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Sample records for reactor target rod

  1. In-reactor performance of LWR-type tritium target rods

    SciTech Connect

    Lanning, D.D. ); Paxton, M.M. ); Crumbaugh, L. )

    1992-01-01

    Pacific Northwest Laboratory has conducted several 1-yr irradiation tests of light water reactor-type tritium target rods. These tests have been sponsored by the U.S. Department of Energy's Office of New Production Reactors. The first test, designated water capsule-1 (WC-1), was conducted in the Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory from November 1989 to December 1990. The test vehicle contained a single 4-ft target rod within a pressurized water capsule. The capsule maintained the rod at pressurized water reactor (PWR)-type water temperature and pressure conditions. Posttest nondestructive examinations of the WC-1 rod involved visual examinations, dimensional checks, gamma scanning, and neutron radiography. The results indicate that the rod maintained the integrity of its pressure seal and was otherwise unaltered both mechanically and dimensionally by its irradiation and posttest handling.

  2. 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)

  3. 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.

  4. Reactor control rod timing system

    DOEpatents

    Wu, Peter T. K.

    1982-01-01

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

  5. 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.

  6. Automatic safety rod for reactors. [LMFBR

    DOEpatents

    Germer, J.H.

    1982-03-23

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

  7. 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.

  8. 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.

  9. 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.

  10. 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.

  11. 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.

  12. 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.

  13. Fuel rod retention device for a nuclear reactor

    DOEpatents

    Hylton, Charles L.

    1984-01-01

    A device is described for supporting a nuclear fuel rod in a fuel rod assembly which allows the rod to be removed without disturbing other rods in the assembly. A fuel rod cap connects the rod to a bolt which is supported in the assembly end fitting by means of a locking assembly. The device is designed so that the bolt is held securely during normal reactor operation yet may be easily disengaged and the fuel rod removed when desired.

  14. 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.

  15. 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.

  16. 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.

  17. 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.

  18. 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.

  19. In-reactor performance of LWR-type tritium targe rods

    SciTech Connect

    Lanning, D.D.; Paxton, M.M.; Crumbaugh, L.

    1992-06-01

    Pacific Northwest Laboratory (PNL) has conducted several one-year irradiation tests of light-water reactor (LWR)-type tritium target rods. This report discusses these tests which have been sponsored by DOE`s Office of New Production Reactors. The first test, designated water capsule-I (WC-1), was conducted in the Advanced Test Reactor (ATR) at DOE`s Idaho National Engineering Laboratory from November 1989 to December 1990. The test vehicle contained a single four-foot target rod within a pressurized water capsule. The capsule maintained the rod at PWR-type water temperature and pressure conditions.

  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. Regulation of nuclear reactors by duplexed control rods; Linearized analysis

    SciTech Connect

    Haidar, N.H.S. ); Diab, H.B. )

    1991-10-01

    In this paper the dynamic behavior of a duplex control rod for a nuclear reactor is analyzed by linearized modeling. A simulation-aided design for this duplex, which has performance and safety characteristics that are improved over those of conventional single control rods with the same stability margins, is reported.

  5. 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.

  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. Boiling water reactor radiation shielded Control Rod Drive Housing Supports

    SciTech Connect

    Baversten, B.; Linden, M.J.

    1995-03-01

    The Control Rod Drive (CRD) mechanisms are located in the area below the reactor vessel in a Boiling Water Reactor (BWR). Specifically, these CRDs are located between the bottom of the reactor vessel and above an interlocking structure of steel bars and rods, herein identified as CRD Housing Supports. The CRD Housing Supports are designed to limit the travel of a Control Rod and Control Rod Drive in the event that the CRD vessel attachement went to fail, allowing the CRD to be ejected from the vessel. By limiting the travel of the ejected CRD, the supports prevent a nuclear overpower excursion that could occur as a result of the ejected CRD. The Housing Support structure must be disassembled in order to remove CRDs for replacement or maintenance. The disassembly task can require a significant amount of outage time and personnel radiation exposure dependent on the number and location of the CRDs to be changed out. This paper presents a way to minimize personal radiation exposure through the re-design of the Housing Support structure. The following paragraphs also delineate a method of avoiding the awkward, manual, handling of the structure under the reactor vessel during a CRD change out.

  10. Control Rod Reactivity Curves for the Annular Core Research Reactor

    NASA Astrophysics Data System (ADS)

    Depriest, K. Russell; Kajder, Karen C.; Frye, Jason N.; Denman, Matthew R.

    2009-08-01

    Experiments were conducted at the Annular Core Research Reactor (ACRR) to increase the fidelity of the control rod integral reactivity worth curve. This experiment series was designed to refine the integral reactivity curve used for pulse yield prediction and eliminate the need for operator compensation in the pulse setup. The experiment series consisted of delayed critical and positive period measurements with various ACRR cavity configurations. An improved integral reactivity worth curve for the ACRR control rods has been constructed using the positive period measurements, the delayed critical measurements, and radiation transport modeling of the reactor. A series of prompt period measurements is used to validate that the new control rod curve more accurately predicts the energy yield of the pulse operations. The new reactivity worth curve is compared with the current curve that was developed using traditional approaches.

  11. 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)

  12. 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.

  13. 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.

  14. 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.

  15. Subchannel thermal-hydraulic modeling of an APT tungsten target rod bundle

    SciTech Connect

    Hamm, L.L.; Shadday, M.A. Jr.

    1997-09-01

    The planned target for the Accelerator Production of Tritium (APT) neutron source consists of an array of tungsten rod bundles through which D{sub 2}O coolant flows axially. Here, a scoping analysis of flow through an APT target rod bundle was conducted to demonstrate that lateral cross-flows are important, and therefore subchannel modeling is necessary to accurately predict thermal-hydraulic behavior under boiling conditions. A local reactor assembly code, FLOWTRAN, was modified to model axial flow along the rod bundle as flow through three concentric heated annular passages.

  16. 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).

  17. 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.

  18. 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.

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

    SciTech Connect

    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. 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.

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

    DOE PAGES

    Christon, Mark A.; Lu, Roger; Bakosi, Jozsef; ...

    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 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

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

    SciTech Connect

    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. 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.

  2. 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.

  3. Analysis of High Temperature Reactor Control Rod Worth for the Initial and Full Core

    NASA Astrophysics Data System (ADS)

    Oktajianto, Hammam; Setiawati, Evi; Anam, Khoirul; Sugito, Heri

    2017-01-01

    Control rod is one important component in a nuclear reactor. In nuclear reactor operations the control rod functions to shut down the reactor. This research analyses ten control rods worth of HTR (High Temperature Reactor) at initial and full core. The HTR in this research adopts HTR-10 China and HTR- of pebble bed. Core calculations are performed by using MCNPX code after modelling the entire parts of core in condition of ten control rods fully withdrawn, all control rods in with 20 cm ranges of depth and the use of one control rod. Pebble bed and moderator balls are distributed in the core zone using a Body Centred Cubic (BCC) lattice by ratio of 57:43. The research results are obtained that the use of one control rod will decrease the reactor criticality of 2.04±0.12 %Δk/k at initial core and 1.57±0.10 %Δk/k at full core. The deeper control rods are in, the lesser criticality of reactor is with reactivity of ten control rods of 16.41±0.11 %Δk/k at initial core and 15.43±0.11 %Δk/k at full core. The results show that the use of ten control rods at full core will keep achieving subcritical condition even though the reactivity is smaller than reactivity at initial core.

  4. Experimental analysis of liquid-metal reactor scram rod kinematic characteristics

    NASA Astrophysics Data System (ADS)

    Konovalenko, F. D.; Kondrashov, S. I.

    2017-01-01

    This article represents the results of computational and experimental research of liquid-metal research reactor control rod kinematics. In this research liquid-metal coolant (sodium) was simulated by water. Investigation of control rod scram-mode movement duration and investigation of velocity of movable parts near the bump of damper are the purposes of this research. Also mathematic simulation of control rod movement in scram mode was performed. Computational results for some modes of water circulation comply with experimental results well. Results of this work will be used for tests of scram rod drive of above-named research reactor. It will significantly simplify the scram rod drive testing stand construction.

  5. 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.

  6. 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)

  7. 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)

  8. 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.

  9. 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)

  10. 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.

  11. 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.

  12. CASL virtual reactor predictive simulation: Grid-to-Rod Fretting wear

    NASA Astrophysics Data System (ADS)

    Lu, Roger Y.; Karoutas, Zeses; Sham, T.-L.

    2011-08-01

    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.

  13. 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.

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

    SciTech Connect

    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 effective way to evaluate contact forces and provide guidance for simulative bench fretting-impact tests.

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

    DOE PAGES

    Jiang, Hao; Qu, Jun; Lu, Roger Y.; ...

    2016-06-10

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

  16. 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.

  17. 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.

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

    SciTech Connect

    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 effective way to evaluate contact forces and provide guidance for simulative bench fretting-impact tests.

  19. 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.

  20. 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.

  1. A model of the control rod housing overflow incident in a Savannah River Production Reactor

    SciTech Connect

    Shadday, M.A. Jr.; Flach, G.P.; Steimke, J.L.; Koffman, L.D.; Askew, N.M.

    1993-05-01

    On May 25, 1992, K-reactor at the Savannah River Site was being configured for start up when a D20 leak of unknown origin was detected on the top of the reactor. The reactor was immediately shutdown, and reactor start up was put on hold until the cause of the leak was determined and the problem eliminated. The source of the leak was suspected to be overflow of the control rod guide tubes, and on May 27 hydraulic tests in the laboratory confirmed this and identified the cause to be unanticipated secondary circulations in the control rod housings. The modelling effort, that in conjunction with the experimental program, identified a temporary fix that allowed K-reactor to be restarted two weeks after the initial overflow incident is herein described.

  2. 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.

  3. Apparent target strength in long-rod penetration

    SciTech Connect

    Godwin, R.P.; Chapyak, E.J.

    1996-03-01

    The authors investigate the apparent enhancement of target strength in the steady-state Tate model of long-rod penetration. They show that computing the effective area over which the target behaves as a fluid provides an explanation of the effective 1-D target strength measured empirically. Expressing the effective target strength as R{sub t} = a {times} Y{sub t}, they postulate that a = A{sub e}/A{sub p}, where Y{sub t} is the nominal strength; A{sub e} is the effective target fluid cross-sectional area and A{sub p} the projectile cross-sectional area. For the case of a rod and projectile of the same material, they use the Tate model together with the jet model of Birkhoff et al. to show a {approx} 4 is likely. Simultaneously satisfying Newton`s Second Law and the Tate model yields a very general derivation of a = 4. By explicitly including strength terms in both the Tate equation and Newton`s Second Law, an even more general a = f(v,{rho}{sub p},{rho}{sub t},Y{sub p},Y{sub t}) can be derived.

  4. 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.

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

    DOE PAGES

    Bakosi, J.; Christon, M. A.; Lowrie, R. B.; ...

    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

  6. 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.

  7. 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.

  8. Consistent neutron-physical and thermal-physical calculations of fuel rods of VVER type reactors

    NASA Astrophysics Data System (ADS)

    Tikhomirov, Georgy; Saldikov, Ivan; Ternovykh, Mikhail; Gerasimov, Alexander

    2017-09-01

    For modeling the isotopic composition of fuel, and maximum temperatures at different moments of time, one can use different algorithms and codes. In connection with the development of new types of fuel assemblies and progress in computer technology, the task makes important to increase accuracy in modeling of the above characteristics of fuel assemblies during the operation. Calculations of neutronphysical characteristics of fuel rods are mainly based on models using averaged temperature, thermal conductivity factors, and heat power density. In this paper, complex approach is presented, based on modern algorithms, methods and codes to solve separate tasks of thermal conductivity, neutron transport, and nuclide transformation kinetics. It allows to perform neutron-physical and thermal-physical calculation of the reactor with detailed temperature distribution, with account of temperature-depending thermal conductivity and other characteristics. It was applied to studies of fuel cell of the VVER-1000 reactor. When developing new algorithms and programs, which should improve the accuracy of modeling the isotopic composition and maximum temperature in the fuel rod, it is necessary to have a set of test tasks for verification. The proposed approach can be used for development of such verification base for testing calculation of fuel rods of VVER type reactors

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

    NASA Astrophysics Data System (ADS)

    Cutic, Avdo

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

  10. 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.

  11. 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

  12. 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.

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

    DOE PAGES

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

    2016-10-01

    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

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

    SciTech Connect

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

    2016-10-01

    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 criticality 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.

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

    SciTech Connect

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

    2016-10-01

    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 criticality 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.

  16. 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.

  17. 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.

  18. 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.

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

    DOE PAGES

    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

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

    SciTech Connect

    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 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.

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

    SciTech Connect

    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 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.

  2. 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.

  3. 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.

    2016-11-01

    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.

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

    SciTech Connect

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

    1980-01-01

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

  5. 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

  6. 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.

  7. 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.

  8. Cone and rod cells have different target preferences in vitro as revealed by optical tweezers

    PubMed Central

    Clarke, Robert J.; Högnason, Kormákur; Brimacombe, Michael

    2008-01-01

    Purpose When neural circuits are damaged in adulthood, regenerating and sprouting processes must distinguish appropriate targets to recreate the normal circuitry. We tested the ability of adult nerve cells to target specific cells in culture using the retina as a model system. Methods Under sterile culture conditions, retinal cells, isolated from tiger salamander retina, were micromanipulated with optical tweezers to create pairs of first-order photoreceptor cells with second- or third-order retinal neurons. The development of cell contact and presynaptic varicosities, the direction and amount of neuritic growth, and nerve cell polarity were assessed after seven days in vitro. Cultures were labeled for rod opsin to distinguish rod from cone cells and for the alpha subunit of the trimeric G protein Go (Goα) to identify cone-dominated and mixed rod-cone ON bipolar cells. Results Quantitative analysis of growth demonstrated that target preferences were cell-specific: Cone cells preferred second-order bipolar cells, whereas rod cells grew toward third-order neurons, which include amacrine and ganglion cells. In addition, when rod cells grew toward bipolar cells, they chose an abnormally high number of Goα-positive bipolar cells. These growth patterns were not affected by tweezers manipulation or the amount of growth. Cell orientation of the photoreceptor also did not affect preferences: Cells oriented away from dendritic processes could reorient their axonal pole toward the target cell. Conclusions Cone cells preferred normal partners, and rod cells preferred novel partners. These intrinsic preferences indicate that adult nerve cells can have differing capacities for targeting even if they come from the same cell class. Further, these differences may help explain the patterns of photoreceptor sprouting seen in retinal degeneration in which rod, but not cone, cells invade the inner retinal layers where third-order neurons are located. PMID:18432315

  9. 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.

  10. 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.

  11. 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)

  12. 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.

  13. Pretest Predictions of Long-Rod Interactions with Armor Technology Targets

    DTIC Science & Technology

    1994-04-01

    July 2 (1993). I 39. Y. Partom, "Ceramic Armor Resistance to Long-Rod Penetration (R.) and its Dependence on Projectile Velocity," Report No. IAT.R...I * AD - A281 384I Jllllll I PRETEST PREDICTIONS OF LONG-ROD INTERACTIONS WITH ARMOR TECHNOLOGY TARGETS DTIC I ELECTE Charles E. Anderson, Jr. JUL... Armor Technology Targets 6. AUTHR(S) j A 9-~-O Charles E. Anderson, Jr. DAvid L. Littlefield 7. PERFORMING ORGANIZATION NAME(S) AND ADDRESS(ES) 8

  14. Preliminary design studies for an iridium rod target at the BNL-AGS

    SciTech Connect

    Ludewig, H.; Hastings, J.; Montanez, P.; Todosow, M.

    1998-12-31

    The BNL-AGS is an intense source of 24 GeV protons. It is proposed to explore the potential to use these protons as the driver for a Pulsed Spallation Neutron Source target. The proposed target design is based on an edge cooled iridium rod concept--similar to the anti-proton production target which operated reliably at CERN under similar conditions. Lead, lead fluoride, and beryllium are investigated as possible reflector materials, and ambient temperature light water and 80 K light water ice are proposed as initial moderator materials. Both moderators are decoupled by cadmium containing moderator chamber walls. The small size of the target has the advantage that the moderators can be placed close to the target (resulting in a bright source), and since a large fraction of the radioactive inventory is contained in the iridium rod, removal and disposition of this inventory should be relatively simple and inexpensive.

  15. 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.

    2017-01-01

    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.

  16. Target-fueled nuclear reactor for medical isotope production

    DOEpatents

    Coats, Richard L.; Parma, Edward J.

    2017-06-27

    A small, low-enriched, passively safe, low-power nuclear reactor comprises a core of target and fuel pins that can be processed to produce the medical isotope .sup.99Mo and other fission product isotopes. The fuel for the reactor and the targets for the .sup.99Mo production are the same. The fuel can be low enriched uranium oxide, enriched to less than 20% .sup.235U. The reactor power level can be 1 to 2 MW. The reactor is passively safe and maintains negative reactivity coefficients. The total radionuclide inventory in the reactor core is minimized since the fuel/target pins are removed and processed after 7 to 21 days.

  17. 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.

  18. 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.

  19. Evaluation of the tritium content in light water reactor control and absorber rods to obtain data for the fuel cycle backend

    SciTech Connect

    Bleier, A.; Neeb, K.H.; Gelfort, E.; Mischke, J.

    1986-08-01

    Tritium inventories and tritium distribution have been determined in boron glass absorber rods discharged from a pressurized water reactor first-cycle core and in spent boron carbide (B/sub 4/C) control rods from a boiling water reactor. The total tritium inventory in the boron glass absorber rods from the Stade nuclear reactor amounts to approx. =8.0 x 10/sup 10/ Bq (2.2 Ci) per rod. Of this, 99.6% was fixed in the boron glass itself and 0.4% in the Al/sub 2/O/sub 3/ pellets. The 4 x 10/sup -3/% fractions in the tube cladding and support pipe and the 1 x 10/sup -2/% fraction in the fill gas accounted for an insignificant part of the total tritium inventory of the rod. This experimentally determined tritium inventory was a factor of 5 larger than that suggested by the calculated estimate. The discrepancy between analyzed and calculated values can be explained by tritium formation from lithium impurities in the boron glass, where a 30-ppm lithium content would be adequate for this tritium inventory to be generated by the reaction /sup 6/Li(n,..cap alpha..)/sup 3/H. Evaluation of the B/sub 4/C control rods from the Lingen nuclear reactor after 3 yr of operation gave a 3.2 x 10/sup 10/Bq(0.85-Ci)tritium inventory per B/sub 4/C rod, while the total tritium inventory for a control rod assembly containing 60 B/sub 4/C rods was approx. =1.9 x 10/sup 12/ Bq (50 Ci). The tritium generated was essentially bound 100% in the B/sub 4/C, since the hulls contained only 6 x 10/sup -3/% and the fill gas only 2 x 10/sup -4/%.

  20. Subchannel Thermal-Hydraulic Experimental Program (STEP). Volume 1. Mixing in a pressurized water reactor (PWR) rod bundle. Final report

    SciTech Connect

    Barber, A.R.; Zielke, L.A.

    1980-08-01

    This volume describes an experiment that was performed to determine the mixing characteristics of a pressurized water reactor (PWR) rod bundle. The objective of this project was to improve the subchannel computer code models of the reactor core. The experimental technique was isokinetic subchannel withdrawal of the entire flow from two sample subchannels. Once withdrawn, the sample fluid was condensed and its enthalpy was measured by regenerative heat exchange calorimetry. The test bundle was a 4 x 6 electrically heated array with a 50% power upset. The COBRA IIIC code was used to model the experiment and to determine the value of the thermal mixing coefficient, ..beta.., that was necessary to predict the measured results. Both single- and two-phase data were obtained over a range of PWR operating conditions. The results indicate that both single- and two-phase mixing is small. The COBRA model predicts the enthalpy data using a turbulent mixing coefficient, ..beta.. approx. = 0.002.

  1. Analysis of the Noneroding Penetration of Tungsten Alloy Long Rods into Aluminum Targets

    DTIC Science & Technology

    2003-09-01

    Solid Mechanics Team for lending photographic support to this effort. The author would like to graciously thank Messrs. Dan Scheffler, Konrad Frank , and...at higher impact velocities. Kinslow (8) noted in the 1960s the disparity between the low-velocity and high- velocity penetration behaviors...transition, the penetration-mode variation of rod strength and target resistance parameters, Y and H, respectively. Frank (26) modeled the penetration

  2. miR Cluster 143/145 Directly Targets Nrl and Regulates Rod Photoreceptor Development.

    PubMed

    Sreekanth, Sreekumaran; Rasheed, Vazhanthodi A; Soundararajan, Lalitha; Antony, Jayesh; Saikia, Minakshi; Sivakumar, Krishnankutty Chandrika; Das, Ani V

    2016-11-23

    Retinal histogenesis requires coordinated and temporal functioning of factors by which different cell types are generated from multipotent progenitors. Development of rod photoreceptors is regulated by multiple transcription factors, and Nrl is one of the major factors involved in their fate specification. Presence or absence of Nrl at the postnatal stages decides the generation of cone photoreceptors or other later retinal cells. This suggests the need for regulated expression of Nrl in order to accelerate the generation of other cell types during retinal development. We found that miR cluster 143/145, comprising miR-143 and miR-145, targets and imparts a posttranscriptional inhibition of Nrl. Expression of both miRNAs was differentially regulated during retinal development and showed least expression at PN1 stage in which most of the rod photoreceptors are generated. Downregulation of rod photoreceptor regulators and markers upon miR cluster 143/145 overexpression demonstrated that this cluster indeed negatively regulates rod photoreceptors. Further, we prove that Nrl positively regulates miR cluster 143/145, thus establishing a feedback loop regulatory mechanism. This may be one possible mechanism by which Nrl is posttranscriptionally regulated to facilitate the generation of other cell types in retina.

  3. Super-characteristic x-ray generator utilizing a pipe and rod target

    NASA Astrophysics Data System (ADS)

    Sato, Eiichi; Tanaka, Etsuro; Mori, Hidezo; Kawai, Toshiaki; Inoue, Takashi; Ogawa, Akira; Izumisawa, Mitsuru; Takahashi, Kiyomi; Sato, Shigehiro; Ichimaru, Toshio; Takayama, Kazuyoshi

    2006-08-01

    This generator consists of the following components: a constant high-voltage power supply, a filament power supply, a turbomolecular pump, and an x-ray tube. The x-ray tube is a demountable diode which is connected to the turbomolecular pump and consists of the following major devices: a tungsten hairpin cathode (filament), a focusing (Wehnelt) electrode, a polyethylene terephthalate x-ray window 0.25 mm in thickness, a stainless-steel tube body, a pipe target, and a rod target. The pipe and rod targets are useful for forming linear and cone beams, respectively. In the x-ray tube, the positive high voltage is applied to the anode (target) electrode, and the cathode is connected to the tube body (ground potential). In this experiment, the tube voltage applied was from12 to 20 kV, and the tube current was regulated to within 0.10 mA by the filament temperature. The exposure time is controlled in order to obtain optimum x-ray intensity. The electron beams from the cathode are converged to the target by the focusing electrode, and clean K-series characteristic x-rays are produced through the focusing electrode without using a filter. The x-ray intensities of the pipe and rod targets were 1.29 and 4.28 μGy/s at 1.0 m from the x-ray source with a tube voltage of 15 kV and a tube current of 0.10 mA, and quasi-monochromatic radiography was performed using a computed radiography system.

  4. Prediction of the lifetime of the elements of the safety and control rods of nuclear reactors

    SciTech Connect

    Voskoboinikov, V.V.; Emel'yanov, I. Ya.; Lineva, A.F.; Pushkin, S.N.; Semchenko, E.L.; Usov, P.P.

    1987-07-01

    The authors construct a mathematical model based on the analytical solution to such parameters as magnetic flux, alignment, sliding friction, erosion, abrasion, corrosion, and irradiation for the purpose of predicting the service life of electromagnetically driven control rods and their drives. The analytical solution was verified experimentally on a bench simulation where it was found that incorrect assembly and alignment not only serve as the largest contributors to shortened service life of the rods and drives but also render the calculations of the model invalid.

  5. 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.

  6. 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.

  7. Dynamic Inversion Enables External Magnets To Concentrate Ferromagnetic Rods to a Central Target

    PubMed Central

    2015-01-01

    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. PMID:25457292

  8. 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.

  9. 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

  10. Control rod drive

    SciTech Connect

    Hawke, Basil C.

    1986-01-01

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

  11. 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.

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

    NASA Astrophysics Data System (ADS)

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

    1998-07-01

    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.

  13. CONTROL ROD

    DOEpatents

    Walker, D.E.; Matras, S.

    1963-04-30

    This patent shows a method of making a fuel or control rod for a nuclear reactor. Fuel or control material is placed within a tube and plugs of porous metal wool are inserted at both ends. The metal wool is then compacted and the tube compressed around it as by swaging, thereby making the plugs liquid- impervious but gas-pervious. (AEC)

  14. 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.

  15. 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.

  16. 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.

  17. Variable reluctance electric motor for the translation of control rods in a nuclear reactor

    SciTech Connect

    Guedj, F.; Defaucheux, J.; Wiart, A.

    1985-04-02

    In accordance with the invention, the motor includes: a rotor component composed of a shaft of magnetic material provided with regularly spaced projections, and a stator component comprising at least three multipolar stators encircling the shaft consecutively, each polar core having cavities on the shaft side so as to constitute projections of the same width and spacing as those of the shaft, with the spacing between stators being such that the projections of the shaft and those of the polar cores face one another for only one of the stators. A tight casing is arranged to intersect the stator magnetic circuit and magnetic parts extend the polar core to the interior of the casing. Application to nuclear reactors.

  18. Tumor Targeting and Imaging in Live Animals with Functionalized Semiconductor Quantum Rods

    PubMed Central

    Yong, Ken-Tye; Hu, Rui; Roy, Indrajit; Ding, Hong; Vathy, Lisa A.; Bergey, Earl J.; Mizuma, Masamichi; Maitra, Anirban; Prasad, Paras N.

    2009-01-01

    In this contribution, we demonstrate that highly luminescent CdSe/CdS/ZnS quantum rods (QRs) coated with PEGylated phospholipids and conjugated with cyclic RGD-peptide can be successfully used for tumor targeting and imaging in live animals. The design of these targeted luminescent probes involves encapsulating hydrophobic CdSe/CdS/ZnS QRs with PEGylated phospholipids, followed by conjugation of these PEGylated phospholipids to ligands that specifically target the tumor vasculature. In vivo optical imaging studies in nude mice bearing pancreatic cancer xenografts, both subcutaneous and orthotopic, indicate that the QR probes accumulate at tumor sites via the cyclic RGD-peptides on the QR surface binding to theαVβ3 integrins overexpressed in the tumor vasculature, following systemic injection. In vivo tumor detection studies showed no adverse effects even at a dose roughly 6.5 times higher than has been reported for in vivo imaging studies using QDs. Cytotoxicity studies indicated absence of any toxic effect in the cellular and tissue levels arising from functionalized QRs. These results demonstrate the vast potential of QRs as bright, photostable, and biocompatible luminescent probes for the early diagnosis of cancer. PMID:20160901

  19. 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.

  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. Method of targeted delivery of laser beam to isolated retinal rods by fiber optics

    PubMed Central

    Sim, Nigel; Bessarab, Dmitri; Jones, C. Michael; Krivitsky, Leonid

    2011-01-01

    A method of controllable light delivery to retinal rod cells using an optical fiber is described. Photo-induced current of the living rod cells was measured with the suction electrode technique. The approach was tested with measurements relating the spatial distribution of the light intensity to photo-induced current. In addition, the ion current responses of rod cells to polarized light at two different orientation geometries of the cells were studied. PMID:22076256

  2. Method of targeted delivery of laser beam to isolated retinal rods by fiber optics.

    PubMed

    Sim, Nigel; Bessarab, Dmitri; Jones, C Michael; Krivitsky, Leonid

    2011-11-01

    A method of controllable light delivery to retinal rod cells using an optical fiber is described. Photo-induced current of the living rod cells was measured with the suction electrode technique. The approach was tested with measurements relating the spatial distribution of the light intensity to photo-induced current. In addition, the ion current responses of rod cells to polarized light at two different orientation geometries of the cells were studied.

  3. 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.

  4. 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.

  5. Hydraulic Actuator for Ganged Control Rods

    NASA Technical Reports Server (NTRS)

    Thompson, D. C.; Robey, R. M.

    1986-01-01

    Hydraulic actuator moves several nuclear-reactor control rods in unison. Electromagnetic pump pushes liquid lithium against ends of control rods, forcing them out of or into nuclear reactor. Color arrows show lithium flow for reactor startup and operation. Flow reversed for shutdown. Conceived for use aboard spacecraft, actuator principle applied to terrestrial hydraulic machinery involving motion of ganged rods.

  6. Multiplexing thermography for International Thermonuclear Experimental Reactor divertor targets

    SciTech Connect

    Itami, K.; Sugie, T.; Vayakis, G.; Walker, C.

    2004-10-01

    The concept of multiplexing thermography is applied to the design of the divertor thermography system for International Thermonuclear Experimental Reactor (ITER). The combination of the front mirror with multiellipticity and a Czerney-Turner spectrometer with a 0.2 mm pitched multichannel detector enables a spatial resolution of 3 mm and a time resolution of 20 {mu}s above a target temperature of 300 deg. C to be achieved. This should be sufficient to measure ELM heat fluxes to the targets in ITER. To satisfy the measurement requirement, it is very important to keep an accurate alignment around the optical axis against movement of the vessel during the plasma discharges. Several key engineering problems, such as the survivability of components against mirror coating by redeposited divertor material, remain to be solved. Potential solutions have been identified.

  7. Penetration of Rod Projectiles in Semi-Infinite Targets: A Validation Test for Eulerian X-FEM in Alegra

    DTIC Science & Technology

    2013-03-12

    the semi-infinite system. The equation of state ( EOS ) material properties for the projectile rod and target plate were obtained from Sesame tabular...libraries, using the “KEOS” (Kerley) interface in ALEGRA, as listed in Table 2. The KEOS Sesame model is an implementation of the Sesame EOS in...exists which establishes any certain combination of material response models for this problem as the most accurate. The tabular Sesame EOS models are

  8. 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.

  9. 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)

  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.

    2016-11-01

    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. Evaluation of a uranium zirconium hydride fuel rod option for conversion of the MIT research reactor (MITR) from highly-enriched uranium to low-enriched uranium

    DOE PAGES

    Dunn, F. E.; Wilson, E. H.; Feldman, E. E.; ...

    2017-03-23

    The conversion of the Massachusetts Institute of Technology Reactor (MITR) from the use of highly-enriched uranium (HEU) fuel-plate assemblies to low-enriched uranium (LEU) by replacing the HEU fuel plates with specially designed General Atomics (GA) uranium zirconium hydride (UZrH) LEU fuel rods is evaluated in this paper. The margin to critical heat flux (CHF) in the core, which is cooled by light water at low pressure, is evaluated analytically for steady-state operation. A form of the Groeneveld CHF lookup table method is used and described in detail. A CHF ratio of 1.41 was found in the present analysis at 10more » MW with engineering hot channel factors included. Therefore, the nominal reactor core power, and neutron flux performance, would need to be reduced by at least 25% in order to meet the regulatory requirement of a minimum CHF ratio of 2.0.« less

  12. Control rod

    SciTech Connect

    Dixon, R.C.; Cearley, J.E.; VanDiemen, P.; Sayre, E.D.; Gordon, G.M.

    1990-02-20

    This patent describes in a nuclear reactor control rod having elongate planar members for absorption of neutrons within a nuclear reactor for control of the nuclear reaction, the elongate planar members being formed of a plurality of tubes arranged side-by-side in abutting contact and joined together. The tube comprises: a tube defining a cylindrical pressure vessel for containment of neutron absorbing poisons. The tube defining constant side wall thickness sufficient to define there within a cylindrical volume for the containment of neutron absorbing poisons and having sufficient side wall thickness to retain the poisons under all anticipated pressures from decomposition of the neutron absorbing poisons; and the tube integrally defining in addition to the cylindrical pressure vessel four discrete right angle corner sections placed at 90{degree} intervals to the side wall of the constant side wall thickness tube; and neutron absorbing poisons confined within the tube for absorption of neutrons for control of the nuclear reaction.

  13. 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.

  14. Long rod penetration test of hot isostatically pressed Ti-based targets

    NASA Astrophysics Data System (ADS)

    Nesterenko, Vitali F.; Indrakanti, Sastry S.; Brar, Singh; Gu, YaBei

    2000-04-01

    Hot Isostatic Pressing (HIP) is one of the most efficient techniques to produce high quality materials from powders. Nevertheless there is a shortage of data on high-strain-rate behavior and penetration resistance of such materials. In this paper the results of penetration test with tungsten (93%) heavy alloy penetrators of solid and porous composite samples of Ti-6Al-4V alloy with different microstructures (Widmanstatten pattern and equiaxed) are presented. Penetration depth for HIPed materials is smaller than in baseline samples of Ti-6Al-4V alloy (forged rod MIL-T-9047G). Composite materials with alumina rods and tubes filled with B4C powders demonstrated a new features of penetration: projectile deflection with self sealing of hole and forced shear localization caused by tubes fracture. The results demonstrate the applicability of HIPing for Ti-based armor materials.

  15. Inverted Control Rod Lock-In Device

    DOEpatents

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

    1962-12-01

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

  16. 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.

  17. Comparing targeted and non-targeted high-resolution mass spectrometric approaches for assessing advanced oxidation reactor performance.

    PubMed

    Parry, Emily; Young, Thomas M

    2016-11-01

    High resolution mass spectrometry (HR-MS) offers the opportunity to track large numbers of non-target analytes through water treatment processes, providing a more comprehensive view of reactor performance than targeted evaluation. Both approaches were used to evaluate the performance of a pilot scale advanced oxidation process (AOP) employing ultraviolet light and hydrogen peroxide (UV/H2O2) to treat municipal wastewater effluent. Twelve pharmaceuticals and personal care products were selected as target compounds and added to reactor influent. Target compound removal over a range of flow rates and hydrogen peroxide addition levels was assessed using a liquid chromatograph combined with a quadrupole time-of-flight mass spectrometer (LC-qTOF-MS). Target compound removals were used to determine hydroxyl radical concentrations and UV fluence under pilot scale conditions. The experiments were also analyzed using a nontarget approach, which identified "molecular features" in either reactor influent or effluent. Strong correlation (r = 0.94) was observed between target compound removals calculated using the targeted and non-targeted approaches across the range of reactor conditions tested. The two approaches also produced consistent rankings of the performance of the various reactor operating conditions, although the distribution of compound removal efficiencies was usually less favorable with the broader, nontarget approach. For example, in the UV only treatment 8.3% of target compounds and 2.2% of non-target compounds exhibited removals above 50%, while 100% of target compounds and 74% of non-target compounds exhibited removals above 50% in the best condition tested. These results suggest that HR-MS methods can provide more holistic evaluation of reactor performance, and may reduce biases caused by selection of a limited number of target compounds. HR-MS methods also offer insights into the composition of poorly removed compounds and the formation of transformation

  18. 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.

  19. 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.

  20. 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.

  1. 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.

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

    DOEpatents

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

    1982-08-19

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

  3. 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

    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.

  4. 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)

  5. 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.

  6. Solid-state flow, mechanical alloying, and melt-related phenomena for [001] single-crystal tungsten ballistic rod penetrators interacting with steel targets

    NASA Astrophysics Data System (ADS)

    Pizana, Carlos

    This research program consists of a detailed microstructural investigation of in-target, single-crystal [001], clad (with Inconel 718) and unclad, W long-rod, ballistic penetrators. The rods were shot into rolled homogeneous armor (RHA) steel targets approximately 76 mm in thickness at impact velocities ranging from 1100 m/s to 1350 m/s. A comprehensive microstructural overview of the penetration process was obtained from this investigation. Solid-state flow/erosion, solid-state target/rod mixing as well as influencing factors such as strain rate, penetration performance, cladding interference and the interaction between target and projectile were emphasized. Some of the microstructural features observed, including deformation twins, cleaving, adiabatic shear bands and DRX support an overall solid-state penetration process. Furthermore they provide for a unifying perspective for the applicability of the hydrodynamic paradigm (DOP ≈ l∘rp/rt ) and earlier mechanistic erosion approaches. DRX and grain growth within adiabatic shear bands observed at specific high strain/strain-rate zones within the rods suggest that the projectile erodes by means of these microstructures in a solid-state form. This erosion process contributes to the performance of the rod by either allowing optimum flow of rod material which would increase penetration depth, or by maximizing rod material consumption which would reduce it. Since flow and/or erosion are also necessary in the target for perforation to occur, it is not surprising that the erosion process in the target was observed to mirror the one in the projectile. That is both target and projectile developed erosion zones with DRX facilitating the extreme deformation via dense overlapping shear band formation. Mechanical alloying and/or mixing of the target (steel) and rod (W, or W-Inconel 718) was also observed and investigated. Selective etching techniques as well as energy-dispersive x-ray mapping revealed unambiguous evidence of

  7. 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.

  8. Properties of reactor fuel rod materials at high temperatures: Final summary report: Severe Core Damage Property Tests Program

    SciTech Connect

    Prater, J.T.; Courtright, E.L.

    1987-07-01

    This report summarizes work sponsored by the US Nuclear Regulatory Commission Division of Accident Evaluation to investigate those physical properties that are needed to predict the behavior of fuel-rod assemblies during a loss-of-coolant accident. The results include a determination of the oxidation kinetics of Zircaloy and Zircaloy-uranium oxide mixtures in steam and steam-hydrogen gas mixtures at 1300 to 2400C, viscosity measurements of zirconium-oxide mixtures at 1800 to 2100C, an estimate of the heat of reaction for the dissolution of uranium oxide by molten zirconium at 2000C, and thermal diffusivity measurements on prereacted Zircaloy-uranium oxide mixtures at 800 to 1500C.

  9. Qualification of a Tritium-Producing Target for the light water reactor application

    SciTech Connect

    Apley, W.J.; Beeman, G.H.; Ethridge, J.L.

    1992-06-01

    The Pacific Northwest Laboratory (PNL) currently manages the Light Water Reactor (LWR) Tritium Target Development Project (TTDP) for the Office of New Production Reactors (NP), US Department of Energy. The project`s objective is to demonstrate and qualify a high-temperature LWR tritium target system with fabrication and extraction processes sufficiently confirmed to ensure a deployable system consistent with variable tritium production demands. The project has also examined and reported on technical and institutional issues associated with acquisition and conversion of the yet uncompleted Washington Public Power Supply System Unit I (WNP-1) for tritium production purposes. WNP-1 is a 63% complete, Babcock and Wilcox, 3800 MW thermal, 205 assembly, pressurized water reactor located at Hanford, Washington. WNP-1 has served as the reference LWR plant for the technical evaluation and target development activities. This report discussed the evaluation and development necessary to provide a complete LWR target qualification package.

  10. CTH analyses of steel rod penetration into aluminum and concrete targets with comparisons to experimental data

    SciTech Connect

    Kmetyk, L.N.; Yarrington, P.

    1994-10-01

    Calculational results are presented here for a class of intermediate-velocity penetration problems. The problems of interest involve penetration of moderate-strength target materials by high-strength projectiles. Two series of metal penetration experiments and a series of concrete slab perforation tests were simulated in this study. The computer code used for the calculations was the CTH code, which employs a recently-developed ``boundary layer`` algorithm for treating penetration problems such as these.

  11. Signal sequence cleavage and plasma membrane targeting of the retinal rod NCKX1 and cone NCKX2 Na+/Ca2+ - K+ exchangers.

    PubMed

    Kang, KyeongJin; Schnetkamp, Paul P M

    2003-08-12

    Retinal rod and cone photoreceptors express two distinct Na(+)/Ca(2+)-K(+) exchanger (NCKX) gene products. Both the rod NCKX1 and cone NCKX2 are polytopic membrane proteins thought to contain a putative cleavable signal peptide. A cleavable signal peptide is unusual for plasma membrane proteins; moreover, predictive algorithms suggest the presence of a cleavable signal peptide for all rod NCKX1 proteins and a noncleavable signal anchor for the cone NCKX2 proteins. In this study we have placed a peptide tag at different positions of the NCKX sequence to examine whether the putative signal sequence is indeed cleaved in either NCKX1 or NCKX2 proteins expressed in heterologous systems. The signal peptide was found to be, at least in part, cleaved in dolphin rod NCKX1 and in chicken and human cone NCKX2 expressed in HEK293 cells; no signal peptide cleavage was observed for chicken rod NCKX1 despite the fact that the SignalP predictive algorithm assigned this sequence to have the highest likelihood for a cleavable signal peptide among the three NCKX sequences tested here. For the two NCKX proteins that contained a cleavable signal peptide, only cleaved NCKX protein was found in the plasma membrane of HEK293 cells. Deletion of the signal sequence in both dolphin rod NCKX1 or cone NCKX2 did not affect NCKX protein synthesis but did disrupt plasma membrane targeting as judged from abolition of NCKX function and from lack of surface biotinylation. These results are consistent with delayed signal peptide cleavage for the rod and cone NCKX proteins.

  12. 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.

  13. 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.

  14. NEUTRONIC REACTORS

    DOEpatents

    Wigner, E.P.

    1960-11-22

    A nuclear reactor is described wherein horizontal rods of thermal- neutron-fissionable material are disposed in a body of heavy water and extend through and are supported by spaced parallel walls of graphite.

  15. Targeting of GFP to newborn rods by Nrl promoter and temporal expression profiling of flow-sorted photoreceptors.

    PubMed

    Akimoto, Masayuki; Cheng, Hong; Zhu, Dongxiao; Brzezinski, Joseph A; Khanna, Ritu; Filippova, Elena; Oh, Edwin C T; Jing, Yuezhou; Linares, Jose-Luis; Brooks, Matthew; Zareparsi, Sepideh; Mears, Alan J; Hero, Alfred; Glaser, Tom; Swaroop, Anand

    2006-03-07

    The Maf-family transcription factor Nrl is a key regulator of photoreceptor differentiation in mammals. Ablation of the Nrl gene in mice leads to functional cones at the expense of rods. We show that a 2.5-kb Nrl promoter segment directs the expression of enhanced GFP specifically to rod photoreceptors and the pineal gland of transgenic mice. GFP is detected shortly after terminal cell division, corresponding to the timing of rod genesis revealed by birthdating studies. In Nrl-/- retinas, the GFP+ photoreceptors express S-opsin, consistent with the transformation of rod precursors into cones. We report the gene profiles of freshly isolated flow-sorted GFP+ photoreceptors from wild-type and Nrl-/- retinas at five distinct developmental stages. Our results provide a framework for establishing gene regulatory networks that lead to mature functional photoreceptors from postmitotic precursors. Differentially expressed rod and cone genes are excellent candidates for retinopathies.

  16. Targeting of GFP to newborn rods by Nrl promoter and temporal expression profiling of flow-sorted photoreceptors

    PubMed Central

    Akimoto, Masayuki; Cheng, Hong; Zhu, Dongxiao; Brzezinski, Joseph A.; Khanna, Ritu; Filippova, Elena; Oh, Edwin C. T.; Jing, Yuezhou; Linares, Jose-Luis; Brooks, Matthew; Zareparsi, Sepideh; Mears, Alan J.; Hero, Alfred; Glaser, Tom; Swaroop, Anand

    2006-01-01

    The Maf-family transcription factor Nrl is a key regulator of photoreceptor differentiation in mammals. Ablation of the Nrl gene in mice leads to functional cones at the expense of rods. We show that a 2.5-kb Nrl promoter segment directs the expression of enhanced GFP specifically to rod photoreceptors and the pineal gland of transgenic mice. GFP is detected shortly after terminal cell division, corresponding to the timing of rod genesis revealed by birthdating studies. In Nrl−/− retinas, the GFP+ photoreceptors express S-opsin, consistent with the transformation of rod precursors into cones. We report the gene profiles of freshly isolated flow-sorted GFP+ photoreceptors from wild-type and Nrl−/− retinas at five distinct developmental stages. Our results provide a framework for establishing gene regulatory networks that lead to mature functional photoreceptors from postmitotic precursors. Differentially expressed rod and cone genes are excellent candidates for retinopathies. PMID:16505381

  17. Analysis of the in-vessel control rod guide tube and subpile room shielding design for the advanced neutron source reactor

    SciTech Connect

    Gallmeier, F.X.; Bucholz, J.A.; Engle, W.W. Jr.; Williams, L.R.

    1995-08-01

    An extensive sheilding analysis of the control rod guide tube (CRGT) and the subpile room was performed for the Advanced Neutron Source (ANS) reactor. A two-dimensional model for the CRGT and subpile room was developed. Coupled 39 neutron group and 44 gamma group calculations with the multigroup DORT discrete originates transport code were done using cross sections from the ANSL-V library including photoneutron production. Different shield designs were investigated with a shield thickness of 10 to 15 mm. None of the shields affected the neutron dose rate and gamma dose rate at the top of the subpile room, which were 1 {center_dot} 10{sup 5} mrem/h and 1 {center_dot} 10{sup 3} mrem/h, respectively. An L-shaped cylindrical boral shield positioned around the core pressure boundary tube at the bottom of the reflector vessel with the horizontal part extended over the whole bottom of the reflector vessel reduced the maximal displacements per atom (DPA) level and helium production level in the primary coolant supply adapter and its flange after 40 years of reactor operation from 1 and 500 appm to 5 {center_dot} 10{sup -2} and 2 {center_dot} 10{sup -2} appm compared with the unshielded arrangement. Shields of boral and hafnium with the horizontal part of the shield restricted to a radius of 485 mm gave a maximal DPA of 5 {center_dot} 10{sup -2} and a helium production of up to 20 appm. Heat loads of up to 70 W{center_dot}cm{sup -3} were calculated at the most exposed parts of the shield both for boral and hafnium shields. A depletion/activation analysis of the hafnium shield showed that at the most exposed part of the shield, the naturally occurring isotope {sup 177}Hf is 34% depleted at the end of two years of reactor operation. This high burnup is somewhat balanced by a subsequent buildup of {sup 178}Hf, {sup 179}Hf, and {sup 180}Hf. In all other parts of the shield, the burnup is much smaller.

  18. 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.

  19. 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.

  20. A Computational Study of Segmented Tungsten Rod Penetration into a Thick Steel Target Plate at High Velocities

    NASA Astrophysics Data System (ADS)

    Presnell, M.; Rajendran, A.

    2011-06-01

    This paper presents results from computational simulations of tungsten alloy segmented rod projectiles (SRP) penetrating an RHA semi-infinite target plate at high velocities. For SRP with an aspect ratio (L/D) = 1/8, a loss in penetration efficiency was seen upon successive segment impacts. Numerical simulations of a configuration in which a tungsten heavy alloy SRP penetrated a thick RHA 4340 steel at 2.6 km/s were performed using the 2006 version of the EPIC - a Lagrangian code. The configuration consisted of eight collinear impacts of discs which measured 2 mm thick and 16mm in diameter. The numerical simulations considered a range of parameters including element-particle conversion, spacing and number of fragments, failure criterion, and mesh resolution that influenced the Depth of Penetration (DOP). The EPIC results using the element-to-particle conversion capability in the EPIC code are also compared with open-literature DOP data from simulations using an Eulerian finite element code, AUTODYN for a similar configuration. The present results showed a unique phenomenon of back-flowing ejecta from the crater and fragmented segments penetrating the in-coming subsequent segment. The penetration efficiency seems to be influenced by the back-flowing ejecta. Further computational investigation considered additional simulations with an impact configuration designed to minimize the ejecta effects by using washer-shaped segments; however, the results showed insignificant improvement.

  1. 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.

  2. 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.

  3. Qualification of a tritium-producing target for light water reactor application

    SciTech Connect

    Apley, W.J.; Beeman, G.H.; Ethridge, J.L. )

    1992-01-01

    The Pacific Northwest Laboratory (PNL) currently manages the light water reactor (LWR) tritium target development project (TTDP) for the Office of New Production Reactors, U.S. Department of Energy (DOE). The project's objective is to demonstrate and qualify a high-temperature LWR tritium target system with fabrication and extraction processes sufficiently confirmed to ensure a deployable system consistent with variable tritium production demands. The project has also examined and reported on technical and institutional issues associated with acquisition and conversion of the yet uncompleted Washington Public Power Supply System Unit 1 (WNP-1) for tritium production purposes; WNP-1 is a 63% complete, Babcock and Wilcox, 3,800-MW(thermal), 205 assembly, pressurized water reactor located at Hanford, Washington. WNP-1 has served as the reference LWR plant for the technical evaluation and target development activities. The current project has conducted significant engineering and test activities in support of the evaluation and development necessary to provide a complete LWR target qualification package. The overall performance of the reference design, as measured by achieving in-reactor tritium retention goals, has been quantified. Therefore, the high-temperature LWR tritium target design continues to represent a viable contingency option for future defense production of tritium in the United States.

  4. 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.

  5. Fuel and core testing plan for a target fueled isotope production reactor.

    SciTech Connect

    Coats, Richard Lee; Dahl, James J.; Parma, Edward J., Jr.

    2010-12-01

    In recent years there has been an unstable supply of the critical diagnostic medical isotope 99Tc. Several concepts and designs have been proposed to produce 99Mo the parent nuclide of 99Tc, at a commercial scale sufficient to stabilize the world supply. This work lays out a testing and experiment plan for a proposed 2 MW open pool reactor fueled by Low Enriched Uranium (LEU) 99Mo targets. The experiments and tests necessary to support licensing of the reactor design are described and how these experiments and tests will help establish the safe operating envelop for a medical isotope production reactor is discussed. The experiments and tests will facilitate a focused and efficient licensing process in order to bring on line a needed production reactor dedicated to supplying medical isotopes. The Target Fuel Isotope Reactor (TFIR) design calls for an active core region that is approximately 40 cm in diameter and 40 cm in fuel height. It contains up to 150 cylindrical, 1-cm diameter, LEU oxide fuel pins clad with Zircaloy (zirconium alloy), in an annular hexagonal array on a {approx}2.0 cm pitch surrounded, radially, by a graphite or a Be reflector. The reactor is similar to U.S. university reactors in power, hardware, and safety/control systems. Fuel/target pin fabrication is based on existing light water reactor fuel fabrication processes. However, as part of licensing process, experiments must be conducted to confirm analytical predictions of steady-state power and accident conditions. The experiment and test plan will be conducted in phases and will utilize existing facilities at the U.S. Department of Energy's Sandia National Laboratories. The first phase is to validate the predicted reactor core neutronics at delayed critical, zero power and very low power. This will be accomplished by using the Sandia Critical Experiment (CX) platform. A full scale TFIR core will be built in the CX and delayed critical measurements will be taken. For low power experiments

  6. 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)

  7. 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.

  8. 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.

  9. Cryogenic hydrogen fuel for controlled inertial confinement fusion (formation of reactor-scale cryogenic targets)

    SciTech Connect

    Aleksandrova, I. V.; Koresheva, E. R. Krokhin, O. N.; Osipov, I. E.

    2016-12-15

    In inertial fusion energy research, considerable attention has recently been focused on low-cost fabrication of a large number of targets by developing a specialized layering module of repeatable operation. The targets must be free-standing, or unmounted. Therefore, the development of a target factory for inertial confinement fusion (ICF) is based on methods that can ensure a cost-effective target production with high repeatability. Minimization of the amount of tritium (i.e., minimization of time and space at all production stages) is a necessary condition as well. Additionally, the cryogenic hydrogen fuel inside the targets must have a structure (ultrafine layers—the grain size should be scaled back to the nanometer range) that supports the fuel layer survivability under target injection and transport through the reactor chamber. To meet the above requirements, significant progress has been made at the Lebedev Physical Institute (LPI) in the technology developed on the basis of rapid fuel layering inside moving free-standing targets (FST), also referred to as the FST layering method. Owing to the research carried out at LPI, unique experience has been gained in the development of the FST-layering module for target fabrication with an ultrafine fuel layer, including a reactor- scale target design. This experience can be used for the development of the next-generation FST-layering module for construction of a prototype of a target factory for power laser facilities and inertial fusion power plants.

  10. Cryogenic hydrogen fuel for controlled inertial confinement fusion (formation of reactor-scale cryogenic targets)

    NASA Astrophysics Data System (ADS)

    Aleksandrova, I. V.; Koresheva, E. R.; Krokhin, O. N.; Osipov, I. E.

    2016-12-01

    In inertial fusion energy research, considerable attention has recently been focused on low-cost fabrication of a large number of targets by developing a specialized layering module of repeatable operation. The targets must be free-standing, or unmounted. Therefore, the development of a target factory for inertial confinement fusion (ICF) is based on methods that can ensure a cost-effective target production with high repeatability. Minimization of the amount of tritium (i.e., minimization of time and space at all production stages) is a necessary condition as well. Additionally, the cryogenic hydrogen fuel inside the targets must have a structure (ultrafine layers—the grain size should be scaled back to the nanometer range) that supports the fuel layer survivability under target injection and transport through the reactor chamber. To meet the above requirements, significant progress has been made at the Lebedev Physical Institute (LPI) in the technology developed on the basis of rapid fuel layering inside moving free-standing targets (FST), also referred to as the FST layering method. Owing to the research carried out at LPI, unique experience has been gained in the development of the FST-layering module for target fabrication with an ultrafine fuel layer, including a reactor- scale target design. This experience can be used for the development of the next-generation FST-layering module for construction of a prototype of a target factory for power laser facilities and inertial fusion power plants.

  11. Clinical associations and potential novel antigenic targets of autoantibodies directed against rods and rings in chronic hepatitis C infection

    PubMed Central

    2013-01-01

    Background Chronic hepatitis C virus (HCV) infection is frequently associated with extrahepatic autoimmune disorders while interferon (IFN) and ribavirin treatment may exacerbate these conditions. Autoantibodies from HCV patients identify a novel indirect immunofluorescence (IIF) pattern on HEp-2 cells characterized by cytoplasmic rods and rings (RR). Our objectives were to determine the prevalence and clinical associations of RR autoantibodies in HCV patients, and identify related novel autoantibody targets. Methods Sera from 315 patients with HCV (301 treatment naive, 14 treated with interferon and/or ribavirin) were analyzed for the presence of RR antibodies by IIF on commercially available HEp-2 cell substrates. Antibodies to inosine monophosphate dehydrogenase 2 (IMPDH2) and cytidine triphosphate synthase 1 (CTPS1) were detected by addressable laser bead assay and other potential targets were identified by immunoscreening a protein microarray. Clinical and demographic data including HCV genotype, mode of infection, prior antiviral therapy, and histological findings were compared between RR antibody positive (RR+) and negative (RR-) patients. Results The median age of the HCV cohort was 51 years, 61% were male, and 76% were infected with HCV genotype 1 (G1). Four percent (n=14) had been treated with IFN-based therapy (IFN monotherapy, n=3; IFN/ribavirin, n=11); all had a sustained virologic response. In total, 15 patients (5% of the cohort) were RR+. RR+ and RR- patients had similar demographic and clinical characteristics including age, sex, mode of HCV infection, prevalence of the G1 HCV genotype, and moderate to severe fibrosis. Nevertheless, RR+ patients were significantly more likely than RR- cases to have been treated with IFN-based therapy (33% vs. 3%; adjusted odds ratio 20.5 [95% confidence interval 5.1-83.2]; P<0.0005). Only 1/10 RR positive sera had detectable antibodies to IMPHD2 and none had antibodies to CTPS1. Potentially important autoantibody

  12. 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)

  13. 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)

  14. 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.

  15. 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.

  16. 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.

  17. 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.

  18. 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.

  19. 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.

  20. 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.

  1. 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.

  2. 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.

  3. 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.

  4. 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.

  5. 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.

  6. Tumor Bioimaging: Morphology-Tailoring of a Red AIEgen from Microsized Rods to Nanospheres for Tumor-Targeted Bioimaging (Adv. Mater. 16/2016).

    PubMed

    Li, Yongsheng; Shao, Andong; Wang, Yao; Mei, Ju; Niu, Dechao; Gu, Jinlou; Shi, Ping; Zhu, Weihong; Tian, He; Shi, Jianlin

    2016-04-01

    Y. Li, W. Zhu, and co-workers develop a convenient and versatile "make-up" strategy to modulate the micro-sized rods of a near-infrared-emissive AIEgen probe integrated into nanospheres via a self-assembly encapsulation process, as presented on page 3187. The obtained nanospheres outperform microrods in terms of brightness, photostability, biocompatibility, tumor-accumulation, and targeting ability, making them perfect bioprobes for tumor-targeted bioimaging. © 2016 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  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. REACTORS

    DOEpatents

    Spitzer, L. Jr.

    1961-10-01

    Thermonuclear reactors, methods, and apparatus are described for controlling and confining high temperature plasma. Main axial confining coils in combination with helical windings provide a rotational transform that avoids the necessity of a figure-eight shaped reactor tube. The helical windings provide a multipolar helical magnetic field transverse to the axis of the main axial confining coils so as to improve the effectiveness of the confining field by counteracting the tendency of the more central lines of force in the stellarator tube to exchange positions with the magnetic lines of force nearer the walls of the tube. (AEC)

  9. MicroRNA-499 expression distinctively correlates to target genes sox6 and rod1 profiles to resolve the skeletal muscle phenotype in Nile tilapia.

    PubMed

    Nachtigall, Pedro G; Dias, Marcos C; Carvalho, Robson F; Martins, Cesar; Pinhal, Danillo

    2015-01-01

    A class of small non-coding RNAs, the microRNAs (miRNAs), has been shown to be essential for the regulation of specific cell pathways, including skeletal muscle development, maintenance and homeostasis in vertebrates. However, the relative contribution of miRNAs for determining the red and white muscle cell phenotypes is far from being fully comprehended. To better characterize the role of miRNA in skeletal muscle cell biology, we investigated muscle-specific miRNA (myomiR) signatures in Nile tilapia fish. Quantitative (RT-qPCR) and spatial (FISH) expression analyses revealed a highly differential expression (forty-four-fold) of miR-499 in red skeletal muscle compared to white skeletal muscle, whereas the remaining known myomiRs were equally expressed in both muscle cell types. Detailed examination of the miR-499 targets through bioinformatics led us to the sox6 and rod1 genes, which had low expression in red muscle cells according to RT-qPCR, FISH, and protein immunofluorescence profiling experiments. Interestingly, we verified that the high expression of miR-499 perfectly correlates with a low expression of sox6 and rod1 target genes, as verified by a distinctive predominance of mRNA destabilization and protein translational decay to these genes, respectively. Through a genome-wide comparative analysis of SOX6 and ROD1 protein domains and through an in silico gene regulatory network, we also demonstrate that both proteins are essentially similar in vertebrate genomes, suggesting their gene regulatory network may also be widely conserved. Overall, our data shed light on the potential regulation of targets by miR-499 associated with the slow-twitch muscle fiber type phenotype. Additionally the results provide novel insights into the evolutionary dynamics of miRNA and target genes enrolled in a putative constrained molecular pathway in the skeletal muscle cells of vertebrates.

  10. MicroRNA-499 Expression Distinctively Correlates to Target Genes sox6 and rod1 Profiles to Resolve the Skeletal Muscle Phenotype in Nile Tilapia

    PubMed Central

    Carvalho, Robson F.; Martins, Cesar; Pinhal, Danillo

    2015-01-01

    A class of small non-coding RNAs, the microRNAs (miRNAs), has been shown to be essential for the regulation of specific cell pathways, including skeletal muscle development, maintenance and homeostasis in vertebrates. However, the relative contribution of miRNAs for determining the red and white muscle cell phenotypes is far from being fully comprehended. To better characterize the role of miRNA in skeletal muscle cell biology, we investigated muscle-specific miRNA (myomiR) signatures in Nile tilapia fish. Quantitative (RT-qPCR) and spatial (FISH) expression analyses revealed a highly differential expression (forty-four-fold) of miR-499 in red skeletal muscle compared to white skeletal muscle, whereas the remaining known myomiRs were equally expressed in both muscle cell types. Detailed examination of the miR-499 targets through bioinformatics led us to the sox6 and rod1 genes, which had low expression in red muscle cells according to RT-qPCR, FISH, and protein immunofluorescence profiling experiments. Interestingly, we verified that the high expression of miR-499 perfectly correlates with a low expression of sox6 and rod1 target genes, as verified by a distinctive predominance of mRNA destabilization and protein translational decay to these genes, respectively. Through a genome-wide comparative analysis of SOX6 and ROD1 protein domains and through an in silico gene regulatory network, we also demonstrate that both proteins are essentially similar in vertebrate genomes, suggesting their gene regulatory network may also be widely conserved. Overall, our data shed light on the potential regulation of targets by miR-499 associated with the slow-twitch muscle fiber type phenotype. Additionally the results provide novel insights into the evolutionary dynamics of miRNA and target genes enrolled in a putative constrained molecular pathway in the skeletal muscle cells of vertebrates. PMID:25793727

  11. Flux trap void effects in the high-flux isotope reactor with curium targets

    SciTech Connect

    Rothrock, R.B.

    1990-06-01

    The reactivity effect of voids in the high-flux isotope reactor (HFIR) flux trap/target region has been reevaluated to determine the impact of the use of curium target material in lieu of the {sup 242}Pu targets originally used. The HFIR, located at the Oak Ridge National Laboratory (ORNL) and operated for the US Department of Energy by Martin Marietta Energy Systems, Inc., produces transplutonium elements for research and industrial purposes by irradiating target materials in a central flux trap region of intense thermal neutron flux. The target region void worth was extensively investigated during the HFIR nuclear design and critical experiments. Subsequently, as larger quantities of heavy elements became available, HFIR targets were fabricated from curium. This change substantially increased the amount of transplutonium isotopes present in the target during irradiation; it is therefore necessary to determine the impact on the target region void worth. Nuclear calculations of the target void worth were made using the two-dimensional transport code DORT. To benchmark the methods and data, flux trap void worth experiments performed during the HFIR start-up physics tests were calculated with the two-dimensional model for void worth measurements with and without the simulated target present in the flux trap.

  12. REFLECTOR FOR NEUTRONIC REACTORS

    DOEpatents

    Fraas, A.P.

    1963-08-01

    A reflector for nuclear reactors that comprises an assembly of closely packed graphite rods disposed with their major axes substantially perpendicular to the interface between the reactor core and the reflector is described. Each graphite rod is round in transverse cross section at (at least) its interface end and is provided, at that end, with a coaxial, inwardly tapering hole. (AEC)

  13. Rodding Surgery

    MedlinePlus

    ... operation; anesthesia issues,  Reason for the choice of rod,  Time in the hospital,  Length of recovery time at home,  Pain management including control of muscle spasms,  The rehabilitation plan. It is ...

  14. 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)

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

    DOE PAGES

    Byerly, Benjamin; Tandon, Lav; Hayes-Sterbenz, Anna; ...

    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.

  16. 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.

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

    DOEpatents

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

    1983-06-10

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

  18. 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.

  19. Isotope production target irradiation experience at the annular core research reactor

    SciTech Connect

    Talley, D.G.

    1997-02-01

    As a result of an Environmental Impact Statement (EIS) recently issued by the Department of Energy, Sandia National Laboratories (SNL) has been selected as the {open_quotes}most appropriate facility{close_quotes} for the production of {sup 99}Mo. The daughter product of {sup 99}Mo is {sup 99m}Tc, a radioisotope used in 36,000 medical procedures per day in the U.S.{close_quote} At SNL, the {sup 99}Mo would be created by the fission process in UO{sub 2} coated {open_quotes}targets{close_quotes} and chemically separated in the SNL Hot Cell Facility (HCF). SNL has recently completed the irradiation of five production targets at its Annular Core Research Reactor (ACRR). Following irradiation, four of the targets were chemically processed in the HCF using the Cintichem process.

  20. 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.

  1. 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.

  2. 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.

  3. 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.

  4. Fast reactor core concepts to improve transmutation efficiency

    NASA Astrophysics Data System (ADS)

    Fujimura, Koji; Kawashima, Katsuyuki; Itooka, Satoshi

    2015-12-01

    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.

  5. 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.

  6. 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.

  7. 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.

  8. 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.

  9. 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.

  10. A computational study of segmented tungsten rod penetration into a thick steel target plate at high velocities

    NASA Astrophysics Data System (ADS)

    Presnell, Michael B.; Rajendran, A. M.

    2012-03-01

    This paper presents results from numerical simulations of a configuration in which a tungsten heavy alloy SRP penetrates a thick RHA 4340 steel at 2.6 km/s using the 2006 version of the Lagrangian finite element code EPIC. Penetration experimental data show improved penetration efficiency by the segmented projectiles when compared to monolithic (single solid rod) projectiles. For SRP with an aspect ratio (L/D) = 1/8, a loss in penetration efficiency was seen upon successive segment impacts. The projectile configuration considered in this study was collinear impacts of eight successive discs which measured 2mm in thickness and 16mm in diameter. The EPIC simulations considered a range of parameters that influenced the Depth of Penetration (DOP) including element-particle conversion, spacing and number of segments, failure criteria, impact velocity, and mesh resolution. The EPIC results are also compared with open-literature DOP data from simulations using an Eulerian finite element code, AUTODYN for a similar configuration. In addition, the effects of back-flowing ejecta generated by the impact of first segment on the penetration processes of subsequent segments were studied in details. An alternate SRP design is proposed in this paper to alleviate the ejecta problem.

  11. Dominant Cone-Rod Dystrophy: A Mouse Model Generated by Gene Targeting of the GCAP1/Guca1a Gene

    PubMed Central

    Buch, Prateek K.; Mihelec, Marija; Cottrill, Phillippa; Wilkie, Susan E.; Pearson, Rachael A.; Duran, Yanai; West, Emma L.; Michaelides, Michel; Ali, Robin R.; Hunt, David M.

    2011-01-01

    Cone dystrophy 3 (COD3) is a severe dominantly inherited retinal degeneration caused by missense mutations in GUCA1A, the gene encoding Guanylate Cyclase Activating Protein 1 (GCAP1). The role of GCAP1 in controlling cyclic nucleotide levels in photoreceptors has largely been elucidated using knock-out mice, but the disease pathology in these mice cannot be extrapolated directly to COD3 as this involves altered, rather than loss of, GCAP1 function. Therefore, in order to evaluate the pathology of this dominant disorder, we have introduced a point mutation into the murine Guca1a gene that causes an E155G amino acid substitution; this is one of the disease-causing mutations found in COD3 patients. Disease progression in this novel mouse model of cone dystrophy was determined by a variety of techniques including electroretinography (ERG), retinal histology, immunohistochemistry and measurement of cGMP levels. It was established that although retinal development was normal up to 3 months of age, there was a subsequent progressive decline in retinal function, with a far greater alteration in cone than rod responses, associated with a corresponding loss of photoreceptors. In addition, we have demonstrated that accumulation of cyclic GMP precedes the observed retinal degeneration and is likely to contribute to the disease mechanism. Importantly, this knock-in mutant mouse has many features in common with the human disease, thereby making it an excellent model to further probe disease pathogenesis and investigate therapeutic interventions. PMID:21464903

  12. Rod Microglia: A Morphological Definition

    PubMed Central

    Taylor, Samuel E.; Morganti-Kossmann, Cristina

    2014-01-01

    Brain microglial morphology relates to function, with ramified microglia surveying the micro-environment and amoeboid microglia engulfing debris. One subgroup of microglia, rod microglia, have been observed in a number of pathological conditions, however neither a function nor specific morphology has been defined. Historically, rod microglia have been described intermittently as cells with a sausage-shaped soma and long, thin processes, which align adjacent to neurons. More recently, our group has described rod microglia aligning end-to-end with one another to form trains adjacent to neuronal processes. Confusion in the literature regarding rod microglia arises from some reports referring to the sausage-shaped cell body, while ignoring the spatial distribution of processes. Here, we systematically define the morphological characteristics of rod microglia that form after diffuse brain injury in the rat, which differ morphologically from the spurious rod microglia found in uninjured sham. Rod microglia in the diffuse-injured rat brain show a ratio of 1.79±0.03 cell length∶cell width at day 1 post-injury, which increases to 3.35±0.05 at day 7, compared to sham (1.17±0.02). The soma length∶width differs only at day 7 post-injury (2.92±0.07 length∶width), compared to sham (2.49±0.05). Further analysis indicated that rod microglia may not elongate in cell length but rather narrow in cell width, and retract planar (side) processes. These morphological characteristics serve as a tool for distinguishing rod microglia from other morphologies. The function of rod microglia remains enigmatic; based on morphology we propose origins and functions for rod microglia after acute neurological insult, which may provide biomarkers or therapeutic targets. PMID:24830807

  13. 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)

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

    SciTech Connect

    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.

  15. 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

  16. 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.

  17. 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.

  18. 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)

  19. FLASH simulations of 120MJ target explosions in LIFE reactor chamber

    NASA Astrophysics Data System (ADS)

    Sacks, Ryan; Moses, Gregory; Fatenejad, Milad

    2012-10-01

    The LIFE conceptual reactor designfootnotetextMoses, E.I., Ignition on the National Ignition Facility: a path towards inertial fusion energy, Nucl. Fusion 49 104022 is a 12 m diameter reaction chamber with a steel first wall. The chamber is filled with 6 μg/cm^3 Xenon gas to protect the wall from fusion burn products. Indirect drive 120 MJ fusion targets are shot at 13 Hz repetition rate. For purposes of simulating the target explosion the target is approximated as a 1 g lead hohlraum. Fusion burn product energy is added to the Pb in a 100 ps flattop source at a rate of 12 MJ / 100 ps. The additional 13 MJ of fusion energy is assumed to be radiated as prompt x-rays. The resulting spherical micro-explosion of the heated Pb target into the surrounding Xe is simulated in 2D using the FLASH radiation hydrodynamics code. The FLASH codefootnotetextFryxell, B., Olson, K. et al.,FLASH: An Adaptive Mesh Hydrodynamics Code for Modeling Astrophysical Thermonuclear Flashes, Astro. Journal Sup. Series., 131, 273 is an AMR block-structured, parallel scalable radiation hydrodynamics code. FLASH has separate electron and ion temperatures and single group or multi-group radiation diffusion. Shock generation in the Xe and mixing of the Pb and Xe behind the shock due to Rayleigh-Taylor instability is investigated. Comparison with results from the 1D BUCKY radiation hydrodynamics code will be presented. This work was supported by Lawrence Livermore National Laboratory under contract number B587835.

  20. Possibility of using tube shell tables for calculating the heat-transfer crisis in rod bundles of water-cooled reactors

    SciTech Connect

    Zhukov, Y.M.

    1995-02-01

    Empirical correlations are used to calculate the conditions for the appearance of a crisis of heat transfer in the core of a water-cooled reactor under normal operating conditions or during accidents. Experimental data obtained on models which simulate, to one degree of another, the geometry, the operating parameters, and the form of the radial energy-release field in real fuel assemblies are used to develop and verify such recommendations.

  1. 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.

  2. 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)

  3. About the possibility of use of different types of targets as a neutron source for subcritical nuclear reactor driven by particle beam accelerator

    SciTech Connect

    Avdeev, E.F.; Dorokhovich, S.L.; Chusov, I.A.

    1995-10-01

    The schemes of jet gas and liquid targets as well as the gastargets with a solid phase dispersion are introduced to use to receive the neutrons admitted to a subcritical reactor core. The possible variants of target position in the reactor are considered, target characteristics are calculated. The authors pay a great attention to the estimation of radioactive products yield receiving due to the interaction of the beam with the target.

  4. 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)

  5. 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.

  6. Improving fuel-rod performance. [PWR; BWR

    SciTech Connect

    Ocken, H.; Knott, S.

    1981-03-01

    To reduce the risk of fuel-rod failures, utilities operate their nuclear reactors within conservative limits on power increases proposed by nuclear-fuel vendors. Of particular concern to US utilities is that adopting these limits results in an industrywide average plant capacity loss of 3% in BWR designs and 0.3% in PWR designs. To replace lost BWR capacity by other generating means currently costs the utilities $150 million annually, and losses for PWRs are about $20 million. Efforts are therefore being made to identify the factors responsible for Zircaloy degradation under PCI condition and to improve nuclear-fuel-rod design and reactor operation.

  7. 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.

  8. 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.

  9. 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.

  10. 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.

  11. Tritium Producing Burnable Absorber Rod and Its Inherent Safety Features

    SciTech Connect

    Reid, Bruce D.; Lanning, Donald D.

    2000-06-24

    Description of the Tritium-producing burnable absorber rods (TPBARs) used in the U.S. Department of Energy's Commercial Light Water Reactor (CLWR) Project and a discussion of their inherent safety features.

  12. 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.

  13. 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.

  14. DRX-Induced Solid-State Flow and Projectile-Target Mixing During [001] Single-Crystal Tungsten Rod Penetration into Steel Targets

    NASA Astrophysics Data System (ADS)

    Pizana, Carlos; Tamoria, T. L.; Cytron, Sheldon

    2005-07-01

    Residual [001] single-crystal W penetrators have been examined by light and electron microscopy. The post-impact residual penetrators examined using energy-dispersive x-ray mapping, revealed target and penetrator mechanical mixing. Considerable intercalation activity was found to concentrate specifically within the material being eroded by DRX-assisted flow. The solid-state flow features including shear bands facilitate the mixing of the two. Peripherally along the head of the penetrator and adjacent to the shear band itself, large bands of high Ni steel appear to influence the solid-state flow of the penetrator. Residual microstructures obtained within the penetrator suggest localized melt zones due to thermal instabilities caused by the turbulent behavior in the high-pressure regime. Supported by the U.S. Army TACOM-Picatinny Arsenal.

  15. 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.

  16. 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.

  17. Theoretical assessment and targeted modeling of TiO2 in reactor towards the scandium radioisotopes estimation.

    PubMed

    Hosseini, Seyedeh Fatemeh; Sadeghi, Mahdi; Aboudzadeh, Mohammad Reza

    2017-09-01

    (47)Sc radioisotope with beta particle emission can be used for targeted radionuclide therapy in nuclear medicine and can be produced by nuclear reactor (with high activities) and accelerator. In this work, the specific activity of (47)Sc with the isotopic impurities produced through irradiating the enriched (47)Tio2 and natural TiO2 targets have been calculated by the MCNPX-2.6 and the TALYS-1.8 codes and also theoretical approach in a fast neutron flux 3×10(13)ncm(-2)s(-1) in the reactor. In addition, the cross sections of (46)Ti(n,p)(46)Sc reaction have been interpolated in the 1-10MeV energy range and compared with the corresponding experimental data, as well. Moreover, the average cross sections of (46)Ti(n,p)(46)Sc, (47)Ti(n,p)(47)Sc and (48)Ti(n,p)(48)Sc reactions were reproduced. Acceptable agreement between the experimental data and calculated results confirms the ability of the used methods to design and predict the production process in reactor. Copyright © 2017 Elsevier Ltd. All rights reserved.

  18. 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.

  19. 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.

  20. 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.

  1. 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.

  2. 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

  3. 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.

  4. 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.

  5. 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.

  6. 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.

  7. Two-sided conical laser target for a neutron source of a hybrid nuclear-thermonuclear reactor

    NASA Astrophysics Data System (ADS)

    Lebo, I. G.; Isaev, E. A.; Lebo, A. I.

    2017-03-01

    Numerical simulations suggest that a source of thermonuclear neutrons with a high pulse repetition rate and the number of neutrons of ~1017 per pulse, which is required for the development of nuclear-thermonuclear reactors, can be realised in the irradiation of a two-sided conical target simultaneously by a long and short laser pulses with energies of ~1 MJ and 50 kJ and durations of 100 – 250 ns and 0.1 – 1 ns. We consider the feasibility of verifying separate propositions of the proposed conception with the use of existing laser facilities.

  8. Report on the evaluation of the tritium producing burnable absorber rod lead test assembly. Revision 1

    SciTech Connect

    1997-03-01

    This report describes the design and fabrication requirements for a tritium-producing burnable absorber rod lead test assembly and evaluates the safety issues associated with tritium-producing burnable absorber rod irradiation on the operation of a commercial light water reactor. The report provides an evaluation of the tritium-producing burnable absorber rod design and concludes that irradiation can be performed within U.S. Nuclear Regulatory Commission regulations applicable to a commercial pressurized light water reactor.

  9. 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.

  10. 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)

  11. 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.

  12. 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.

  13. 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.

  14. 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.

  15. Neutronic and thermal-hydraulic analysis of fission molybdenum-99 production at Tehran Research Reactor using LEU plate targets.

    PubMed

    Abedi, Ebrahim; Ebrahimkhani, Marzieh; Davari, Amin; Mirvakili, Seyed Mohammad; Tabasi, Mohsen; Maragheh, Mohammad Ghannadi

    2016-12-01

    Efficient and safe production of molybdenum-99 ((99)Mo) radiopharmaceutical at Tehran Research Reactor (TRR) via fission of LEU targets is studied. Neutronic calculations are performed to evaluate produced (99)Mo activity, core neutronic safety parameters and also the power deposition values in target plates during a 7 days irradiation interval. Thermal-hydraulic analysis has been also carried out to obtain thermal behavior of these plates. Using Thermal-hydraulic analysis, it can be concluded that the safety parameters are satisfied in the current study. Consequently, the present neutronic and thermal-hydraulic calculations show efficient (99)Mo production is accessible at significant activity values in TRR current core configuration. Copyright © 2016 Elsevier Ltd. All rights reserved.

  16. 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.

  17. 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

  18. Directions for reactor target design based on the US heavy ion fusion systems assessment

    SciTech Connect

    Wilson, D.C.; Dudziak, D.; Magelssen, G.; Zuckerman, D.; Dreimeyer, D.

    1986-01-01

    We studied areas of major uncertainty in target design using the cost of electricity as our figure of merit. Net electric power from the plant was fixed at 1000 MW to eliminate large effects due to economies of scale. The system is relatively insensitive to target gain. Factors of three changes in gain cause only 8 to 12% changes in electricity cost. An increase in the peak power needed to drive targets poses only a small cost risk, but requires many more beamlets be transported to the target. A shortening of the required ion range causes both cost and beamlet difficulties. A factor of 4 decrease in the required range at a fixed driver energy increases electricity cost by 44% and raises the number of beamlets to 240. Finally, the heavy ion fusion system can accommodate large increases in target costs. To address the major uncertainties, target design should concentrate on the understanding requirements for ion range and peak driver power.

  19. 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.

  20. Estimation of the activity generated by neutron activation in control rods of a BWR.

    PubMed

    Ródenas, José; Gallardo, Sergio; Abarca, Agustín; Juan, Violeta

    2010-01-01

    Control rods are activated by neutron reactions into the reactor. The activation is produced mainly in stainless steel and its impurities. The dose produced by this activity is not important inside the reactor, but it has to be taken into account when the rod is withdrawn from the reactor. Activation reactions produced have been modelled by the MCNP5 code based on the Monte Carlo method. The code gives the number of reactions that can be converted into activity.

  1. 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.

  2. 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.

  3. Thermal modeling of W rod armor.

    SciTech Connect

    Nygren, Richard Einar

    2004-09-01

    Sandia has developed and tested mockups armored with W rods over the last decade and pioneered the initial development of W rod armor for International Thermonuclear Experimental Reactor (ITER) in the 1990's. We have also developed 2D and 3D thermal and stress models of W rod-armored plasma facing components (PFCs) and test mockups and are applying the models to both short pulses, i.e. edge localized modes (ELMs), and thermal performance in steady state for applications in C-MOD, DiMES testing and ITER. This paper briefly describes the 2D and 3D models and their applications with emphasis on modeling for an ongoing test program that simulates repeated heat loads from ITER ELMs.

  4. 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. © 2015 by The American Society for Biochemistry and Molecular Biology, Inc.

  5. 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

  6. IMPACT OF ENERGY GROUP STRUCTURE ON NUCLEAR DATA TARGET ACCURACY REQUIREMENTS FOR ADVANCED REACTOR SYSTEMS

    SciTech Connect

    G. Palmiotti; M. Salvatores; H. Hiruta

    2011-06-01

    A target accuracy assessment study using both a fine and a broad energy structure has shown that less stringent nuclear data accuracy requirements are needed for the latter energy structure. However, even though a reduction is observed, still the requirements will be very difficult to be met unless integral experiments are also used to reduce nuclear data uncertainties. Target accuracy assessment is the inverse problem of the uncertainty evaluation. To establish priorities and target accuracies on data uncertainty reduction, a formal approach can be adopted by defining target accuracy on design parameters and finding out required accuracy on data in order to meet them. In fact, the unknown uncertainty data requirements can be obtained by solving a minimization problem where the sensitivity coefficients in conjunction with the constraints on the integral parameters provide the needed quantities for finding the solutions.

  7. 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.

  8. 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.

  9. Fuel-rod temperature transients during LWR degraded-core accidents

    SciTech Connect

    Briscoe, F; Rivard, J B; Young, M F

    1982-01-01

    Heat transfer models of fuel rods and coolant have been developed in support of LWR damaged fuel studies underway at Sandia National Laboratories for the NRC. A one-dimensional, full-length model simulates a PWR fuel rod; a two-dimensional, 0.5 m model simulates 9-rod bundle experiments to be performed in the Annular Core Research Reactor. The models include zircaloy oxidation, heat transfer by convecting steam/hydrogen flow, and radiation between surfaces through an absorbing/emitting gas. Characteristics of the one-dimensional reactor fuel rod model for two types of accident sequence are reported, as well as comparisons with MARCH code results.

  10. Final design review of boron carbide safety rod

    SciTech Connect

    Lutz, R.N.

    1991-09-24

    The object of this paper discusses the design review of the boron carbide safety rod for the Westinghouse Savannah River Company. This paper reviewed information presented by personnel of the Savannah River Laboratory (SRL) Equipment Engineering Section, SRL Materials Technology Section and Reactor Materials Engineering and Technology. From this report, views, opinions and recommendations were made on the safety rod from materials testing to production.

  11. Final design review of boron carbide safety rod

    SciTech Connect

    Lutz, R.N.

    1991-09-24

    The object of this paper discusses the design review of the boron carbide safety rod for the Westinghouse Savannah River Company. This paper reviewed information presented by personnel of the Savannah River Laboratory (SRL) Equipment Engineering Section, SRL Materials Technology Section and Reactor Materials Engineering and Technology. From this report, views, opinions and recommendations were made on the safety rod from materials testing to production.

  12. 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.

  13. 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.

  14. Scaled Long Rod Penetration Experiments: Tungsten - RHA

    NASA Astrophysics Data System (ADS)

    Proud, William

    2011-06-01

    Scaled, reverse ballistic, long-rod experiments were performed at an impact velocity in the range of ~700 m s-1. The targets were tungsten alloy rods and the projectiles either 3 or 6 mm thick rolled homogeneous armour (RHA) plates. The plate was inclined at 30 to the direction of travel and the interaction was recorded using high-speed photography, strain gauges and laser velocimetry. The pitch of the rod was varied in steps of 3 over a total range of 15. In this range the rod deformation changed dramatically the bending process moved from a flexing of the tip away from the plate, to a marked motion into the surface. Cross comparison of the diagnostic outputs reveals the time windows for these process and also the varying sensitivity of the measurement system to that process. Post-impact recovery was also performed.

  15. Substitute safety rods: Physics design and NTG calibration

    SciTech Connect

    Baumann, N.P.

    1991-07-01

    Under certain assumed accident conditions, an SRS reactor may loose most of its bulk moderator while maintaining flow to fuel assemblies. If this occurs immediately after operation at power, components normally dependent on convective heat transfer to the moderator will heat up with the possibility of melting that component. One component at risk is the safety rod. Tests have shown that the current cadmium safety rod, which contains aluminum as well as cadmium, can fail at temperatures only slightly in excess of 500 deg C. Computations indicate that such temperatures can be reached with operating powers well below the 50% power limit now imposed by other accident scenarios. Safety rod melting would thus establish a new lower operating limit. A substitute safety rod that could tolerate much higher temperatures would eliminate this limit. This memorandum details the physics characteristics of a suitable replacement rod. 7 refs.

  16. 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.

  17. 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.

  18. 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.

  19. 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.

  20. 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.

  1. Irradiation damage analysis on the flat plate type target plate of the divertor for fusion experimental reactors

    NASA Astrophysics Data System (ADS)

    Ishiyama, S.; Akiba, M.; Eto, M.

    1996-04-01

    To design the relevant plasma facing components of fusion experimental reactors such as ITER, irradiation damage analysis, especially on divertor structures exposed to high heat flux and heavy neutron irradiation, is one of the most important problems. This paper presents finite element analytical results of the thermal and irradiation induced stresses which occurred in the divertor structures which are exposed to neutron irradiation at 0-1 dpa with a high heat flux up to 15 MW/m 2. A type of target plate model of the divertor structure studied in present study e.g. flat plate model has bonded structure of one-dimensional high thermal conductivity carbon-carbon composite (C/C) and oxygen-free high conductivity copper (OFHC), as armor and substrate/heat sink materials, respectively. These results show that irradiation induced stresses at edges of bonded interface between an armor and a substrate/heat sink, become higher with increase of dpa and reach up to the critical values of the materials at 0 and 1 dpa. This indicates that drop-off of armor tiles from substrate structure is one of very serious problems for the safety design of target plate; thus the reduction of service conditions and change of divertor materials are important to extend lifetime of the model.

  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. 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. Rat Hippocampal Neurons Express Genes for Both Rod Retinal and Olfactory Cyclic Nucleotide-Gated Channels: Novel Targets for cAMP/cGMP Function

    NASA Astrophysics Data System (ADS)

    Kingston, Paul A.; Zufall, Frank; Barnstable, Colin J.

    1996-09-01

    Cyclic nucleotide-gated (CNG) channels are Ca2+-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.

  5. Trunnion Rod Microcrack Detection

    DTIC Science & Technology

    2013-08-01

    Engineers needs reliable non-destructive testing (NDT) methods that are rapid, robust, and capable of detecting and quantifying defects , as well as...Robustness is required to handle the significant variations in design, construction, and field conditions that are known to exist. Defect detection and...constructed to help contain failing rods. It is not known if the defect growth and resulting rod failure rate will stay at its relatively low number of

  6. Evaluations for the choice of the best target material in a accelerator-driven reactor

    SciTech Connect

    Neuhold, P.

    1995-10-01

    From the point of view of the overall credibility of the machine the choice of the best target material play a fundamental role. That has to fulfill at least three requirements: (1) the maximum number of neutrons produced per incident proton; (2) the maximum melting point; (3) the minimum activation under neutron irradiation. A number of different candidates can be investigated, keeping in mind that the first requirement is roughly fulfilled by high atomic number elements: the most popular among them appear to be W, Pb, and Np. While the characteristics for the second point are well known (3410C, 327C and 640C), as well as many evaluations are done about the first point, even if showing numbers ranging roughly a factor of 2 between the lowest and the highest ones, poor attention instead has been given to the third one. As far as the operation/maintenance of the machine is concerned, is important to know the dose rate of the target due to its neutron activation. According to the author`s calculations, for example after one year of 800 MeV/62.5 mA proton irradiation of W, one can have, after one day of decay, dose rates as high as 2*10**6 Sv/hr, cumulative of the effects due to the direct proton beam and to the arising neutron flux. The author made use of the codes MSMPN, ANITA for the neutron activation analysis and of NMTC, ORIGEN for the proton interaction analysis. In particular it`s worth noting how dramatic differences can arise from different isotopic composition, that is for example between natural W or only W-186, due to the very different behavior of the daughter elements as decay time and gamma-energy. For example the dose rate due to the neutron activation alone, after one month of decay, for natural W is a factor of 500 greater than for W-186.

  7. 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.

  8. 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.

  9. 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.

  10. 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.

  11. Long-Rod Moving-Plate Interaction

    NASA Astrophysics Data System (ADS)

    Partom, Yehuda

    2001-06-01

    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 AUTODYN2D/Euler in planar symmetry, although the problem is 3D. We assume that this is a good enough approximation, as we're only interested in the main features, and are not comparing fine detailes 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 dL is diverted at an angle and becomes ineffective for penetrating the main target. We made many such runs to determine the dependence of dL on the material and kinematic parameters of the problem. This dependence makes it possible to predict dL for a variaty of rod-plate situations.

  12. 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.

  13. 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.

  14. Investigation of water-logged spent fuel rods under dry storage conditions

    SciTech Connect

    Kohli, R.; Pasupathi, V.

    1986-09-01

    Tests were conducted to determine the amount of moisture contained in breached, water-logged spent fuel rods and the rate of release. Two well-characterized BWR fuel rods with reactor-induced breaches were tested in a hot cell. These rods contained approximately 6 to 10 g of moisture, most of which was released during heating tests simulating normal cask drying operations. Additional testing with two intentionally defected fuel rods (BWR and PWR) was performed to evaluate the effect of the cladding breach on migration of moisture along the length of the fuel rod. The results showed that the moisture released from reactor-breached spent fuel rods was insufficient to cause degradation of fuel or dry storage system components.

  15. Investigation of Swirling Flow in Rod Bundle Subchannels Using Computational Fluid Dynamics

    SciTech Connect

    Holloway, Mary V.; Beasley, Donald E.; Conner, Michael E.

    2006-07-01

    The fluid dynamics for turbulent flow through rod bundles representative of those used in pressurized water reactors is examined using computational fluid dynamics (CFD). The rod bundles of the pressurized water reactor examined in this study consist of a square array of parallel rods that are held on a constant pitch by support grids spaced axially along the rod bundle. Split-vane pair support grids are often used to create swirling flow in the rod bundle in an effort to improve the heat transfer characteristics for the rod bundle during both normal operating conditions and in accident condition scenarios. Computational fluid dynamics simulations for a two subchannel portion of the rod bundle were used to model the flow downstream of a split-vane pair support grid. A high quality computational mesh was used to investigate the choice of turbulence model appropriate for the complex swirling flow in the rod bundle subchannels. Results document a central swirling flow structure in each of the subchannels downstream of the split-vane pairs. Strong lateral flows along the surface of the rods, as well as impingement regions of lateral flow on the rods are documented. In addition, regions of lateral flow separation and low axial velocity are documented next to the rods. Results of the CFD are compared to experimental particle image velocimetry (PIV) measurements documenting the lateral flow structures downstream of the split-vane pairs. Good agreement is found between the computational simulation and experimental measurements for locations close to the support grid. (authors)

  16. 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)

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

    NASA Astrophysics Data System (ADS)

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

    1991-02-01

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

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

    SciTech Connect

    Guenther, R.J.

    1983-01-01

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

  19. Rod Contributions to Color Perception: Linear with Rod Contrast

    PubMed Central

    Cao, Dingcai; Pokorny, Joel; Smith, Vivianne C.; Zele, Andrew J.

    2008-01-01

    At mesopic light levels, an incremental change in rod activation causes changes in color appearance. In this study, we investigated how rod mediated changes in color perception varied as a function of the magnitude of the rod contrast. Rod-mediated changes in color appearance were assessed by matching them with cone-mediated color changes. A two-channel four-primary colorimeter allowed independent control of the rods and each of the L-, M- and S-cone photoreceptor types. At all light levels, rod contributions to inferred PC, KC and MC pathway mediated vision were linearly related to the rod incremental contrast. This linear relationship could be described by a model based on primate ganglion cell responses with the assumption that rod signals were conveyed via rod-cone gap junctions at mesopic light levels. PMID:18561973

  20. 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.

  1. Dysprosium titanate as an absorber material for control rods

    NASA Astrophysics Data System (ADS)

    Risovany, V. D.; Varlashova, E. E.; Suslov, D. N.

    2000-09-01

    Disprosium titanate is an attractive control rod material for the thermal neutron reactors. Its main advantages are: insignificant swelling, no out-gassing under neutron irradiation, rather high neutron efficiency, a high melting point (˜1870°C), non-interaction with the cladding at temperatures above 1000°C, simple fabrication and easily reprocessed non-radioactive waste. It can be used in control rods as pellets and powder. The disprosium titanate control rods have worked off in the MIR reactor for 17 years, in VVER-1000 - for 4 years without any operating problems. After post-irradiation examinations this type of control rod having high lifetime was recommended for the VVER and RBMK. The paper presents the examination results of absorber element dummies containing dysprosium titanate, irradiated in the SM reactor to the neutron fluence of 3.4×10 22 cm -2 ( E>0.1 MeV) and, also, the data on structure, thermal-physical properties of dysprosium titanate, efficiency of dysprosium titanate control rods.

  2. 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.

  3. Tritium analysis of irradiated burnable poison rods

    SciTech Connect

    D'Annucci, F.; Baro, G.; Pardue, E.B.S.; Rommelaere, W.

    1982-10-01

    To investigate the tritium content in the various components and to determine the /sup 10/B burnup, a postirradiation examination was carried out on three burnable poison rods that had been irradiated in the first cycle of the Oconee 2 Reactor. The results of the analysis reveal that the Al/sub 2/O/sub 3/-B/sub 4/C pellets retain the major portion (greater than or equal to99%) of all the tritium generated; only a very small quantity (less than or equal to0.5%) of the tritium produced is absorbed by the cladding and no tritium was detected in the plenum gas. Comparison of the average postirradiation /sup 10/B content with the preirradiation content indicates that almost all of the /sup 10/B has been consumed. The experimental results are in good agreement with the calculated tritium content of an irradiated poison rod.

  4. Failure of latch mechanism for motion control of safety rods

    SciTech Connect

    Yau, W.W.F.; Leader, D.R.

    1992-01-16

    During safety rod tests in K-reactor prior to startup, one safety rod could not be lifted because the ``button`` broke off and became lodged in the mechanism. Examination of the failed latch assembly along with other assemblies from both K-Area and L-Area revealed several missing buttons as well as severely deformed ``jaw hanger extensions.`` We participated in the investigation of the damage by request of the Reactor Restart Section. Based on our study of the latch mechanism, the modifications to the ``safety rod extension,`` and the operating history of the machine, this memorandum describes the causes of the observed damage with experimental evidence and calculations to support the findings. 3 refs.

  5. Failure of latch mechanism for motion control of safety rods

    SciTech Connect

    Yau, W.W.F.; Leader, D.R.

    1992-01-16

    During safety rod tests in K-reactor prior to startup, one safety rod could not be lifted because the button'' broke off and became lodged in the mechanism. Examination of the failed latch assembly along with other assemblies from both K-Area and L-Area revealed several missing buttons as well as severely deformed jaw hanger extensions.'' We participated in the investigation of the damage by request of the Reactor Restart Section. Based on our study of the latch mechanism, the modifications to the safety rod extension,'' and the operating history of the machine, this memorandum describes the causes of the observed damage with experimental evidence and calculations to support the findings. 3 refs.

  6. Intramedullary rodding in osteogenesis imperfecta.

    PubMed

    Mulpuri, K; Joseph, B

    2000-01-01

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

  7. 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.

  8. 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.

  9. 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.

  10. Carbon rod furnace infrared source

    NASA Technical Reports Server (NTRS)

    Boyd, W. J.; Jennings, D. E.; Blass, W. E.; Gailar, N. M.

    1974-01-01

    A carbon rod furnace infrared source has been built and has proven to be a reliable and trouble-free source despite the high rod temperature of 2500 K. The furnace offers several advantages over previous furnaces. These include an increase in usable rod length to 6.35 cm, extended rod life, and a fully automated power supply for the furnace. Construction and operational details are discussed.

  11. Anchor for Fiberglas Guy Rod

    NASA Technical Reports Server (NTRS)

    Wilson, A. H.

    1982-01-01

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

  12. 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.

  13. 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.

  14. Learning with Rods: One Account.

    ERIC Educational Resources Information Center

    Cherry, Donald Esha

    This paper discusses one English as a Second Language (ESL) teacher's attempts to use cuisenaire rods as a language learning tool. Cuisenaire rods (sometimes called algebricks) vary in size from 1 x 1 x 10 centimeter sticks to 1 x 1 x 1 centimeter cubes, with each of the 10 sizes a different color. Although such rods have been used to teach…

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

    SciTech Connect

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

    1981-06-01

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

  16. Gas chromatography/isotope ratio mass spectrometry of recalcitrant target compounds: performance of different combustion reactors and strategies for standardization.

    PubMed

    Reinnicke, Sandra; Juchelka, Dieter; Steinbeiss, Sibylle; Meyer, Armin; Hilkert, Andreas; Elsner, Martin

    2012-05-15

    Compound-specific isotope analysis (CSIA) relies on continuous flow combustion of organic substances to CO(2) and N(2) in a miniature reactor to measure (13)C/(12)C and (15)N/(14) N stable isotope ratios. Accurate analysis is well established for many volatile hydrocarbons. In contrast, compounds which contain hetero and halogen atoms are less volatile and may be more recalcitrant to combustion. This study tested carbon and nitrogen isotope analysis of atrazine, desethylatrazine (DEA), dichlobenil and 2,6-dichlorobenzamide (BAM) by gas chromatography/isotope ratio mass spectrometry (GC/IRMS) with multiple reactor tubes of two different kinds (conventional CuO/NiO/Pt and a NiO tube/CuO-NiO reactor prototype). The advantages of the NiO tube/CuO-NiO reactor were the absence of an additional reduction reactor, the possibility of routine reoxidation in nitrogen isotope analysis, and reliable atrazine and DEA measurements over several hundred injections. In contrast, BAM analysis showed good accuracy for carbon, but notable variations in the trueness of nitrogen isotope ratios. Accurate carbon and nitrogen analysis was nevertheless possible by bracketing samples with external compound-specific standards and subsequent offset correction. We conclude that instrument data should never be taken at its 'face value', but must consistently be validated with compound-specific standards of the respective analytes. Copyright © 2012 John Wiley & Sons, Ltd.

  17. 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.

  18. 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

  19. A subchannel analysis code for supercritical-pressure LWR with downward-flowing water rods

    SciTech Connect

    Tanabe, T.; Koshizuka, S.; Oka, Y.; Kitou, K.

    2004-07-01

    This paper describes development of a subchannel analysis code for the square fuel assembly with downward-flowing water rods, which is the new design of high temperature thermal reactor (SCLWR-H) in Univ. Tokyo. PLANDTL experiment for liquid metal cooled fast breeder reactor and ASFRE-III code analysis are used to verify the present code. Subchannel analysis for SCLWR-H, which considers the effect of water rods, is executed. Since the hydraulic diameters of the subchannels are almost constant in the square fuel assembly, the flow and temperature distributions are more uniform than those in the hexagonal assembly. However, coolant temperature slightly depends on the subchannel types. The temperature in the corner subchannels is the lowest. This is because flow area is small against the contact length with the water rod and heat transfer to the water rod is larger than that in other subchannels. The coolant temperature in the corner subchannels is 17 deg. C lower than the average temperature (500 deg. C). The temperature distributions in the water rods are also evaluated. The water rods outside the channel box have a higher temperature distribution than that in the inside water rods. The temperature in the outside water rods increases up to 378 deg. C which is close to pseudo-critical temperature. It can be reduced by increasing the flow rate in the outside water rods. (authors)

  20. 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

  1. 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.

  2. 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.

  3. Actuator system history of safety rod lower latch problems review of latch inspection video tapes

    SciTech Connect

    Banks, J.J.

    1992-06-24

    During pre-restart testing the safety rod at position X26-YlO bound after being driven approximately two (2) feet out of the reactor. Subsequently, the rod was manually returned to it's seated position. Inspection of the lower latch showed that the latch locking plunger button (screwed on to the bottom of the plunger shaft and retained by a pin through a hole drilled through the button and the plunger shaft) was missing. The shaft failed through the hole drilled for the retaining pin. The button, with the retaining pin intact, was found lodged between the safety rod upper adapter collar and the top of the safety rod thimble top fitting. Analysis of the safety rod latch and accompanying forest guide tube design provided assurance that this type of failure would not cause binding during the scramming'' of the safety rods. Inspection of all of the K'' safety rod lower latches revealed six other latches with missing plunger buttons, and nine with other non-conformances which required latch replacement. A history search conducted by Reactor Engineering Design, Components Handling Group, is included in this report. The history search shows that latch design modifications, as a part of initial development of the latch system and later to improve the delatching operation, were made from 1950 to 1960. These modifications created a condition where latch damage could occur. Video tapes were made during inspection of the safety rod latches in K area and control rod latches in L area. These tapes were reviewed by Reactor Engineering Design Components Handling engineers. The reviews were used for correlation of latch problems reported by the engineers/mechanics making the inspections. The K area tapes showed inspection of 65 of the 66 safety rod latches. The review of the tapes showed the plunger buttons to be missing from five latches. RED-CH reviewed the L Area video taped inspection of 35 control rod clusters (245 latches). No non-conformances were noted.

  4. Actuator system history of safety rod lower latch problems review of latch inspection video tapes

    SciTech Connect

    Banks, J.J.

    1992-06-24

    During pre-restart testing the safety rod at position X26-YlO bound after being driven approximately two (2) feet out of the reactor. Subsequently, the rod was manually returned to it`s seated position. Inspection of the lower latch showed that the latch locking plunger button (screwed on to the bottom of the plunger shaft and retained by a pin through a hole drilled through the button and the plunger shaft) was missing. The shaft failed through the hole drilled for the retaining pin. The button, with the retaining pin intact, was found lodged between the safety rod upper adapter collar and the top of the safety rod thimble top fitting. Analysis of the safety rod latch and accompanying forest guide tube design provided assurance that this type of failure would not cause binding during the ``scramming`` of the safety rods. Inspection of all of the ``K`` safety rod lower latches revealed six other latches with missing plunger buttons, and nine with other non-conformances which required latch replacement. A history search conducted by Reactor Engineering Design, Components Handling Group, is included in this report. The history search shows that latch design modifications, as a part of initial development of the latch system and later to improve the delatching operation, were made from 1950 to 1960. These modifications created a condition where latch damage could occur. Video tapes were made during inspection of the safety rod latches in K area and control rod latches in L area. These tapes were reviewed by Reactor Engineering Design Components Handling engineers. The reviews were used for correlation of latch problems reported by the engineers/mechanics making the inspections. The K area tapes showed inspection of 65 of the 66 safety rod latches. The review of the tapes showed the plunger buttons to be missing from five latches. RED-CH reviewed the L Area video taped inspection of 35 control rod clusters (245 latches). No non-conformances were noted.

  5. 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.

  6. 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.

  7. Electrochemically induced deuterium-tritium fusion power reactor; Preliminary design of a reactor system

    SciTech Connect

    Oka, Y.; Koshizuka, S.; Kondo, S. . Faculty of Engineering )

    1989-09-01

    Conceptual design of an electrochemically induced deuterium-tritium fusion power reactor has been carried out. A double-tube-type fuel cell is proposed for efficient electrolysis and to provide a large cathode area. The fuel cell tubes are assembled like a pressurized water reactor (PWR) control rod cluster. The tritium fuel is continuously fed through the cluster rod to the cell. The voltage for the electrolysis is supplied through the rod. The tritium breeding Li/sub 2/O is contained in a hexagonal blanket through which coolant tubes penetrate. The fuel cell tube is inserted in the coolant tube and the water coolant flows through the annuli.

  8. 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.

  9. A measurement of a control rod drop using an LVDT

    NASA Astrophysics Data System (ADS)

    Choi, Myoung-Hwan; Kim, Ji-Ho; Huh, Hyung; Yu, Je-Yong; Sohn, Dong-Seong

    2010-03-01

    A control element drive mechanism is a reactor regulating system, which is to insert, withdraw, or maintain a control rod containing neutron-absorbing material within a reactor core to control the reactivity of the reactor. The ball-screw type CEDM for the small and medium research reactor has a spring-hydraulic damper to reduce the impact force due to the free drop of the CEDM. This paper describes the experimental results to obtain the drop characteristics of the CEDM. The tests are performed by using a full-scale structure except the control element assembly, and a drop time and displacement after an impact are measured by using an LVDT. The influences of the rod weight and the drop height on the drop behavior are also estimated on the basis of test results. In case of the longest stroke, the drop time of the control rod is within 4.5 seconds to meet the design requirement. The behavior after the impact shows a general damping motion of the spring-damper system, and the maximum displacement is measured as 15.6 mm.

  10. A measurement of a control rod drop using an LVDT

    NASA Astrophysics Data System (ADS)

    Choi, Myoung-Hwan; Kim, Ji-Ho; Huh, Hyung; Yu, Je-Yong; Sohn, Dong-Seong

    2009-12-01

    A control element drive mechanism is a reactor regulating system, which is to insert, withdraw, or maintain a control rod containing neutron-absorbing material within a reactor core to control the reactivity of the reactor. The ball-screw type CEDM for the small and medium research reactor has a spring-hydraulic damper to reduce the impact force due to the free drop of the CEDM. This paper describes the experimental results to obtain the drop characteristics of the CEDM. The tests are performed by using a full-scale structure except the control element assembly, and a drop time and displacement after an impact are measured by using an LVDT. The influences of the rod weight and the drop height on the drop behavior are also estimated on the basis of test results. In case of the longest stroke, the drop time of the control rod is within 4.5 seconds to meet the design requirement. The behavior after the impact shows a general damping motion of the spring-damper system, and the maximum displacement is measured as 15.6 mm.

  11. Range shortening, radiation transport, and Rayleigh-Taylor instability phenomena in ion-beam-driven inertial-fusion-reactor-size targets: Implosion, ignition, and burn phases

    NASA Astrophysics Data System (ADS)

    Long, Keith A.; Tahir, Naeem A.

    1987-03-01

    In this paper we present an analysis of the theory of the energy deposition of ions in cold materials and hot dense plasmas together with numerical calculations for heavy and light ions of interest to ion-beam fusion. We have used the gorgon computer code of Long, Moritz, and Tahir (which is an extension of the code originally written for protons by Nardi, Peleg, and Zinamon) to carry out these calculations. The energy-deposition data calculated in this manner has been used in the design of heavy-ion-beam-driven fusion targets suitable for a reactor, by its inclusion in the medusa code of Christiansen, Ashby, and Roberts as extended by Tahir and Long. A number of other improvements have been made in this code and these are also discussed. Various aspects of the theoretical analysis of such targets are discussed including the calculation of the hydrodynamic stability, the hydrodynamic efficiency, and the gain. Various different target designs have been used, some of them new. In general these targets are driven by Bi+ ions of energy 8-12 GeV, with an input energy of 4-6.5 MJ, with output energies in the range 600-900 MJ, and with gains in the range 120-180. The peak powers are in the range of 500-750 TW. We present detailed calculations of the ablation, compression, ignition, and burn phases. By the application of a new stability analysis which includes ablation and density-gradient effects we show that these targets appear to implode in a stable manner. Thus the targets designed offer working examples suited for use in a future inertial-confinement fusion reactor.

  12. 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.

  13. ELECTRONUCLEAR REACTOR

    DOEpatents

    Lawrence, E.O.; McMillan, E.M.; Alvarez, L.W.

    1960-04-19

    An electronuclear reactor is described in which a very high-energy particle accelerator is employed with appropriate target structure to produce an artificially produced material in commercial quantities by nuclear transformations. The principal novelty resides in the combination of an accelerator with a target for converting the accelerator beam to copious quantities of low-energy neutrons for absorption in a lattice of fertile material and moderator. The fertile material of the lattice is converted by neutron absorption reactions to an artificially produced material, e.g., plutonium, where depleted uranium is utilized as the fertile material.

  14. Distributed temperature sensing inside a 19-rod bundle

    DOE PAGES

    Lomperski, S.; Bremer, N.; Gerardi, C.

    2017-05-23

    The temperature field within a model of a sodium-cooled fast reactor fuel rod bundle was measured using Ø155 μm fiber optic distributed temperature sensors (DTS). The bundle consists of 19 electrically-heated rods Ø6.3 mm and 865 mm long. Working fluids were argon and air at atmospheric pressure and Reynolds numbers up to 300. A 20 m-long DTS was threaded through Ø1 mm capillaries wound around rods as wire-wraps. The sensor generated 173 measurements along each rod at 5 mm resolution for a total of 3300 data locations. A second DTS, 58 m long, was suspended between rods to provide 9300more » fluid temperature measurements at 20 mm resolution. Such data density makes it possible to construct 3D maps of the temperature field that are beyond the reach of traditional sensors such as thermocouples. This is illustrated through a series of steady-state and transient tests. As a result, the work demonstrates the feasibility of mapping temperature within the close confines of a rod bundle at resolutions suitable for validation of computational fluid dynamics codes.« less

  15. 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.

  16. 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.

  17. Concept for LEU Burst Reactor

    SciTech Connect

    Klein, Steven Karl; Kimpland, Robert Herbert

    2016-03-07

    Design and performance of a proposed LEU burst reactor are sketched. Salient conclusions reached are the following: size would be ~1,500 kg or greater, depending on the size of the central cavity; internal stresses during burst require split rings for relief; the reactor would likely require multiple control and safety rods for fine control; the energy spectrum would be comparable to that of HEU machines; and burst yields and steady-state power levels will be significantly greater in an LEU reactor.

  18. Neurocontrol of Pressurized Water Reactors in Load-Follow Operations

    SciTech Connect

    Lin Chaung; Shen Chihming

    2000-12-15

    The neurocontrol technique was applied to control a pressurized water reactor (PWR) in load-follow operations. Generalized learning or direct inverse control architecture was adopted in which the neural network was trained off-line to learn the inverse model of the PWR. Two neural network controllers were designed: One provided control rod position, which controlled the axial power distribution, and the other provided the change in boron concentration, which adjusted core total power. An additional feedback controller was designed so that power tracking capability was improved. The time duration between control actions was 15 min; thus, the xenon effect is limited and can be neglected. Therefore, the xenon concentration was not considered as a controller input variable, which simplified controller design. Center target strategy and minimum boron strategy were used to operate the reactor, and the simulation results demonstrated the effectiveness and performance of the proposed controller.

  19. 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

  20. 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.

  1. 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.

  2. 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.

  3. 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.

  4. Cone rod dystrophies

    PubMed Central

    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

  5. Making Highly Pure Glass Rods

    NASA Technical Reports Server (NTRS)

    Naumann, R. J.

    1986-01-01

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

  6. 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.

  7. 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.

  8. 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.

  9. 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.

  10. Analysis of Dynamic Insertion of Control Rod of BWR under Seismic Excitation

    NASA Astrophysics Data System (ADS)

    Koide, Yuichi; Nakagawa, Masaki; Fukushi, Naoki; Ishigaki, Hirokuni; Okumura, Kazue

    The dynamic characteristics of control rod for boiling water reactor being inserted under seismic excitation were investigated using non-linear analytical models. The capability of managing the insertion of control rod is one of the most important factors affecting the safety of nuclear power plant undergoing seismic events. Predicting the behavior of control rod being inserted during earthquakes is important when designing how rod should be controlled during seismic events. We developed analytical models using the finite element method (FEM). The effect of the interaction force between the control rod and the fuel assemblies is considered in non-linear analysis. This interaction force causes resistance force to be applied to the control rod when they are being inserted. The validity of the analytical models was confirmed by comparing the analytical results with the experimental ones. The effects of input seismic motion and structural parameters on the insertion time ware investigated using the analytical models. These analytical methods can be used to predict the time to insert the control rod into the core region of reactor, and are useful for designing control rod system that can survive seismic events.

  11. Active Brownian rods

    NASA Astrophysics Data System (ADS)

    Peruani, Fernando

    2016-11-01

    Bacteria, chemically-driven rods, and motility assays are examples of active (i.e. self-propelled) Brownian rods (ABR). The physics of ABR, despite their ubiquity in experimental systems, remains still poorly understood. Here, we review the large-scale properties of collections of ABR moving in a dissipative medium. We address the problem by presenting three different models, of decreasing complexity, which we refer to as model I, II, and III, respectively. Comparing model I, II, and III, we disentangle the role of activity and interactions. In particular, we learn that in two dimensions by ignoring steric or volume exclusion effects, large-scale nematic order seems to be possible, while steric interactions prevent the formation of orientational order at large scales. The macroscopic behavior of ABR results from the interplay between active stresses and local alignment. ABR exhibit, depending on where we locate ourselves in parameter space, a zoology of macroscopic patterns that ranges from polar and nematic bands to dynamic aggregates.

  12. 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

  13. Autophagy supports survival and phototransduction protein levels in rod photoreceptors

    PubMed Central

    Zhou, Z; Doggett, T A; Sene, A; Apte, R S; Ferguson, T A

    2015-01-01

    Damage and loss of the postmitotic photoreceptors is a leading cause of blindness in many diseases of the eye. Although the mechanisms of photoreceptor death have been extensively studied, few studies have addressed mechanisms that help sustain these non-replicating neurons for the life of an organism. Autophagy is an intracellular pathway where cytoplasmic constituents are delivered to the lysosomal pathway for degradation. It is not only a major pathway activated in response to cellular stress, but is also important for cytoplasmic turnover and to supply the structural and energy needs of cells. We examined the importance of autophagy in photoreceptors by deleting the essential autophagy gene Atg5 specifically in rods. Loss of autophagy led to progressive degeneration of rod photoreceptors beginning at 8 weeks of age such that by 44 weeks few rods remained. Cone photoreceptor numbers were only slightly diminished following rod degeneration but their function was significantly decreased. Rod cell death was apoptotic but was not dependent on daily light exposure or accelerated by intense light. Although the light-regulated translocation of the phototransduction proteins arrestin and transducin were unaffected in rods lacking autophagy, Atg5-deficient rods accumulated transducin-α as they degenerated suggesting autophagy might regulate the level of this protein. This was confirmed when the light-induced decrease in transducin was abolished in Atg5-deficient rods and the inhibition of autophagy in retinal explants cultures prevented its degradation. These results demonstrate that basal autophagy is essential to the long-term health of rod photoreceptors and a critical process for maintaining optimal levels of the phototransduction protein transducin-α. As the lack of autophagy is associated with retinal degeneration and altered phototransduction protein degradation in the absence of harmful gene products, this process may be a viable therapeutic target where rod

  14. 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)

  15. 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.

  16. 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.

  17. Control of Rod-Rod Interactions in Poly(3-alkylthiophenes)

    NASA Astrophysics Data System (ADS)

    Ho, Victor; Boudouris, Bryan W.; Segalman, Rachel A.

    2010-03-01

    Poly(3-hexylthiophene) is a commonly used semiconducting polymer because of its relatively high charge transport ability, low band gap, and solution processiblity. Strong intermolecular interactions lead to the formation of nanofibers during crystallization, which prevents long-range microstructural ordering. We show rod-rod interactions, parameterized by the Maier-Saupe parameter, can be controlled by rational polythiophene side chain design. Effects of side chain passivation are evidenced by a depressed melting temperature and the presence of a liquid crystalline region. Additionally, the Maier-Saupe parameters are estimated for poly(3-dodecylthiophene) and poly(3-ethylhexylthiophene); the relative magnitudes of each are related to the interchain spacings obtained by x-ray diffraction experiments. The systematic tuning of the rod-rod interactions in polythiophenes allows for manipulation of the ratio of Maier-Saupe to the Flory-Huggins parameter, a crucial value in obtaining long-range order in rod-coil block copolymer morphologies.

  18. The effect of fuel rod oxidation on PCMI-induced fuel failure

    NASA Astrophysics Data System (ADS)

    Kim, Kyu-Tae

    2011-11-01

    It was found in a one of the PWRs operating in Korea that a few three cycle-burned Zry-4 fuel assemblies which were loaded in a core center region at control bank positions were leaking. The leaking cycle has experienced a few reactor trips and some fuel rods started to leak at about a month after a power ramp following the second reactor trip. To investigate a root cause of such fuel failure as well as to examine intact and leak rod oxidation behaviors, one intact and one leaking fuel rods were selected from one intact and one failed three cycle-burned fuel assemblies, respectively, and in parallel one intact fuel rod was selected from a two cycle-burned fuel assembly to examine the effect of burnup on fuel rod oxidation and cladding stress during the power ramp. The maximum oxide thicknesses for the intact two cycle-burned and three cycle-burned fuel rods were measured to be about 70 and 140 μm, respectively, whereas that for the leaking three cycle-burned fuel rod to be about 200 μm. The leaking fuel rods generated a very sharp increase in oxide thickness in the fuel rod upper region having a relatively high axial power, resulting in through-wall axial cracks. The root cause of the fuel rod leaks was evaluated to be the pellet-clad mechanical interaction (PCMI)-induced failure combined with excessive Zry-4 oxidation and cladding stress, based on the evaluations of pellet-clad friction coefficient-dependent cladding hoop stresses after the power ramp following the second trip, measured oxide thicknesses and axial cracks on the cladding surface, a fuel leak initiation time and failed fuel rod locations at the control bank positions.

  19. 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.

  20. Cuisenaire Rods Go to College.

    ERIC Educational Resources Information Center

    Chinn, Phyllis; And Others

    1992-01-01

    Presents examples of questions and answers arising from a hands-on and exploratory approach to discrete mathematics using cuisenaire rods. Combinatorial questions about trains formed of cuisenaire rods provide the setting for discovering numerical patterns by experimentation and organizing the results using induction and successive differences.…

  1. 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.

  2. Experiments with a falling rod

    NASA Astrophysics Data System (ADS)

    Oliveira, Vitor

    2016-02-01

    We study the motion of a uniform thin rod released from rest, with the bottom end initially in contact with a horizontal surface. Our focus here is the motion of the bottom end as the rod falls. For small angles of release with respect to the horizontal, the rod falls without the bottom end slipping. For larger angles, the slipping direction depends on the static friction coefficient between the rod and the horizontal surface. Small friction coefficients cause the end to slip initially in one direction and then in the other, while for high coefficients, the end slips in one direction only. For intermediate values, depending on the angle of release, both situations can occur. We find the initial slipping direction to depend on a relation between the angle at which the rod slips, and a critical angle at which the frictional force vanishes. Comparison between experimental data and numerical simulations shows good agreement.

  3. Topological mixing with ghost rods.

    PubMed

    Gouillart, Emmanuelle; Thiffeault, Jean-Luc; Finn, Matthew D

    2006-03-01

    Topological chaos relies on the periodic motion of obstacles in a two-dimensional flow in order to form nontrivial braids. This motion generates exponential stretching of material lines, and hence efficient mixing. Boyland, Aref, and Stremler [J. Fluid Mech. 403, 277 (2000)] have studied a specific periodic motion of rods that exhibits topological chaos in a viscous fluid. We show that it is possible to extend their work to cases where the motion of the stirring rods is topologically trivial by considering the dynamics of special periodic points that we call "ghost rods", because they play a similar role to stirring rods. The ghost rods framework provides a new technique for quantifying chaos and gives insight into the mechanisms that produce chaos and mixing. Numerical simulations for Stokes flow support our results.

  4. 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.

  5. Efficiency Calibration and Testing of a Commercial Active Well Coincidence Neutron to Inventory U-Al Reactor Fuel and Target Elements at the Savannah River Site

    SciTech Connect

    Dewberry, R.A.

    2001-06-20

    In this paper we describe testing, efficiency calibration and use of the Aquila active well neutron coincidence counter for nuclear material assays of U-235 in Savannah River Site (SRS) U-Al fuel. All of the SRS production reactors have been in standby mode since 1993. However the Department of Energy still requires material control and accountability of the fuel and target material stored in the reactor assembly areas. Because the SRS fuel is pure of plutonium, it has a very low passive neutron emission rate and is very well suited for assay of the U-235 content by active well induced fission neutron counting. These U-Al active well assays for inventory verification are compared to the fuel element assays attained from the SRS fuel production facility Californium shuffler. In addition we describe experiments performed at Argonne West National Laboratory with the Aquila active well counter. In these experiments we tested the capability of the instrument to determine U-235 content in metallic uranium chunks of non-uniform shape.

  6. 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.

  7. Scaled long rod penetration experiments: Tungsten against rolled homogeneous armour

    NASA Astrophysics Data System (ADS)

    Proud, William; Cross, Daniel

    2012-03-01

    Scaled, reverse ballistic, long-rod experiments were performed at an impact velocity of ~700 m s-1. The targets were tungsten alloy rods and the projectiles either 3 or 6 mm thick rolled homogeneous armour (RHA) plates. The plate was inclined at 60° to the direction of travel and the interaction was recorded using high-speed photography, strain gauges and laser velocimetry. The pitch of the rod was varied in steps of 3° over a total range of 15°. In this range the rod deformation changed dramatically the bending process moved from a flexing of the tip away from the plate, to a marked motion into the surface. Cross comparison of the diagnostic outputs reveals the time windows for these process and also the varying sensitivity of the measurement system to that process. Post-impact recovery was also performed.

  8. 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.

  9. 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

  10. Fast Reactors

    NASA Astrophysics Data System (ADS)

    Esposito, S.; Pisanti, O.

    The following sections are included: * Elementary Considerations * The Integral Equation to the Neutron Distribution * The Critical Size for a Fast Reactor * Supercritical Reactors * Problems and Exercises

  11. 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.

  12. 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.

  13. 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.

  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. 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.

  16. 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.

  17. 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.

  18. Topological mixing with ghost rods

    NASA Astrophysics Data System (ADS)

    Gouillart, Emmanuelle; Thiffeault, Jean-Luc; Finn, Matthew D.

    2006-03-01

    Topological chaos relies on the periodic motion of obstacles in a two-dimensional flow in order to form nontrivial braids. This motion generates exponential stretching of material lines, and hence efficient mixing. Boyland, Aref, and Stremler [J. Fluid Mech. 403, 277 (2000)] have studied a specific periodic motion of rods that exhibits topological chaos in a viscous fluid. We show that it is possible to extend their work to cases where the motion of the stirring rods is topologically trivial by considering the dynamics of special periodic points that we call “ghost rods”, because they play a similar role to stirring rods. The ghost rods framework provides a new technique for quantifying chaos and gives insight into the mechanisms that produce chaos and mixing. Numerical simulations for Stokes flow support our results.

  19. 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.

  20. Cadmium control/safety rod disposal at the Savannah River Site

    SciTech Connect

    McInnis, S.H.

    1995-05-01

    Four heavy-water-moderated reactors at the Savannah River Site will undergo the removal of 862 activated cadmium control/safety rods. Although these reactors are 40 years old, they offer 4 basic advantages for decommissioning: the equipment is still in some sort of operable state; the reactor is blow the floor in a large process room, allowing access; Control/safety rods can be handled remotely by existing equipment; a radiologically shielded removal path exists. Drawbacks include the following: age of reactors; improvements in technology have caused incompatibility problems; more strigent standards; compliance with environmental regulations. This article details how the removal was carried out and the current status of the project, keeping in mind the above considerations.

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

    NASA Astrophysics Data System (ADS)

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

    2013-01-01

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

  2. Hydrothermal synthesis of hydroxyapatite rods

    NASA Astrophysics Data System (ADS)

    Zhang, Xing; Vecchio, Kenneth S.

    2007-10-01

    Hydroxyapatite (HAP) rods were synthesized from dicalcium phosphate anhydrous (CaHPO 4, DCPA) and calcium carbonate (CaCO 3) by the hydrothermal method from 120 to 180 °C. Both cuttlebone (aragonite polymorph of CaCO 3) and CaCO 3 chemical (calcite polymorph of CaCO 3) were used as CaCO 3 sources. The nucleation and growth of HAP rods mainly occurred on DCPA particles, while some HAP rods also grew from aragonite particles. The nucleation and growth of β-tricalcium phosphate (β-TCP) particles on the surface of calcite particles were observed at the beginning of the reaction of DCPA and calcite, and some HAP rods were also found to grow out of β-TCP particles. After the hydrothermal reaction at 140 °C for 24 h, most products are HAP with a small amount of β-TCP synthesized as a byproduct. The HAP rods synthesized were ˜200 nm in width and several microns in length. The reaction mechanism and growth process of HAP rods are discussed.

  3. 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.

  4. 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.

  5. 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.

  6. Actin dynamics and cofilin-actin rods in Alzheimer disease

    PubMed Central

    Bamburg, James R.; Bernstein, Barbara W.

    2017-01-01

    Cytoskeletal abnormalities and synaptic loss, typical of both familial and sporadic Alzheimer disease (AD), are induced by diverse stresses such as neuroinflammation, oxidative stress, and energetic stress, each of which may be initiated or enhanced by proinflammatory cytokines or amyloid-β (Aβ) peptides. Extracellular Aβ-containing plaques and intracellular phospho-tau-containing neurofibrillary tangles are postmortem pathologies required to confirm AD and have been the focus of most studies. However, AD brain, but not normal brain, also have increased levels of cytoplasmic rod-shaped bundles of filaments composed of ADF/cofilin-actin in a 1:1 complex (rods). Cofilin, the major ADF/cofilin isoform in mammalian neurons, severs actin filaments at low cofilin/actin ratios and stabilizes filaments at high cofilin/actin ratios. It binds cooperatively to ADP-actin subunits in F-actin. Cofilin is activated by dephosphorylation and may be oxidized in stressed neurons to form disulfide-linked dimers, required for bundling cofilin-actin filaments into stable rods. Rods form within neurites causing synaptic dysfunction by sequestering cofilin, disrupting normal actin dynamics, blocking transport, and exacerbating mitochondrial membrane potential loss. Aβ and proinflammatory cytokines induce rods through a cellular prion protein-dependent activation of NADPH oxidase and production of reactive oxygen species. Here we review recent advances in our understanding of cofilin biochemistry, rod formation, and the development of cognitive deficits. We will then discuss rod formation as a molecular pathway for synapse loss that may be common between all three prominent current AD hypotheses, thus making rods an attractive therapeutic target. PMID:26873625

  7. Evidence of fast nonlinear feedback in EBR-II rod-drop measurements

    SciTech Connect

    Grimm, K.N.; Meneghetti, D.

    1987-01-01

    Feedback reactivities determine the time dependence of a reactor during and after a transient initiating event. Recent analysis of control-rod drops in the Experimental Breeder Reactor II (EBR-II) has indicated that some relatively fast feedback may exist that cannot be accounted for by the linear feedback mechanisms. The magnitude of the positive insertion appears dependent on the amount of inserted reactivity and the run configuration. This phenomenon may be caused by a small, but rapid, change in core dimensions.

  8. 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.

  9. 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.

  10. Behavior of Segmented Rods during Penetration

    DTIC Science & Technology

    1990-07-01

    full-scale penetrators which had been swaged 24%. The density of this tungsten alloy was 17.2 Mg/m 3. Gold-alloy penetrators were composed of 92Au-4.9Ag...of behavior. Segmented rods of tungsten alloy always penetrated less than the equivalent unitary rod. Successive rod segments were found to...gold-alloy penetrators because unitary rods of this material surpassed the perform- ance of unitary tungsten -alloy rods, while leaving almost no

  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. 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.

  13. 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

  14. 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)

  15. The evolution of rod photoreceptors.

    PubMed

    Morshedian, Ala; Fain, Gordon L

    2017-04-05

    Photoreceptors in animals are generally of two kinds: the ciliary or c-type and the rhabdomeric or r-type. Although ciliary photoreceptors are found in many phyla, vertebrates seem to be unique in having two distinct kinds which together span the entire range of vision, from single photons to bright light. We ask why the principal photoreceptors of vertebrates are ciliary and not rhabdomeric, and how rods evolved from less sensitive cone-like photoreceptors to produce our duplex retina. We suggest that the principal advantage of vertebrate ciliary receptors is that they use less ATP than rhabdomeric photoreceptors. This difference may have provided sufficient selection pressure for the development of a completely ciliary eye. Although many of the details of rod evolution are still uncertain, present evidence indicates that (i) rods evolved very early before the split between the jawed and jawless vertebrates, (ii) outer-segment discs make no contribution to rod sensitivity but may have evolved to increase the efficiency of protein renewal, and (iii) evolution of the rod was incremental and multifaceted, produced by the formation of several novel protein isoforms and by changes in protein expression, with no one alteration having more than a few-fold effect on transduction activation or inactivation.This article is part of the themed issue 'Vision in dim light'.

  16. 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).

  17. 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)

  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. Topological Optimization of Rod Mixers

    NASA Astrophysics Data System (ADS)

    Finn, Matthew D.; Thiffeault, Jean-Luc

    2006-11-01

    Stirring of fluid with moving rods is necessary in many practical applications to achieve homogeneity. These rods are topological obstacles that force stretching of fluid elements. The resulting stretching and folding is commonly observed as filaments and striations, and is a precursor to mixing. In a space-time diagram, the trajectories of the rods form a braid [1], and the properties of this braid impose a minimal complexity in the flow. We discuss how optimal mixing protocols can be obtained by a judicious choice of braid, and how these protocols can be implemented using simple gearing [2].[12pt] [1] P. L. Boyland, H. Aref, and M. A. Stremler, JFM 403, 277 (2000).[8pt] [2] J.-L. Thiffeault and M. D. Finn, http://arxiv.org/nlin/0603003

  20. 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.

  1. BOILING REACTORS

    DOEpatents

    Untermyer, S.

    1962-04-10

    A boiling reactor having a reactivity which is reduced by an increase in the volume of vaporized coolant therein is described. In this system unvaporized liquid coolant is extracted from the reactor, heat is extracted therefrom, and it is returned to the reactor as sub-cooled liquid coolant. This reduces a portion of the coolant which includes vaporized coolant within the core assembly thereby enhancing the power output of the assembly and rendering the reactor substantially self-regulating. (AEC)

  2. 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.

  3. 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.

  4. 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.

  5. 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.

  6. 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.

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

    SciTech Connect

    Michael L. Wilson

    2001-02-08

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

  8. Review of CTF s Fuel Rod Modeling Using FRAPCON-4.0 s Centerline Temperature Predictions

    SciTech Connect

    Toptan, Aysenur; Salko, Robert K; Avramova, Maria

    2017-01-01

    Coolant Boiling in Rod Arrays Two Fluid (COBRA-TF), or CTF1 [1], is a nuclear thermal hydraulic subchannel code used throughout academia and industry. CTF s fuel rod modeling is originally developed for VIPRE code [2]. Its methodology is based on GAPCON [3] and FRAP [4] fuel performance codes, and material properties are included from MATPRO handbook [5]. This work focuses on review of CTF s fuel rod modeling to address shortcomings in CTF s temperature predictions. CTF is compared to FRAPCON which is U.S. NRC s steady-state fuel performance code for light-water reactor fuel rods. FRAPCON calculates the changes in fuel rod variables and temperatures including the eects of cladding hoop strain, cladding oxidation, hydriding, fuel irradiation swelling, densification, fission gas release and rod internal gas pressure. It uses fuel, clad and gap material properties from MATPRO. Additionally, it has its own models for fission gas release, cladding corrosion and cladding hydrogen pickup. It allows finite dierence or finite element approaches for its mechanical model. In this study, FRAPCON-4.0 [6] is used as a reference fuel performance code. In comparison, Halden Reactor Data for IFA432 Rod 1 and Rod 3. CTF simulations are performed in two ways; informing CTF with gap conductance value from FRAPCON, and using CTF s dynamic gap conductance model. First case is chosen to show temperature is predicted correctly with CTF s models for thermal and cladding conductivities once gap conductance is provided. Latter is to review CTF s dynamic gap conductance model. These Halden test cases are selected to be representative of cases with and without any physical contact between fuel-pellet and clad while reviewing functionality of CTF s dynamic gap conductance model. Improving the CTF s dynamic gap conductance model will allow prediction of fuel and cladding thermo-mechanical behavior under irradiation, and better temperature feedbacks from CTF in transient calculations.

  9. 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.

  10. 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)

  11. Method and apparatus for preventing and repairing vibrational damage to nuclear fuel rods

    SciTech Connect

    Busselman, G.J.; Reparaz, A.

    1988-11-15

    This patent describes a nuclear reactor comprising a core barrel, baffle plates within the core barrel, the baffle plates having joints therebetween, stationary fuel assemblies at least partially surrounded by the baffle plates. Each of the assemblies comprising multiple parallel rows of elongated, small diameter, fuel rods, some but not all, of the assemblies having rods adjacent and exposed to the joints between baffle plates, and means for introducing cooling water between the core barrel and the baffle plates, which water flows down, around the baffle plates, and up through the assemblies whereby water jetting through the joints may induce vibrations in the rods adjacent the joints; the improvement comprising: spring clip means disposed between the ends of the fuel rods adjacent the joints, gripping and securing together a plurality of the rods, thereby reducing the vibration, the clip means being removable from the fuel rods in a non-parallel direction to the axes of the fuel rods without requiring even partial disassembly of the latter.

  12. 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.

  13. 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.

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

    NASA Astrophysics Data System (ADS)

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

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

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

    SciTech Connect

    Holger Pfeifer; Norman Meinert

    2000-06-04

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

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

    SciTech Connect

    Pfeifer, H.; Meinert, N.

    2000-07-01

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

  17. Evidence of fast non-linear feedback in EBR-II rod-drop measurements

    SciTech Connect

    Grimm, K.N.; Meneghetti, D.

    1987-06-01

    Feedback reactivities determine the time dependence of a reactor during and after a transient initiating event. Recent analysis of control-rod drops in the Experimental Breeder Reactor II (EBR-II) Reactor has indicated that some relatively fast feedback may exist which cannot be accounted for by the linear feedback mechanisms. The linear and deduced non-linear feedback reactivities from a control-rod drop in EBR-II run 93A using detailed temperature coefficients of reactivity in the EROS kinetics code have been reported. The transient analyses have now been examined in more detail for times close to the drop to ascertain if additional positive reactivity is being built-in early in the drop which could be gradually released later in the drop.

  18. Investigation of Lightning Rod Shielding Angle

    NASA Astrophysics Data System (ADS)

    Nayel, Mohamed

    This paper studies those parameters affecting the shielding angle of the lightning rod (Franklin Rod) above very tall buildings. It was recommended that the shielding angle of the lightning rod is about 45°∼60°. The downward lightning leader is modeled by using discrete line charges to consider the exponential distribution of charges through the downward leader. The voltage condition used by Rizk is used to investigate the inception of the upward lightning leader. Different air conditions (relative air density and air humidity) are considered for more practical simulation. The influences of lightning parameters and lightning rod height on the shielding angle are studied. The results shows that, lightning leader parameters, lightning rod height and ground slope have series effects on the lightning rod shielding angle. Based on the results, a lightning rod shielding angle for shielding design is recommended to decrease the lightning stroke to the lightning rod.

  19. Control rods in LMFBRs: a physics assessment

    SciTech Connect

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

    1982-08-01

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

  20. Plum Brook Reactor Facility Control Room during Facility Startup

    NASA Image and Video Library

    1961-02-21

    Operators test the National Aeronautics and Space Administration’s (NASA) Plum Brook Reactor Facility systems in the months leading up to its actual operation. The “Reactor On” signs are illuminated but the reactor core was not yet ready for chain reactions. Just a couple weeks after this photograph, Plum Brook Station held a media open house to unveil the 60-megawatt test reactor near Sandusky, Ohio. More than 60 members of the print media and radio and television news services met at the site to talk with community leaders and representatives from NASA and Atomic Energy Commission. The Plum Brook reactor went critical for the first time on the evening of June 14, 1961. It was not until April 1963 that the reactor reached its full potential of 60 megawatts. The reactor control room, located on the second floor of the facility, was run by licensed operators. The operators manually operated the shim rods which adjusted the chain reaction in the reactor core. The regulating rods could partially or completely shut down the reactor. The control room also housed remote area monitoring panels and other monitoring equipment that allowed operators to monitor radiation sensors located throughout the facility and to scram the reactor instantly if necessary. The color of the indicator lights corresponded with the elevation of the detectors in the various buildings. The reactor could also shut itself down automatically if the monitors detected any sudden irregularities.

  1. 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.

  2. Validation of the Monte Carlo model developed to assess the activity generated in control rods of a BWR

    NASA Astrophysics Data System (ADS)

    Ródenas, José; Abarca, Agustín; Gallardo, Sergio; Sollet, Eduardo

    2010-07-01

    Control rods are activated by neutron reactions into the reactor. The activation is produced mainly in stainless steel and its impurities. The dose produced by this activity is not important inside the reactor, but it has to be taken into account when the rod is withdrawn from it. The neutron activation has been modeled with the MCNP5 code based on the Monte Carlo method. The number of reactions obtained with the code can be converted into activity. In this work, a detailed model of the control rod has been developed considering all its components: handle, tubes, gain, and central core. On the other hand, the rod has been divided into 5 zones in order to consider the different axial exposition to neutron flux into the reactor. Results of the Monte Carlo simulation for the neutron activation constitute a gamma source in the control rod. With this source, applying again the Monte Carlo method, doses at certain distance of the rod have been calculated. Comparison of calculated doses with experimental measurements leads to the validation of the model developed.

  3. Reactor production of Thorium-229

    DOE PAGES

    Boll, Rose Ann; Murphy, Karen E.; Denton, David L.; ...

    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.

  4. 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 is...

  5. 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 is...

  6. 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.

  7. A Heterogeneous Sodium Fast Reactor Designed to Transmute Minor Actinide Actinide Waste Isotopes into Plutonium Fuel

    SciTech Connect

    Samuel E. Bays

    2011-02-01

    An axial heterogeneous sodium fast reactor design is developed for converting minor actinide waste isotopes into plutonium fuel. The reactor design incorporates zirconium hydride moderating rods in an axial blanket above the active core. The blanket design traps the active core’s axial leakage for the purpose of transmuting Am-241 into Pu-238. This Pu-238 is then co-recycled with the spent driver fuel to make new driver fuel. Because Pu-238 is significantly more fissile than Am-241 in a fast neutron spectrum, the fissile worth of the initial minor actinide material is upgraded by its preconditioning via transmutation in the axial targets. Because, the Am-241 neutron capture worth is significantly stronger in a moderated epithermal spectrum than the fast spectrum, the axial targets serve as a neutron trap which recovers the axial leakage lost by the active core. The sodium fast reactor proposed by this work is designed as an overall transuranic burner. Therefore, a low transuranic conversion ratio is achieved by a degree of core flattening which increases axial leakage. Unlike a traditional “pancake” design, neutron leakage is recovered by the axial target/blanket system. This heterogeneous core design is constrained to have sodium void and Doppler reactivity worth similar to that of an equivalent homogeneous design. Because minor actinides are irradiated only once in the axial target region; elemental partitioning is not required. This fact enables the use of metal targets with electrochemical reprocessing. Therefore, the irradiation environment of both drivers and targets was constrained to ensure applicability of the established experience database for metal alloy sodium fast reactor fuels.

  8. 76 FR 68514 - Request for a License To Export Reactor Components

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-11-04

    ... reactor 12 Perform seismic China. LLC, August 18, 2011, October control rod system testing necessary 6, 2011, XR174, 11005963. and associated for qualification equipment. of AP1000 (design) nuclear...

  9. Subplane-based Control Rod Decusping Techniques for the 2D/1D Method in MPACT

    SciTech Connect

    Graham, Aaron M; Collins, Benjamin S; Downar, Thomas

    2017-01-01

    The MPACT transport code is being jointly developed by Oak Ridge National Laboratory and the University of Michigan to serve as the primary neutron transport code for the Virtual Environment for Reactor Applications Core Simulator. MPACT uses the 2D/1D method to solve the transport equation by decomposing the reactor model into a stack of 2D planes. A fine mesh flux distribution is calculated in each 2D plane using the Method of Characteristics (MOC), then the planes are coupled axially through a 1D NEM-P$_3$ calculation. This iterative calculation is then accelerated using the Coarse Mesh Finite Difference method. One problem that arises frequently when using the 2D/1D method is that of control rod cusping. This occurs when the tip of a control rod falls between the boundaries of an MOC plane, requiring that the rodded and unrodded regions be axially homogenized for the 2D MOC calculations. Performing a volume homogenization does not properly preserve the reaction rates, causing an error known as cusping. The most straightforward way of resolving this problem is by refining the axial mesh, but this can significantly increase the computational expense of the calculation. The other way of resolving the partially inserted rod is through the use of a decusping method. This paper presents new decusping methods implemented in MPACT that can dynamically correct the rod cusping behavior for a variety of problems.

  10. 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)

  11. 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.

  12. 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.

  13. 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.

  14. 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).

  15. 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.

  16. 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.

  17. 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.

  18. Wear resistant zirconium base alloy article for water reactors

    SciTech Connect

    Gillett, J.E.; Shockling, L.A.; Sherwood, D.G.

    1988-03-01

    In a water reactor operating environment, the combination having improved fretting wear resistance is described comprising: an elongated tubular water displacer rod; having a low neutron absorption cross section guide support plates distributed along the length of the water displacer rod; the water displacer rod intersecting the guide support plates through apertures in the guide support plates; the water displacer rod having a plurality of spaced apart annular electrospark deposited coatings, each coating facing the wall of a respective aperture, the electrospark deposited coatings comprising Cr/sub 2/C/sub 3/; wherein the water displacer rod has a tube wall composed of a zirconium base alloy; and wherein the guide support plates are composed of a stainless steel alloy.

  19. 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.

  20. 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.

  1. Safety analysis of the advanced thermionic initiative reactor

    NASA Astrophysics Data System (ADS)

    Lee, Hsing H.; Klein, Andrew C.

    1995-01-01

    Previously, detailed analysis was conducted to assess the technology developed for the Advanced Thermionic Initiative reactor. This analysis included the development of an overall system design code capability and the improvement of analytical models necessary for the assessment of the use of single cell thermionic fuel elements in a low power space nuclear reactor. The present analysis extends this effort to assess the nuclear criticality safety of the ATI reactor for various different scenarios. The analysis discusses the efficacy of different methods of reactor control such as control rods, and control drums.

  2. 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.

  3. 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

  4. Rod lengthening with the magnetically controlled growing rod: factors influencing rod slippage and reduced gains during distractions.

    PubMed

    Cheung, Jason Pui Yin; Yiu, Karen Kar Lum; Samartzis, Dino; Kwan, Kenny; Tan, Boon-Beng; Cheung, Kenneth Man Chee

    2017-08-01

    Prospective study OBJECTIVE.: To identify the factors that are associated with rod slippage and to study the pattern of achieved length gain with a standard distraction methodology. Ability to achieve successful magnetically controlled growing rod (MCGR) distraction is crucial for gradual spine lengthening. Rod slippage has been described as a failure of internal magnet rotation leading to a slippage and an inability to distract the rod. However, its onset, significance and risk factors are currently unknown. In addition, how this phenomenon pertains to actual distracted lengths is unknown. A total of 22 patients with MCGR and at least 6 distraction episodes were prospectively studied. Patients with rod slippage occurring ≤6 distraction episodes were considered early rod slippage while those >6 or have yet not slipped were grouped as late rod slippage. The association of parameters including body habitus, maturity status, age of implantation, total number of distractions, months of distraction from initial implantation, initial and postoperative Cobb angle, T1-T12, T1-S1, T5-T12 kyphosis, curve flexibility, fusion block length, and distance between magnets in dual rods and between the magnets and apex of the curve with early or late onset of rod slippage was studied. Differences between expected and achieved distraction lengths were assessed with reference to rod slippage episodes and rod exchanges to determine any patterns of diminishing returns. Patients had mean age of 7.1 years at diagnosis with mean follow-up of 49.8 months. A mean 32.4 distractions were performed per patient. Early rod slippage occurred in 14 patients and late rod slippage occurred in 8 patients. Increased height, weight, body mass index, older age, increased T1-12 and T1-S1 lengths, and less distance between magnets were significantly associated with early rod slippage. Expected distraction lengths did not translate to achieved distraction lengths and reduced gains were only observed after

  5. 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

  6. Computational Fluid Dynamics Modeling of Mono-Silane Siemens Reactor

    NASA Astrophysics Data System (ADS)

    Jung, Hosub; Park, Jong Hoon; Kang, Seung Oh; Jeong, Jong Hyun; Jeon, Soyoung; Jung, Jae Hak; Kim, Woo Kyoung

    2012-10-01

    The computational fluid dynamics-based FLUENT program was employed to model the heat transfer and chemical reaction in a mono-silane Siemens reactor. The kinetic parameters for the 1-step overall reaction SiH4→Si+ 2H2, such as the pre-exponential factor, temperature coefficient, and activation energy, were carefully optimized to satisfy experimental data obtained from the 4-rod Siemens pilot reactor. Established models were successfully used to evaluate the effects of rod diameter, reaction temperature, and reactant gas flow rate on the deposition rate of silicon.

  7. Hypoxic preconditioning protects photoreceptors against light damage independently of hypoxia inducible transcription factors in rods.

    PubMed

    Kast, Brigitte; Schori, Christian; Grimm, Christian

    2016-05-01

    Hypoxic preconditioning protects photoreceptors against light-induced degeneration preserving retinal morphology and function. Although hypoxia inducible transcription factors 1 and 2 (HIF1, HIF2) are the main regulators of the hypoxic response, photoreceptor protection does not depend on HIF1 in rods. Here we used rod-specific Hif2a single and Hif1a;Hif2a double knockout mice to investigate the potential involvement of HIF2 in rods for protection after hypoxic preconditioning. To identify potential HIF2 target genes in rods we determined the retinal transcriptome of hypoxic control and rod-specific Hif2a knockouts by RNA sequencing. We show that rods do not need HIF2 for hypoxia-induced increased survival after light exposure. The transcriptomic analysis revealed a number of genes that are potentially regulated by HIF2 in rods; among those were Htra1, Timp3 and Hmox1, candidates that are interesting due to their connection to human degenerative diseases of the retina. We conclude that neither HIF1 nor HIF2 are required in photoreceptors for protection by hypoxic preconditioning. We hypothesize that HIF transcription factors may be needed in other cells to produce protective factors acting in a paracrine fashion on photoreceptor cells. Alternatively, hypoxic preconditioning induces a rod-intrinsic response that is independent of HIF transcription factors.

  8. Conceptual design of the advanced marine reactor MRX

    NASA Astrophysics Data System (ADS)

    1991-02-01

    Design studies on the advanced marine reactors have been done continuously since 1983 at the Japan Atomic Energy Research Institute (JAERI) in order to develop attractive marine reactors for the next generation. At present, two marine reactor concepts are being formulated. One is 100 MWt MRX (Marine Reactor X) for an icebreaker and the other is 300 kWe DRX (Deep-sea Reactor X) for a deep-sea research vessel. They are characterized by an integral type pressurized water reactor (PWR) built-in type control rod drive mechanisms, a water-filled container and a passive decay heat removal system, which realize highly passive safe and compact reactors. This paper is a detailed report including all major results of the MRX design study.

  9. 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

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

    PubMed

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

    2010-01-01

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

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

    NASA Astrophysics Data System (ADS)

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

    2010-02-01

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

  12. 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. Fractographic Analysis of Long Rod Penetrator-Armor Ceramic Interactions

    DTIC Science & Technology

    1989-10-01

    a 97 wcight-pcrccnt tungsten (97 W) long-rod penietrator with an aspect ratio of’ 10 (L/D = 10) and a laminate armor package consisting of steel ESR ...carefully moved so as not to disturb the comminuted material. The ESR 4340 back plate was removed from the target and the comminuted ceramic was... steel particles were included in the ceramic powder.\\ The steel was present from thrce sources: front plate penetration, back plate splash, and

  14. 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.

  15. 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.

  16. 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.

  17. Chemotaxis of Nonbiological Colloidal Rods

    NASA Astrophysics Data System (ADS)

    Hong, Yiying; Blackman, Nicole M. K.; Kopp, Nathaniel D.; Sen, Ayusman; Velegol, Darrell

    2007-10-01

    Chemotaxis is the movement of organisms toward or away from a chemical attractant or toxin by a biased random walk process. Here we describe the first experimental example of chemotaxis outside biological systems. Platinum-gold rods 2.0μm long exhibit directed movement toward higher hydrogen peroxide concentrations through “active diffusion.” Brownian dynamics simulations reveal that no “temporal sensing” algorithm, commonly attributed to bacteria, is necessary; rather, the observed chemotaxis can be explained by random walk physics in a gradient of the active diffusion coefficient.

  18. 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.

  19. Compact Reactor

    NASA Astrophysics Data System (ADS)

    Williams, Pharis E.

    2007-01-01

    Weyl's Gauge Principle of 1929 has been used to establish Weyl's Quantum Principle (WQP) that requires that the Weyl scale factor should be unity. It has been shown that the WQP requires the following: quantum mechanics must be used to determine system states; the electrostatic potential must be non-singular and quantified; interactions between particles with different electric charges (i.e. electron and proton) do not obey Newton's Third Law at sub-nuclear separations, and nuclear particles may be much different than expected using the standard model. The above WQP requirements lead to a potential fusion reactor wherein deuterium nuclei are preferentially fused into helium nuclei. Because the deuterium nuclei are preferentially fused into helium nuclei at temperatures and energies lower than specified by the standard model there is no harmful radiation as a byproduct of this fusion process. Therefore, a reactor using this reaction does not need any shielding to contain such radiation. The energy released from each reaction and the absence of shielding makes the deuterium-plus-deuterium-to-helium (DDH) reactor very compact when compared to other reactors, both fission and fusion types. Moreover, the potential energy output per reactor weight and the absence of harmful radiation makes the DDH reactor an ideal candidate for space power. The logic is summarized by which the WQP requires the above conditions that make the prediction of DDH possible. The details of the DDH reaction will be presented along with the specifics of why the DDH reactor may be made to cause two deuterium nuclei to preferentially fuse to a helium nucleus. The presentation will also indicate the calculations needed to predict the reactor temperature as a function of fuel loading, reactor size, and desired output and will include the progress achieved to date.

  20. Compact Reactor

    SciTech Connect

    Williams, Pharis E.

    2007-01-30

    Weyl's Gauge Principle of 1929 has been used to establish Weyl's Quantum Principle (WQP) that requires that the Weyl scale factor should be unity. It has been shown that the WQP requires the following: quantum mechanics must be used to determine system states; the electrostatic potential must be non-singular and quantified; interactions between particles with different electric charges (i.e. electron and proton) do not obey Newton's Third Law at sub-nuclear separations, and nuclear particles may be much different than expected using the standard model. The above WQP requirements lead to a potential fusion reactor wherein deuterium nuclei are preferentially fused into helium nuclei. Because the deuterium nuclei are preferentially fused into helium nuclei at temperatures and energies lower than specified by the standard model there is no harmful radiation as a byproduct of this fusion process. Therefore, a reactor using this reaction does not need any shielding to contain such radiation. The energy released from each reaction and the absence of shielding makes the deuterium-plus-deuterium-to-helium (DDH) reactor very compact when compared to other reactors, both fission and fusion types. Moreover, the potential energy output per reactor weight and the absence of harmful radiation makes the DDH reactor an ideal candidate for space power. The logic is summarized by which the WQP requires the above conditions that make the prediction of DDH possible. The details of the DDH reaction will be presented along with the specifics of why the DDH reactor may be made to cause two deuterium nuclei to preferentially fuse to a helium nucleus. The presentation will also indicate the calculations needed to predict the reactor temperature as a function of fuel loading, reactor size, and desired output and will include the progress achieved to date.

  1. How To Make Singing Rods Scream

    NASA Astrophysics Data System (ADS)

    Minnix, Richard B.; Carpenter, D. Rae; McNairy, William W.

    1999-11-01

    The singing rod has long been a favorite of physics demonstrators. Longitudinal vibrations are easily excited for mode numbers n = 1 and n = 2 for short rods and even higher mode numbers for rods about a meter or greater in length. Aluminum is favored because of its higher Q, or tendency to maintain its resonance with a minimum of damping. Occasionally one finds by chance that an aluminum rod screams intermittently when one of the transverse modes becomes coupled to the longitudinal mode as a result of an integer ratio between the longitudinal and transverse frequencies. Computer software is now available to Fourier analyze the various waves emanating from a rod and read out the numerical values of the frequencies. This has made it possible to measure the multiplicity of frequencies present when coupling occurs and compare theoretical and experimental values. Demonstrations, with accompanying data on frequencies and mode numbers, will be shown for rods of both circular and square cross section.

  2. 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.

  3. 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.

  4. 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.

  5. Quantitative uncertainty and sensitivity analysis of a PWR control rod ejection accident

    SciTech Connect

    Pasichnyk, I.; Perin, Y.; Velkov, K.

    2013-07-01

    The paper describes the results of the quantitative Uncertainty and Sensitivity (U/S) Analysis of a Rod Ejection Accident (REA) which is simulated by the coupled system code ATHLET-QUABOX/CUBBOX applying the GRS tool for U/S analysis SUSA/XSUSA. For the present study, a UOX/MOX mixed core loading based on a generic PWR is modeled. A control rod ejection is calculated for two reactor states: Hot Zero Power (HZP) and 30% of nominal power. The worst cases for the rod ejection are determined by steady-state neutronic simulations taking into account the maximum reactivity insertion in the system and the power peaking factor. For the U/S analysis 378 uncertain parameters are identified and quantified (thermal-hydraulic initial and boundary conditions, input parameters and variations of the two-group cross sections). Results for uncertainty and sensitivity analysis are presented for safety important global and local parameters. (authors)

  6. 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.

  7. Growth and Morphology of Rod Eutectics

    SciTech Connect

    Jing Teng; Shan Liu; R. Trivedi

    2008-03-17

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

  8. Rod-climbing effect in Newtonian fluids.

    PubMed

    Bonn, Daniel; Kobylko, Mathias; Bohn, Steffen; Meunier, Jacques; Morozov, Alexander; van Saarloos, Wim

    2004-11-19

    When a rotating rod is brought into a polymer melt or concentrated polymer solution, the meniscus climbs the rod. This spectacular rod climbing is due to the normal stresses present in the polymer fluid and is thus a purely non-Newtonian effect. A similar rod climbing of an interface between two fluids has therefore been taken as a signature that one of the fluids exhibits normal stress effects. We show here, however, that the effect can occur with simple Newtonian fluids: it occurs when a Taylor-Couette instability happens in the less viscous of the two liquids but not in the more viscous one.

  9. Acoustic loading effects on oscillating rod bundles

    SciTech Connect

    Lin, W.H.

    1980-01-01

    An analytical study of the interaction between an infinite acoustic medium and a cluster of circular rods is described. The acoustic field due to oscillating rods and the acoustic loading on the rods are first solved in a closed form. The acoustic loading is then used as a forcing function for rod responses, and the acousto-elastic couplings are solved simultaneously. Numerical examples are presented for several cases to illustrate the effects of various system parameters on the acoustic reaction force coefficients. The effect of the acoustic loading on the coupled eigenfrequencies are discussed.

  10. Equations determine reasonable rod pump submergence depth

    SciTech Connect

    Hu Yongquan; Cai Wizhong

    1997-03-24

    A reasonable rod pump submergence depth can be calculated by combining fluid level changes with piston travel. Submergence depth is affected by the pump fill factor, reservoir fluid viscosity, rod pump type, and pumping parameters such as pump diameter, polished-rod stroke length, and pumping speed. Fluid level velocity can be obtained with an energy balance, and piston travel rate is based on the polished-rod travel. The paper describes the pump fill factor, piston travel velocity, fluid level rise, flow coefficient, reasonable submergence depth, and results from equations.

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

    SciTech Connect

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

    2012-07-01

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

  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. NEUTRONIC REACTOR

    DOEpatents

    Fermi, E.

    1960-04-01

    A nuclear reactor is described consisting of blocks of graphite arranged in layers, natural uranium bodies disposed in holes in alternate layers of graphite blocks, and coolant tubes disposed in the layers of graphite blocks which do not contain uranium.

  14. NEUTRONIC REACTOR

    DOEpatents

    Anderson, H.L.

    1960-09-20

    A nuclear reactor is described comprising fissionable material dispersed in graphite blocks, helium filling the voids of the blocks and the spaces therebetween, and means other than the helium in thermal conductive contact with the graphite for removing heat.

  15. Reactor apparatus

    DOEpatents

    Echtler, J. Paul

    1981-01-01

    A reactor apparatus for hydrocracking a polynuclear aromatic hydrocarbonaceous feedstock to produce lighter hydrocarbon fuels by contacting the hydrocarbonaceous feedstock with hydrogen in the presence of a molten metal halide catalyst.

  16. Chemical Reactors.

    ERIC Educational Resources Information Center

    Kenney, C. N.

    1980-01-01

    Describes a course, including content, reading list, and presentation on chemical reactors at Cambridge University, England. A brief comparison of chemical engineering education between the United States and England is also given. (JN)

  17. NEUTRONIC REACTOR

    DOEpatents

    Hurwitz, H. Jr.; Brooks, H.; Mannal, C.; Payne, J.H.; Luebke, E.A.

    1959-03-24

    A reactor of the heterogeneous, liquid cooled type is described. This reactor is comprised of a central region of a plurality of vertically disposed elongated tubes surrounded by a region of moderator material. The central region is comprised of a central core surrounded by a reflector region which is surrounded by a fast neutron absorber region, which in turn is surrounded by a slow neutron absorber region. Liquid sodium is used as the primary coolant and circulates through the core which contains the fuel elements. Control of the reactor is accomplished by varying the ability of the reflector region to reflect neutrons back into the core of the reactor. For this purpose the reflector is comprised of moderator and control elements having varying effects on reactivity, the control elements being arranged and actuated by groups to give regulation, shim, and safety control.

  18. Reactor Engineering

    NASA Astrophysics Data System (ADS)

    Lema, Juan M.; López, Carmen; Eibes, Gemma; Taboada-Puig, Roberto; Moreira, M. Teresa; Feijoo, Gumersindo

    In this chapter, the engineering aspects of processes catalyzed by peroxidases will be presented. In particular, a discussion of the existing technologies that utilize peroxidases for different purposes, such as the removal of recalcitrant compounds or the synthesis of polymers, is analyzed. In the first section, the essential variables controlling the process will be investigated, not only those that are common in any enzymatic system but also those specific to peroxidative reactions. Next, different reactor configurations and operational modes will be proposed, emphasizing their suitability and unsuitability for different systems. Finally, two specific reactors will be described in detail: enzymatic membrane reactors and biphasic reactors. These configurations are especially valuable for the treatment of xenobiotics with high and poor water solubility, respectively.

  19. NUCLEAR REACTOR

    DOEpatents

    Sherman, J.; Sharbaugh, J.E.; Fauth, W.L. Jr.; Palladino, N.J.; DeHuff, P.G.

    1962-10-23

    A nuclear reactor incorporating seed and blanket assemblies is designed. Means are provided for obtaining samples of the coolant from the blanket assemblies and for varying the flow of coolant through the blanket assemblies. (AEC)

  20. REACTOR SHIELD

    DOEpatents

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

    1959-02-17

    Radiation shield construction is described for a nuclear reactor. The shield is comprised of a plurality of steel plates arranged in parallel spaced relationship within a peripheral shell. Reactor coolant inlet tubes extend at right angles through the plates and baffles are arranged between the plates at right angles thereto and extend between the tubes to create a series of zigzag channels between the plates for the circulation of coolant fluid through the shield. The shield may be divided into two main sections; an inner section adjacent the reactor container and an outer section spaced therefrom. Coolant through the first section may be circulated at a faster rate than coolant circulated through the outer section since the area closest to the reactor container is at a higher temperature and is more radioactive. The two sections may have separate cooling systems to prevent the coolant in the outer section from mixing with the more contaminated coolant in the inner section.

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

    DOEpatents

    Rempe, Joy L [Idaho Falls, ID; Knudson, Darrell L [Firth, ID; Daw, Joshua E [Idaho Falls, ID; Condie, Keith G [Idaho Falls, ID; Stoots, Carl M [Idaho Falls, ID

    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.

  2. 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.

  3. Tritium-producing burnable absorber rod and its inherent safety features

    SciTech Connect

    Reid, B.D.; Lanning, D.D.

    2000-07-01

    Tritium-producing burnable absorber rods (TPBARs) used in the US Department of Energy's Commercial Light Water Reactor (CLWR) Project are designed to produce tritium when irradiated in a typical Westinghouse 17 x 17 pressurized water reactor (PWR). This paper provides an unclassified description of the design baseline for the TPBAR employed in the CLWR. TPBARs are similar in size and nuclear characteristics to standard, commercial PWR, stainless steel--clad burnable absorber rods. The exterior of the TPBAR is a stainless steel tube, {approximately}152 in. from tip to tip with a nominal outer diameter of 0.381 in. The internal components have been designed and selected to produce and retain tritium.

  4. 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

    2016-11-01

    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.

  5. NEUTRONIC REACTORS

    DOEpatents

    Vernon, H.C.

    1959-01-13

    A neutronic reactor of the heterogeneous, fluid cooled tvpe is described. The reactor is comprised of a pressure vessel containing the moderator and a plurality of vertically disposed channels extending in spaced relationship through the moderator. Fissionable fuel material is placed within the channels in spaced relationship thereto to permit circulation of the coolant fluid. Separate means are provided for cooling the moderator and for circulating a fluid coolant thru the channel elements to cool the fuel material.

  6. NUCLEAR REACTOR

    DOEpatents

    Anderson, C.R.

    1962-07-24

    A fluidized bed nuclear reactor and a method of operating such a reactor are described. In the design means are provided for flowing a liquid moderator upwardly through the center of a bed of pellets of a nentron-fissionable material at such a rate as to obtain particulate fluidization while constraining the lower pontion of the bed into a conical shape. A smooth circulation of particles rising in the center and falling at the outside of the bed is thereby established. (AEC)

  7. NUCLEAR REACTOR

    DOEpatents

    Breden, C.R.; Dietrich, J.R.

    1961-06-20

    A water-soluble non-volatile poison may be introduced into a reactor to nullify excess reactivity. The poison is removed by passing a side stream of the water containing the soluble poison to an evaporation chamber. The vapor phase is returned to the reactor to decrease the concentration of soluble poison and the liquid phase is returned to increase the concentration of soluble poison.

  8. NUCLEAR REACTOR

    DOEpatents

    Grebe, J.J.

    1959-07-14

    High temperature reactors which are uniquely adapted to serve as the heat source for nuclear pcwered rockets are described. The reactor is comprised essentially of an outer tubular heat resistant casing which provides the main coolant passageway to and away from the reactor core within the casing and in which the working fluid is preferably hydrogen or helium gas which is permitted to vaporize from a liquid storage tank. The reactor core has a generally spherical shape formed entirely of an active material comprised of fissile material and a moderator material which serves as a diluent. The active material is fabricated as a gas permeable porous material and is interlaced in a random manner with very small inter-connecting bores or capillary tubes through which the coolant gas may flow. The entire reactor is divided into successive sections along the direction of the temperature gradient or coolant flow, each section utilizing materials of construction which are most advantageous from a nuclear standpoint and which at the same time can withstand the operating temperature of that particular zone. This design results in a nuclear reactor characterized simultaneously by a minimum critiral size and mass and by the ability to heat a working fluid to an extremely high temperature.

  9. Effects of rod worth and drop speed on the BWR off-center rod drop accident

    SciTech Connect

    Cokinos, D.M.; Carew, J.F.

    1985-01-01

    BWR off-center RDA calculations have been performed for selected rod worths and drop speeds. While in all cases the peak fuel enthalpy was well below the 280 cal/g fuel criterion, a substantial sensitivity to control rod worth and rod drop speed was observed.

  10. Connexin 36 and Rod Bipolar Cell Independent Rod Pathways Drive Retinal Ganglion Cells and Optokinetic Reflexes

    PubMed Central

    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.

    2016-01-01

    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

  11. 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)

  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. A benchmark for investigating the radial dependence of C/E for control rod worths in large decoupled cores

    SciTech Connect

    Salvatores, M. ); Orechwa, Y. ); Collins, P.J. )

    1990-01-01

    The first physics measurements of a heterogeneous core on the critical assembly ZPPR-7 at ANL showed that the C/E ratios with ENDF/B data for the worths of the control rods in the outer bank were several percent higher than those at the inner bank positions. This radial variation in the C/E for the rod worths was further confirmed in the analysis of the large conventional core ZPPR-10, and again in the analysis of the large heterogeneous core series ZPPR-13. In the design of a power reactor, the number of control rods, and their disposition, are determined by calculations. Misprediction of the worth of the control rods can lead to serious economic penalties by restricting the operation of the core. Retrofitting a core to accommodate more worth will be costly and is likely to lead to a non-optimized core. This document provides a discussion of these calculations. 7 refs., 7 figs., 12 tabs.

  14. Research reactors - an overview

    SciTech Connect

    West, C.D.

    1997-03-01

    A broad overview of different types of research and type reactors is provided in this paper. Reactor designs and operating conditions are briefly described for four reactors. The reactor types described include swimming pool reactors, the High Flux Isotope Reactor, the Mark I TRIGA reactor, and the Advanced Neutron Source reactor. Emphasis in the descriptions is placed on safety-related features of the reactors. 7 refs., 7 figs., 2 tabs.

  15. Criticality of spent reactor fuel

    SciTech Connect

    Harris, D.R.

    1987-01-01

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

  16. 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)

  17. 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…

  18. 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…

  19. 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.

  20. Gas-cooled reactor for space power systems

    SciTech Connect

    Walter, C.E.; Pearson, J.S.

    1987-05-01

    Reactor characteristics based on extensive development work on the 500-MWt reactor for the Pluto nuclear ramjet are described for space power systems useful in the range of 2 to 20 MWe for operating times of 1 y. The modest pressure drop through the prismatic ceramic core is supported at the outlet end by a ceramic dome which also serves as a neutron reflector. Three core materials are considered which are useful at temperatures up to about 2000 K. Most of the calculations are based on a beryllium oxide with uranium dioxide core. Reactor control is accomplished by use of a burnable poison, a variable-leakage reflector, and internal control rods. Reactivity swings of 20% are obtained with a dozen internal boron-10 rods for the size cores studied. Criticality calculations were performed using the ALICE Monte Carlo code. The inherent high-temperature capability of the reactor design removes the reactor as a limiting condition on system performance. The low fuel inventories required, particularly for beryllium oxide reactors, make space power systems based on gas-cooled near-thermal reactors a lesser safeguard risk than those based on fast reactors.