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)

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.

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.

NUT SCREW MECHANISMS

DOEpatents

Glass, J.A.F.

1958-07-01

A reactor control mechanism is described wherein the control is achieved by the partial or total withdrawal of the fissile material which is in the form of a fuel rod. The fuel rod is designed to be raised and lowered from the reactor core area by means of two concentric ball nut and screw assemblies that may telescope one within the other. These screw mechanisms are connected through a magnetic clutch to a speed reduction gear and an accurately controllable prime motive source. With the clutch energized, the fuel rod may be moved into the reactor core area, and fine adjustments may be made through the reduction gearing. However, in the event of a power failure or an emergency signal, the magnetic clutch will become deenergized, and the fuel rod will drop out of the core area by the force of gravity, thus shutting down the operation of the reactor.

Summary of the thermal evaluation of LWBR (LWBR Development Program)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lerner, S.; McWilliams, K.D.; Stout, J.W.

1980-03-01

This report describes the thermal evaluation of the core for the Shippingport Light Water Breeder Reactor. This core contains unique thermal-hydraulic features such as (1) close rod-to-rod proximity, (2) an open-lattice array of fuel rods with two different diameters and rod-to-rod spacings in the same flow region, (3) triplate orifices located at both the entrance and exit of fuel modules and (4) a hydraulically-balanced movable-fuel system coupled with (5) axial-and-radial fuel zoning for reactivity control. Performance studies used reactor thermal principles such as the hot-and-nominal channel concept and related nuclear/engineering design allowances. These were applied to models of three-dimensional roddedmore » arrays comprising the core fuel regions.« less

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.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rowsell, David Leon

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

An Overview of Reactor Concepts, a Survey of Reactor Designs.

DTIC Science & Technology

1985-02-01

may be very different. HTGRs may use highly enriched uranium, thereby yielding better fuel economy and a reduc- tion of the actual core size for a...specific power level. The HTGR core may have fuel and control rods placed in graphite arrays similar to PWR core con- figuration, or they may have fuel ...rods are pulled out. A Peach Bottom core design is another HTGR design. This design is featured by the fuel pin’s ability to purge itself of fission

LOGIC CIRCUIT

DOEpatents

Strong, G.H.; Faught, M.L.

1963-12-24

A device for safety rod counting in a nuclear reactor is described. A Wheatstone bridge circuit is adapted to prevent de-energizing the hopper coils of a ball backup system if safety rods, sufficient in total control effect, properly enter the reactor core to effect shut down. A plurality of resistances form one arm of the bridge, each resistance being associated with a particular safety rod and weighted in value according to the control effect of the particular safety rod. Switching means are used to switch each of the resistances in and out of the bridge circuit responsive to the presence of a particular safety rod in its effective position in the reactor core and responsive to the attainment of a predetermined velocity by a particular safety rod enroute to its effective position. The bridge is unbalanced in one direction during normal reactor operation prior to the generation of a scram signal and the switching means and resistances are adapted to unbalance the bridge in the opposite direction if the safety rods produce a predetermined amount of control effect in response to the scram signal. The bridge unbalance reversal is then utilized to prevent the actuation of the ball backup system, or, conversely, a failure of the safety rods to produce the predetermined effect produces no unbalance reversal and the ball backup system is actuated. (AEC)

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.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

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 amore » 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.« less

POSITIONING MEANS FOR THE CONTROL ROD IN A NUCLEAR REACTOR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bell, F.R.

1963-09-26

An electric motor and transmission means for adjusting the position of the control rod in relation to the reactor core to control the level of activity in either sense and a lock that is retained in a locking condition by a longitudinal force derived electrically from the motor but arranged to be released when the electrical force is removed to allow the control rod in an emergency to drop into its shut-down position are described. (Gmelin Inst.)

Adaptive control method for core power control in TRIGA Mark II reactor

NASA Astrophysics Data System (ADS)

Sabri Minhat, Mohd; Selamat, Hazlina; Subha, Nurul Adilla Mohd

2018-01-01

The 1MWth Reactor TRIGA PUSPATI (RTP) Mark II type has undergone more than 35 years of operation. The existing core power control uses feedback control algorithm (FCA). It is challenging to keep the core power stable at the desired value within acceptable error bands to meet the safety demand of RTP due to the sensitivity of nuclear research reactor operation. Currently, the system is not satisfied with power tracking performance and can be improved. Therefore, a new design core power control is very important to improve the current performance in tracking and regulate reactor power by control the movement of control rods. In this paper, the adaptive controller and focus on Model Reference Adaptive Control (MRAC) and Self-Tuning Control (STC) were applied to the control of the core power. The model for core power control was based on mathematical models of the reactor core, adaptive controller model, and control rods selection programming. The mathematical models of the reactor core were based on point kinetics model, thermal hydraulic models, and reactivity models. The adaptive control model was presented using Lyapunov method to ensure stable close loop system and STC Generalised Minimum Variance (GMV) Controller was not necessary to know the exact plant transfer function in designing the core power control. The performance between proposed adaptive control and FCA will be compared via computer simulation and analysed the simulation results manifest the effectiveness and the good performance of the proposed control method for core power control.

Neutronic safety parameters and transient analyses for Poland's MARIA research reactor.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bretscher, M. M.; Hanan, N. A.; Matos, J. E.

1999-09-27

Reactor kinetic parameters, reactivity feedback coefficients, and control rod reactivity worths have been calculated for the MARIA Research Reactor (Swierk, Poland) for M6-type fuel assemblies with {sup 235}U enrichments of 80% and 19.7%. Kinetic parameters were evaluated for family-dependent effective delayed neutron fractions, decay constants, and prompt neutron lifetimes and neutron generation times. Reactivity feedback coefficients were determined for fuel Doppler coefficients, coolant (H{sub 2}O) void and temperature coefficients, and for in-core and ex-core beryllium temperature coefficients. Total and differential control rod worths and safety rod worths were calculated for each fuel type. These parameters were used to calculate genericmore » transients for fast and slow reactivity insertions with both HEU and LEU fuels. The analyses show that the HEU and LEU cores have very similar responses to these transients.« less

Method for depleting BWRs using optimal control rod patterns

DOE Office of Scientific and Technical Information (OSTI.GOV)

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 nucleonicsmore » calculations.« less

Compact Hybrid Laser Rod and Laser System

NASA Technical Reports Server (NTRS)

Pierrottet, Diego F. (Inventor); Busch, George E. (Inventor); Amzajerdian, Farzin (Inventor)

2017-01-01

A hybrid fiber rod includes a fiber core and inner and outer cladding layers. The core is doped with an active element. The inner cladding layer surrounds the core, and has a refractive index substantially equal to that of the core. The outer cladding layer surrounds the inner cladding layer, and has a refractive index less than that of the core and inner cladding layer. The core length is about 30 to 2000 times the core diameter. A hybrid fiber rod laser system includes an oscillator laser, modulating device, the rod, and pump laser diode(s) energizing the rod from opposite ends. The rod acts as a waveguide for pump radiation but allows for free-space propagation of laser radiation. The rod may be used in a laser resonator. The core length is less than about twice the Rayleigh range. Degradation from single-mode to multi-mode beam propagation is thus avoided.

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.

Validation of updated neutronic calculation models proposed for Atucha-II PHWR. Part I: Benchmark comparisons of WIMS-D5 and DRAGON cell and control rod parameters with MCNP5

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mollerach, R.; Leszczynski, F.; Fink, J.

2006-07-01

In 2005 the Argentine Government took the decision to complete the construction of the Atucha-II nuclear power plant, which has been progressing slowly during the last ten years. Atucha-II is a 745 MWe nuclear station moderated and cooled with heavy water, of German (Siemens) design located in Argentina. It has a pressure-vessel design with 451 vertical coolant channels, and the fuel assemblies (FA) are clusters of 37 natural UO{sub 2} rods with an active length of 530 cm. For the reactor physics area, a revision and update calculation methods and models (cell, supercell and reactor) was recently carried out coveringmore » cell, supercell (control rod) and core calculations. As a validation of the new models some benchmark comparisons were done with Monte Carlo calculations with MCNP5. This paper presents comparisons of cell and supercell benchmark problems based on a slightly idealized model of the Atucha-I core obtained with the WIMS-D5 and DRAGON codes with MCNP5 results. The Atucha-I core was selected because it is smaller, similar from a neutronic point of view, and more symmetric than Atucha-II Cell parameters compared include cell k-infinity, relative power levels of the different rings of fuel rods, and some two-group macroscopic cross sections. Supercell comparisons include supercell k-infinity changes due to the control rods (tubes) of steel and hafnium. (authors)« less

10 CFR 55.41 - Written examination: Operators.

Code of Federal Regulations, 2011 CFR

2011-01-01

... coefficients, and poison effects. (2) General design features of the core, including core structure, fuel elements, control rods, core instrumentation, and coolant flow. (3) Mechanical components and design... changes, and operating limitations and reasons for these operating characteristics. (6) Design, components...

10 CFR 55.41 - Written examination: Operators.

Code of Federal Regulations, 2010 CFR

2010-01-01

... coefficients, and poison effects. (2) General design features of the core, including core structure, fuel elements, control rods, core instrumentation, and coolant flow. (3) Mechanical components and design... changes, and operating limitations and reasons for these operating characteristics. (6) Design, components...

10 CFR 55.41 - Written examination: Operators.

Code of Federal Regulations, 2012 CFR

2012-01-01

... coefficients, and poison effects. (2) General design features of the core, including core structure, fuel elements, control rods, core instrumentation, and coolant flow. (3) Mechanical components and design... changes, and operating limitations and reasons for these operating characteristics. (6) Design, components...

Multi-function magnetic jack control drive mechanism

DOEpatents

Bollinger, L.R.; Crawford, D.C.

1983-10-06

A multi-function magnetic jack control drive mechanism for controlling a nuclear reactor is provided. The mechanism includes an elongate pressure housing in which a plurality of closely spaced drive rods are located. Each drive rod is connected to a rod which is insertable in the reactor core. An electromechanical stationary latch device is provided which is actuatable to hold each drive rod stationary with respect to the pressure housing. An electromechanical movable latch device is also provided for each one of the drive rods. Each movable latch device is provided with a base and is actuatable to hold a respective drive rod stationary with respect to the base. An electromechanical lift device is further provided for each base which is actuatable for moving a respective base longitudinally along the pressure housing. In this manner, one or more drive rods can be moved in the pressure housing by sequentially and repetitively operating the electromechanical devices. Preferably, each latch device includes a pair of opposed latches which grip teeth located on the respective drive rod. Two, three, or four drive rods can be located symmetrically about the longitudinal axis of the pressure housing.

Multi-function magnetic jack control drive mechanism

DOEpatents

Bollinger, Lawrence R.; Crawford, Donald C.

1986-01-01

A multi-function magnetic jack control drive mechanism for controlling a nuclear reactor is provided. The mechanism includes an elongate pressure housing in which a plurality of closely spaced drive rods are located. Each drive rod is connected to a rod which is insertable in the reactor core. An electromechanical stationary latch device is provided which is actuatable to hold each drive rod stationary with respect to the pressure housing. An electromechanical movable latch device is also provided for each one of the drive rods. Each movable latch device is provided with a base and is actuatable to hold a respective drive rod stationary with respect to the base. An electromechanical lift device is further provided for each base which is actuatable for moving a respective base longitudinally along the pressure housing. In this manner, one or more drive rods can be moved in the pressure housing by sequentially and repetitively operating the electromechanical devices. Preferably, each latch device includes a pair of opposed latches which grip teeth located on the respective drive rod. Two, three, or four drive rods can be located symmetrically about the longitudinal axis of the pressure housing.

JPRS Report Science and Technology, Japan: Atomic Energy Society 1989 Annual Meeting.

DTIC Science & Technology

1989-10-13

Control Rod Hole in VHTRC-1 Core [F, Akino, T, Yamane, et al.] ,,, 5 Measurement of MEU [Medium Enriched Uranium ] Fuel Element Characteristics in...K. Yoshida, K. Kobayashi, I. Kimura , C. Yamanaka, and S. Nakai, Laser Laboratory,, Osaka University. Nuclear Reactor Laboratory, Kyoto University...1 core loaded with 278 fuel rods (4 percent enriched uranium ). The PNS target was placed at the back center of the 1/2 assembly on the fixed side

Thermionic switched self-actuating reactor shutdown system

DOEpatents

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

1989-01-01

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

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

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 themore » 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.« less

MOX fuel arrangement for nuclear core

DOEpatents

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

1998-10-13

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

Mox fuel arrangement for nuclear core

DOEpatents

Kantrowitz, Mark L.; Rosenstein, Richard G.

2001-05-15

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

MOX fuel arrangement for nuclear core

DOEpatents

Kantrowitz, Mark L.; Rosenstein, Richard G.

2001-07-17

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

MOX fuel arrangement for nuclear core

DOEpatents

Kantrowitz, Mark L.; Rosenstein, Richard G.

1998-01-01

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

Braided reinforced composite rods for the internal reinforcement of concrete

NASA Astrophysics Data System (ADS)

Gonilho Pereira, C.; Fangueiro, R.; Jalali, S.; Araujo, M.; Marques, P.

2008-05-01

This paper reports on the development of braided reinforced composite rods as a substitute for the steel reinforcement in concrete. The research work aims at understanding the mechanical behaviour of core-reinforced braided fabrics and braided reinforced composite rods, namely concerning the influence of the braiding angle, the type of core reinforcement fibre, and preloading and postloading conditions. The core-reinforced braided fabrics were made from polyester fibres for producing braided structures, and E-glass, carbon, HT polyethylene, and sisal fibres were used for the core reinforcement. The braided reinforced composite rods were obtained by impregnating the core-reinforced braided fabric with a vinyl ester resin. The preloading of the core-reinforced braided fabrics and the postloading of the braided reinforced composite rods were performed in three and two stages, respectively. The results of tensile tests carried out on different samples of core-reinforced braided fabrics are presented and discussed. The tensile and bending properties of the braided reinforced composite rods have been evaluated, and the results obtained are presented, discussed, and compared with those of conventional materials, such as steel.

Anticipatory control of xenon in a pressurized water reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Impink, A.J. Jr.

1987-02-10

A method is described for automatically dampening xenon-135 spatial transients in the core of a pressurized water reactor having control rods which regulate reactor power level, comprising the steps of: measuring the neutron flu in the reactor core at a plurality of axially spaced locations on a real-time, on-line basis; repetitively generating from the neutron flux measurements, on a point-by-point basis, signals representative of the current axial distribution of xenon-135, and signals representative of the current rate of change of the axial distribution of xenon-135; generating from the xenon-135 distribution signals and the rate of change of xenon distribution signals,more » control signals for reducing the xenon transients; and positioning the control rods as a function of the control signals to dampen the xenon-135 spatial transients.« less

Benchmark of Atucha-2 PHWR RELAP5-3D control rod model by Monte Carlo MCNP5 core calculation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pecchia, M.; D'Auria, F.; Mazzantini, O.

2012-07-01

Atucha-2 is a Siemens-designed PHWR reactor under construction in the Republic of Argentina. Its geometrical complexity and peculiarities require the adoption of advanced Monte Carlo codes for performing realistic neutronic simulations. Therefore core models of Atucha-2 PHWR were developed using MCNP5. In this work a methodology was set up to collect the flux in the hexagonal mesh by which the Atucha-2 core is represented. The scope of this activity is to evaluate the effect of obliquely inserted control rod on neutron flux in order to validate the RELAP5-3D{sup C}/NESTLE three dimensional neutron kinetic coupled thermal-hydraulic model, applied by GRNSPG/UNIPI formore » performing selected transients of Chapter 15 FSAR of Atucha-2. (authors)« less

10 CFR 55.41 - Written examination: Operators.

Code of Federal Regulations, 2013 CFR

2013-01-01

... elements, control rods, core instrumentation, and coolant flow. (3) Mechanical components and design..., and functions of reactivity control mechanisms and instrumentation. (7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and...

10 CFR 55.41 - Written examination: Operators.

Code of Federal Regulations, 2014 CFR

2014-01-01

... elements, control rods, core instrumentation, and coolant flow. (3) Mechanical components and design..., and functions of reactivity control mechanisms and instrumentation. (7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and...

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.

Fast-spectrum space-power-reactor concepts using boron control devices

NASA Technical Reports Server (NTRS)

Mayo, W.

1973-01-01

Several fast-spectrum space power reactor concepts that use boron carbide control devices were examined to determine the neutronic feasibility of the designs. The designs considered were (1) a 199-fuel-pin, 12-poison-reflector-control-drum reactor; (2) a 232-fuel-pin reactor with 12 reflector drums and three in-core control rods; (3) a 337-fuel-pin design with 12 incore control rods; and a 181-fuel-pin design with six drums closely coupled to the core to increase reactivity per drum. Adequate reactivity control and excess reactivity could be obtained for each concept, and the goals of 50,000 hours at 2.17 thermal megawatts with a lithium-7 coolant outlet temperature of 1222 K could be met without exceeding the 1-percent-clad-creep criterion. Heating rates in the boron carbide were calculated, but a heat transfer analysis was not done.

75 FR 37876 - Buy America Waiver Notification

Federal Register 2010, 2011, 2012, 2013, 2014

2010-06-30

... hollow core threaded share anchor rods in construction of Federal-aid project X-STP-1525 (004) in Oregon... is appropriate for the use of non- domestic 1'' diameter hollow core threaded share anchor rods for... to issue a waiver for the 1'' diameter hollow core threaded share anchor rods ( http://www.fhwa.dot...

Development of a three-dimensional transient code for reactivity-initiated events of BWRs (boiling water reactors) - Models and code verifications

DOE Office of Scientific and Technical Information (OSTI.GOV)

Uematsu, Hitoshi; Yamamoto, Toru; Izutsu, Sadayuki

1990-06-01

A reactivity-initiated event is a design-basis accident for the safety analysis of boiling water reactors. It is defined as a rapid transient of reactor power caused by a reactivity insertion of over $1.0 due to a postulated drop or abnormal withdrawal of the control rod from the core. Strong space-dependent feedback effects are associated with the local power increase due to control rod movement. A realistic treatment of the core status in a transient by a code with a detailed core model is recommended in evaluating this event. A three-dimensional transient code, ARIES, has been developed to meet this need.more » The code simulates the event with three-dimensional neutronics, coupled with multichannel thermal hydraulics, based on a nonequilibrium separated flow model. The experimental data obtained in reactivity accident tests performed with the SPERT III-E core are used to verify the entire code, including thermal-hydraulic models.« less

Nanopipes in gallium nitride nanowires and rods.

PubMed

Jacobs, Benjamin W; Crimp, Martin A; McElroy, Kaylee; Ayres, Virginia M

2008-12-01

Gallium nitride nanowires and rods synthesized by a catalyst-free vapor-solid growth method were analyzed with cross section high-resolution transmission electron microscopy. The cross section studies revealed hollow core screw dislocations, or nanopipes, in the nanowires and rods. The hollow cores were located at or near the center of the nanowires and rods, along the axis of a screw dislocation. The formation of the hollow cores is consistent with effect of screw dislocations with giant Burgers vector predicted by Frank.

Methodology of the Westinghouse dynamic rod worth measurement technique

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chao, Y.A.; Chapman, D.M.; Easter, M.E.

1992-01-01

During zero-power physics testing, plant operations personnel use one of various techniques to measure the reactivity worth of the control rods to confirm shutdown margin. A simple and fast procedure for measuring rod worths called dynamic rod worth measurement (DRWM) has been developed at Westinghouse. This procedure was tested at the recent startups of Point Beach Nuclear Power Plant Unit 1 cycle 20 and Unit 2 cycle 18. The results of these tests show that DRWM measures rod worths with accuracy comparable to that of both boron dilution and rod bank exchange measurements. The DRWM procedure is a fast processmore » of measuring the reactivity worth of individual banks by inserting and withdrawing the bank continuously at the maximum stepping speed without changing the boron concentration and recording the signals of the ex-core detectors.« less

Experimental results from the VENUS-F critical reference state for the GUINEVERE accelerator driven system project

DOE Office of Scientific and Technical Information (OSTI.GOV)

Uyttenhove, W.; Baeten, P.; Ban, G.

The GUINEVERE (Generation of Uninterrupted Intense Neutron pulses at the lead Venus Reactor) project was launched in 2006 within the framework of FP6 EUROTRANS in order to validate on-line reactivity monitoring and subcriticality level determination in Accelerator Driven Systems. Therefore the VENUS reactor at SCK.CEN in Mol (Belgium) was modified towards a fast core (VENUS-F) and coupled to the GENEPI-3C accelerator built by CNRS The accelerator can operate in both continuous and pulsed mode. The VENUS-F core is loaded with enriched Uranium and reflected with solid lead. A well-chosen critical reference state is indispensable for the validation of the on-linemore » subcriticality monitoring methodology. Moreover a benchmarking tool is required for nuclear data research and code validation. In this paper the design and the importance of the critical reference state for the GUINEVERE project are motivated. The results of the first experimental phase on the critical core are presented. The control rods worth is determined by the rod drop technique and the application of the Modified Source Multiplication (MSM) method allows the determination of the worth of the safety rods. The results are implemented in the VENUS-F core certificate for full exploitation of the critical core. (authors)« less

Development and validation of a low-frequency modeling code for high-moment transmitter rod antennas

NASA Astrophysics Data System (ADS)

Jordan, Jared Williams; Sternberg, Ben K.; Dvorak, Steven L.

2009-12-01

The goal of this research is to develop and validate a low-frequency modeling code for high-moment transmitter rod antennas to aid in the design of future low-frequency TX antennas with high magnetic moments. To accomplish this goal, a quasi-static modeling algorithm was developed to simulate finite-length, permeable-core, rod antennas. This quasi-static analysis is applicable for low frequencies where eddy currents are negligible, and it can handle solid or hollow cores with winding insulation thickness between the antenna's windings and its core. The theory was programmed in Matlab, and the modeling code has the ability to predict the TX antenna's gain, maximum magnetic moment, saturation current, series inductance, and core series loss resistance, provided the user enters the corresponding complex permeability for the desired core magnetic flux density. In order to utilize the linear modeling code to model the effects of nonlinear core materials, it is necessary to use the correct complex permeability for a specific core magnetic flux density. In order to test the modeling code, we demonstrated that it can accurately predict changes in the electrical parameters associated with variations in the rod length and the core thickness for antennas made out of low carbon steel wire. These tests demonstrate that the modeling code was successful in predicting the changes in the rod antenna characteristics under high-current nonlinear conditions due to changes in the physical dimensions of the rod provided that the flux density in the core was held constant in order to keep the complex permeability from changing.

The use of modified scaling factors in the design of high-power, non-linear, transmitting rod-core antennas

NASA Astrophysics Data System (ADS)

Jordan, Jared Williams; Dvorak, Steven L.; Sternberg, Ben K.

2010-10-01

In this paper, we develop a technique for designing high-power, non-linear, transmitting rod-core antennas by using simple modified scale factors rather than running labor-intensive numerical models. By using modified scale factors, a designer can predict changes in magnetic moment, inductance, core series loss resistance, etc. We define modified scale factors as the case when all physical dimensions of the rod antenna are scaled by p, except for the cross-sectional area of the individual wires or strips that are used to construct the core. This allows one to make measurements on a scaled-down version of the rod antenna using the same core material that will be used in the final antenna design. The modified scale factors were derived from prolate spheroidal analytical expressions for a finite-length rod antenna and were verified with experimental results. The modified scaling factors can only be used if the magnetic flux densities within the two scaled cores are the same. With the magnetic flux density constant, the two scaled cores will operate with the same complex permeability, thus changing the non-linear problem to a quasi-linear problem. We also demonstrate that by holding the number of turns times the drive current constant, while changing the number of turns, the inductance and core series loss resistance change by the number of turns squared. Experimental measurements were made on rod cores made from varying diameters of black oxide, low carbon steel wires and different widths of Metglas foil. Furthermore, we demonstrate that the modified scale factors work even in the presence of eddy currents within the core material.

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.

Chemical State Mapping of Degraded B4C Control Rod Investigated with Soft X-ray Emission Spectrometer in Electron Probe Micro-analysis.

PubMed

Kasada, R; Ha, Y; Higuchi, T; Sakamoto, K

2016-05-10

B4C is widely used as control rods in light water reactors, such as the Fukushima Daiichi nuclear power plant, because it shows excellent neutron absorption and has a high melting point. However, B4C can melt at lower temperatures owing to eutectic interactions with stainless steel and can even evaporate by reacting with high-temperature steam under severe accident conditions. To reduce the risk of recriticality, a precise understanding of the location and chemical state of B in the melt core is necessary. Here we show that a novel soft X-ray emission spectrometer in electron probe microanalysis can help to obtain a chemical state map of B in a modeled control rod after a high-temperature steam oxidation test.

TMI-2 (Three Mile Island Unit 2) core region defueling

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rodabaugh, J.M.; Cowser, D.K.

1988-01-01

In July of 1982, a video camera was inserted into the Three Mile Island Unit 2 reactor vessel providing the first visual evidence of core damage. This inspection, and numerous subsequent data acquisition tasks, revealed a central void /approx/1.5 m (5 ft) deep. This void region was surrounded by partial length fuel assemblies and ringed on the periphery by /approx/40 full-length, but partial cross-section, fuel assemblies. All of the original 177 fuel assemblies exhibited signs of damage. The bottom of the void cavity was covered with a bed of granular rubble, fuel assembly upper end fittings, control rod spiders, fuelmore » rod fragments, and fuel pellets. It was obvious that the normal plant refueling system not suitable for removing the damaged core. A new system of defueling tools and equipment was necessary to perform this task. Design of the new system was started immediately, followed by >1 yr of fabrication. Delivery and checkout of the defueling system occurred in mid-1985. Actual defueling was initiated in late 1985 with removal of the debris bed at the bottom of the core void. Obstructions to the debris, such as end fittings and fuel rod fragments ere removed first; then /approx/23,000 kg (50,000lb) of granular debris was quickly loaded into canisters. Core region defueling was completed in late 1987, /approx/2 yr after it was initiated.« less

Nuclear core and fuel assemblies

DOEpatents

Downs, Robert E.

1981-01-01

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

Flux harmonics in large SFR cores in relation with core characteristics such as power peaks

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rimpault, G.; Buiron, L.; Fontaine, B.

2013-07-01

Designing future Sodium Fast Reactors (SFR) requires enhancing their operational performance and reducing the probability to go into core disruption. As a consequence of these constraints, these novel reactors exhibit rather unusual features compared to past designs. The cores are much larger with rather flat shape. The consequences of that shape on the core characteristics deserve to be studied. The approach taken in this paper is to calculate the eigenvalue associated to the first harmonic and its associated flux. It is demonstrated that these values are linked to some core features, in particular, those sensitive to spatial effects such asmore » power peaks induced by the movement of control rods. The uncertainty associated to these characteristics is being tentatively studied and guidelines for further studied are being identified. In the development strategy of these new SFR designs, a first demonstration plant of limited installed power (around 1500 MWth) will have to be built first. Identifying the possibility of going later to higher power plants (around 3600 MWth) without facing new challenges is an important criterion for designing such a plant. That strategy is being studied, in this paper, focusing on some rather frequent initiator such as the inadvertent control rod withdrawal for different core sizes with the help of the perturbation theory and the flux harmonics. (authors)« less

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.

Thermal control of high energy nuclear waste, space option. [mathematical models

NASA Technical Reports Server (NTRS)

Peoples, J. A.

1979-01-01

Problems related to the temperature and packaging of nuclear waste material for disposal in space are explored. An approach is suggested for solving both problems with emphasis on high energy density waste material. A passive cooling concept is presented which utilized conduction rods that penetrate the inner core. Data are presented to illustrate the effectiveness of the rods and the limit of their capability. A computerized thermal model is discussed and developed for the cooling concept.

Results of a Neutronic Simulation of HTR-Proteus Core 4.2 using PEBBED and other INL Reactor Physics Tools: FY-09 Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hans D. Gougar

The Idaho National Laboratory’s deterministic neutronics analysis codes and methods were applied to the computation of the core multiplication factor of the HTR-Proteus pebble bed reactor critical facility. A combination of unit cell calculations (COMBINE-PEBDAN), 1-D discrete ordinates transport (SCAMP), and nodal diffusion calculations (PEBBED) were employed to yield keff and flux profiles. Preliminary results indicate that these tools, as currently configured and used, do not yield satisfactory estimates of keff. If control rods are not modeled, these methods can deliver much better agreement with experimental core eigenvalues which suggests that development efforts should focus on modeling control rod andmore » other absorber regions. Under some assumptions and in 1D subcore analyses, diffusion theory agrees well with transport. This suggests that developments in specific areas can produce a viable core simulation approach. Some corrections have been identified and can be further developed, specifically: treatment of the upper void region, treatment of inter-pebble streaming, and explicit (multiscale) transport modeling of TRISO fuel particles as a first step in cross section generation. Until corrections are made that yield better agreement with experiment, conclusions from core design and burnup analyses should be regarded as qualitative and not benchmark quality.« less

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.

Electrically operated magnetic switch designed to display reduced leakage inductance

DOEpatents

Cook, Edward G.

1994-01-01

An electrically operated magnetic switch is disclosed herein for use in opening and closing a circuit between two terminals depending upon the voltage across these terminals. The switch so disclosed is comprised of a ferrite core in the shape of a toroid having opposing ends and opposite inner and outer sides and an arrangement of electrically conductive components defining at least one current flow path which makes a number of turns around the core. This arrangement of components includes a first plurality of electrically conducive rigid rods parallel with and located outside the outer side of the core and a second plurality of electrically conductive rigid rods parallel with and located inside the inner side of the core. The arrangement also includes means for electrically connecting these rods together so that the define the current flow path. In one embodiment, this latter means uses rigid cross-tab means. In another, preferred embodiment, printed circuits on rigid dielectric substrates located on opposite ends of the core are utilized to interconnect the rods together.

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.

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.

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.

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.

Estimation of free carrier concentrations in high-quality heavily doped GaN:Si micro-rods by photoluminescence and Raman spectroscopy

NASA Astrophysics Data System (ADS)

Mohajerani, M. S.; Khachadorian, S.; Nenstiel, C.; Schimpke, T.; Avramescu, A.; Strassburg, M.; Hoffmann, A.; Waag, A.

2016-03-01

The controlled growth of highly n-doped GaN micro rods is one of the major challenges in the fabrication of recently developed three-dimensional (3D) core-shell light emitting diodes (LEDs). In such structures with a large active area, higher electrical conductivity is needed to achieve higher current density. In this contribution, we introduce high quality heavily-doped GaN:Si micro-rods which are key elements of the newly developed 3D core-shell LEDs. These structures were grown by metal-organic vapor phase epitaxy (MOVPE) using selective area growth (SAG). We employed spatially resolved micro-Raman and micro-photoluminescence (PL) in order to directly determine a free-carrier concentration profile in individual GaN micro-rods. By Raman spectroscopy, we analyze the low-frequency branch of the longitudinal optical (LO)-phonon-plasmon coupled modes and estimate free carrier concentrations from ≍ 2.4 × 1019 cm-3 up to ≍ 1.5 × 1020 cm-3. Furthermore, free carrier concentrations are determined by estimating Fermi energy level from the near band edge emission measured by low-temperature PL. The results from both methods reveal a good consistency.

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.

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)

Results of the Simulation of the HTR-Proteus Core 4.2 Using PEBBED-COMBINE: FY10 Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hans Gougar

2010-07-01

ABSTRACT The Idaho National Laboratory’s deterministic neutronics analysis codes and methods were applied to the computation of the core multiplication factor of the HTR-Proteus pebble bed reactor critical facility. This report is a follow-on to INL/EXT-09-16620 in which the same calculation was performed but using earlier versions of the codes and less developed methods. In that report, results indicated that the cross sections generated using COMBINE-7.0 did not yield satisfactory estimates of keff. It was concluded in the report that the modeling of control rods was not satisfactory. In the past year, improvements to the homogenization capability in COMBINE havemore » enabled the explicit modeling of TRIS particles, pebbles, and heterogeneous core zones including control rod regions using a new multi-scale version of COMBINE in which the 1-dimensional discrete ordinate transport code ANISN has been integrated. The new COMBINE is shown to yield benchmark quality results for pebble unit cell models, the first step in preparing few-group diffusion parameters for core simulations. In this report, the full critical core is modeled once again but with cross sections generated using the capabilities and physics of the improved COMBINE code. The new PEBBED-COMBINE model enables the exact modeling of the pebbles and control rod region along with better approximation to structures in the reflector. Initial results for the core multiplication factor indicate significant improvement in the INL’s tools for modeling the neutronic properties of a pebble bed reactor. Errors on the order of 1.6-2.5% in keff are obtained; a significant improvement over the 5-6% error observed in the earlier This is acceptable for a code system and model in the early stages of development but still too high for a production code. Analysis of a simpler core model indicates an over-prediction of the flux in the low end of the thermal spectrum. Causes of this discrepancy are under investigation. New homogenization techniques and assumptions were used in this analysis and as such, they require further confirmation and validation. Further refinement and review of the complex Proteus core model are likely to reduce the errors even further.« less

Fast reactor core concepts to improve transmutation efficiency

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fujimura, Koji; Kawashima, Katsuyuki; Itooka, Satoshi

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.

Electrically operated magnetic switch designed to display reduced leakage inductance

DOEpatents

Cook, E.G.

1994-05-10

An electrically operated magnetic switch is disclosed herein for use in opening and closing a circuit between two terminals depending upon the voltage across these terminals. The switch so disclosed is comprised of a ferrite core in the shape of a toroid having opposing ends and opposite inner and outer sides and an arrangement of electrically conductive components defining at least one current flow path which makes a number of turns around the core. This arrangement of components includes a first plurality of electrically conducive rigid rods parallel with and located outside the outer side of the core and a second plurality of electrically conductive rigid rods parallel with and located inside the inner side of the core. The arrangement also includes means for electrically connecting these rods together so that the define the current flow path. In one embodiment, this latter means uses rigid cross-tab means. In another, preferred embodiment, printed circuits on rigid dielectric substrates located on opposite ends of the core are utilized to interconnect the rods together. 10 figures.

Self powered neutron detectors as in-core detectors for Sodium-cooled Fast Reactors

NASA Astrophysics Data System (ADS)

Verma, V.; Barbot, L.; Filliatre, P.; Hellesen, C.; Jammes, C.; Svärd, S. Jacobsson

2017-07-01

Neutron flux monitoring system forms an integral part of the design of a Generation IV sodium cooled fast reactor. Diverse possibilities of detector system installation must be studied for various locations in the reactor vessel in order to detect any perturbations in the core. Results from a previous paper indicated that it is possible to detect changes in neutron source distribution initiated by an inadvertent withdrawal of outer control rod with in-vessel fission chambers located azimuthally around the core. It is, however, not possible to follow inner control rod withdrawal and precisely know the location of the perturbation in the core. Hence the use of complimentary in-core detectors coupled with the peripheral fission chambers is proposed to enable robust core monitoring across the radial direction. In this paper, we assess the feasibility of using self-powered neutron detectors (SPNDs) as in-core detectors in fast reactors for detecting local changes in the power distribution when the reactor is operated at nominal power. We study the neutron and gamma contributions to the total output current of the detector modelled with Platinum as the emitter material. It is shown that this SPND placed in an SFR-like environment would give a sufficiently measurable prompt neutron induced current of the order of 600 nA/m. The corresponding induced current in the connecting cable is two orders of magnitude lower and can be neglected. This means that the SPND can follow in-core power fluctuations. This validates the operability of an SPND in an SFR-like environment.

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.

Stair-rod dislocation cores acting as one-dimensional charge channels in GaAs nanowires

NASA Astrophysics Data System (ADS)

Bologna, Nicolas; Agrawal, Piyush; Campanini, Marco; Knödler, Moritz; Rossell, Marta D.; Erni, Rolf; Passerone, Daniele

2018-01-01

Aberration-corrected scanning transmission electron microscopy and density-functional theory calculations have been used to investigate the atomic and electronic structure of stair-rod dislocations connected via stacking faults in GaAs nanowires. At the apexes, two distinct dislocation cores consisting of single-column pairs of either gallium or arsenic were identified. Ab initio calculations reveal an overall reduction in the energy gap with the development of two bands of filled and empty localized states at the edges of valence and conduction bands in the Ga core and in the As core, respectively. Our results suggest the behavior of stair-rod dislocations along the nanowire as one-dimensional charge channels, which could host free carriers upon appropriate doping.

Guiding and amplification properties of rod-type photonic crystal fibers with sectioned core doping

NASA Astrophysics Data System (ADS)

Selleri, S.; Poli, F.; Passaro, D.; Cucinotta, A.; Lægsgaard, J.; Broeng, J.

2009-05-01

Rod-type photonic crystal fibers are large mode area double-cladding fibers with an outer diameter of few millimeters which can provide important advantages for high-power lasers and amplifiers. Numerical studies have recently demonstrated the guidance of higher-order modes in these fibers, which can worsen the output beam quality of lasers and amplifiers. In the present analysis a sectioned core doping has been proposed for Ybdoped rod-type photonic crystal fibers, with the aim to improve the higher-order mode suppression. A full-vector modal solver based on the finite element method has been applied to properly design the low refractive index ring in the fiber core, which can provide an increase of the differential overlap between the fundamental and the higher-order mode. Then, the gain competition among the guided modes along the Yb-doped rod-type fibers has been investigated with a spatial and spectral amplifier model. Simulation results have shown the effectiveness of the sectioned core doping in worsening the higher-order mode overlap on the doped area, thus providing an effective single-mode behavior of the Yb-doped rod-type photonic crystal fibers.

Design and implementation of a simple nuclear power plant simulator

NASA Astrophysics Data System (ADS)

Miller, William H.

1983-02-01

A simple PWR nuclear power plant simulator has been designed and implemented on a minicomputer system. The system is intended for students use in understanding the power operation of a nuclear power plant. A PDP-11 minicomputer calculates reactor parameters in real time, uses a graphics terminal to display the results and a keyboard and joystick for control functions. Plant parameters calculated by the model include the core reactivity (based upon control rod positions, soluble boron concentration and reactivity feedback effects), the total core power, the axial core power distribution, the temperature and pressure in the primary and secondary coolant loops, etc.

Parareal in time 3D numerical solver for the LWR Benchmark neutron diffusion transient model

DOE Office of Scientific and Technical Information (OSTI.GOV)

Baudron, Anne-Marie, E-mail: anne-marie.baudron@cea.fr; CEA-DRN/DMT/SERMA, CEN-Saclay, 91191 Gif sur Yvette Cedex; Lautard, Jean-Jacques, E-mail: jean-jacques.lautard@cea.fr

2014-12-15

In this paper we present a time-parallel algorithm for the 3D neutrons calculation of a transient model in a nuclear reactor core. The neutrons calculation consists in numerically solving the time dependent diffusion approximation equation, which is a simplified transport equation. The numerical resolution is done with finite elements method based on a tetrahedral meshing of the computational domain, representing the reactor core, and time discretization is achieved using a θ-scheme. The transient model presents moving control rods during the time of the reaction. Therefore, cross-sections (piecewise constants) are taken into account by interpolations with respect to the velocity ofmore » the control rods. The parallelism across the time is achieved by an adequate use of the parareal in time algorithm to the handled problem. This parallel method is a predictor corrector scheme that iteratively combines the use of two kinds of numerical propagators, one coarse and one fine. Our method is made efficient by means of a coarse solver defined with large time step and fixed position control rods model, while the fine propagator is assumed to be a high order numerical approximation of the full model. The parallel implementation of our method provides a good scalability of the algorithm. Numerical results show the efficiency of the parareal method on large light water reactor transient model corresponding to the Langenbuch–Maurer–Werner benchmark.« less

Apparatus for controlling molten core debris

DOEpatents

Golden, Martin P. [Trafford, PA; Tilbrook, Roger W. [Monroeville, PA; Heylmun, Neal F. [Pittsburgh, PA

1977-07-19

Apparatus for containing, cooling, diluting, dispersing and maintaining subcritical the molten core debris assumed to melt through the bottom of a nuclear reactor pressure vessel in the unlikely event of a core meltdown. The apparatus is basically a sacrificial bed system which includes an inverted conical funnel, a core debris receptacle including a spherical dome, a spherically layered bed of primarily magnesia bricks, a cooling system of zig-zag piping in graphite blocks about and below the bed and a cylindrical liner surrounding the graphite blocks including a steel shell surrounded by firebrick. Tantalum absorber rods are used in the receptacle and bed.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

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-Pmore » $$_3$$ 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.« less

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.

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.

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.

The influence of MOVPE growth conditions on the shell of core-shell GaN microrod structures

NASA Astrophysics Data System (ADS)

Schimpke, Tilman; Avramescu, Adrian; Koller, Andreas; Fernando-Saavedra, Amalia; Hartmann, Jana; Ledig, Johannes; Waag, Andreas; Strassburg, Martin; Lugauer, Hans-Jürgen

2017-05-01

A core-shell geometry is employed for most next-generation, three-dimensional opto-electric devices based on III-V semiconductors and grown by metal organic vapor phase epitaxy (MOVPE). Controlling the shape of the shell layers is fundamental for device optimization, however no detailed analysis of the influence of growth conditions has been published to date. We study homogeneous arrays of gallium nitride core-shell microrods with height and diameter in the micrometer range and grown in a two-step selective area MOVPE process. Changes in shell shape and homogeneity effected by deliberately altered shell growth conditions were accurately assessed by digital analysis of high-resolution scanning electron microscope images. Most notably, two temperature regimes could be established, which show a significantly different behavior with regard to material distribution. Above 900 °C of wafer carrier temperature, the shell thickness along the growth axis of the rods was very homogeneous, however variations between vicinal rods increase. In contrast, below 830 °C the shell thickness is higher close to the microrod tip than at the base of the rods, while the lateral homogeneity between neighboring microrods is very uniform. This temperature effect could be either amplified or attenuated by changing the remaining growth parameters such as reactor pressure, structure distance, gallium precursor, carrier gas composition and dopant materials. Possible reasons for these findings are discussed with respect to GaN decomposition as well as the surface and gas phase diffusion of growth species, leading to an improved control of the functional layers in next-generation 3D V-III devices.

PM-1 NUCLEAR POWER PLANT PROGRAM. Quarterly Progress Report No. 2 for June 1 to August 31, 1959

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sieg, J.S.; Smith, E.H.

1959-10-01

The objective of the contract is the design, development, fabrication, installation, and initial testing and operation of a prepackaged air- transportable pressurized water reactor nuclear power plant, the PM-1. The specified output is 1 Mwe and 7 million Btu/hr of heat. The plant is to be operational by March 1962. The principal efforts were completion of the plant parametric study and preparation of the preliminary design. A summary of design parameters is given. Systems development work included study and selection of packages for full-scale testing, a survey of in-core instrumentation techniques, control and instrumentation development, and development of components formore » the steam generator, condenser, and turbine generator, which are not commercially available. Reactor development work included completion of the parametric zeropower experiments and preparrtions for a flexible zeropower test program, a revision of plans for irradiation testing PM-1 fuel elements, initiation of a reactor flow test program, outliring of a heat tnansfer test program, completion of the seven-tube test section (SETCH-1) tests, and evaluation of control rod actuators leading to specification of a magnetic jack-type control rod drive similar to that reported in ANL-5768. Completion of the prelimirary design led to initiation of the final design effort, which will be the principal activity during the next two project quarters. Preparations for core fabrication included procurement of core cladding material for the zero-power teat core, arrangement with a subcontractor to convent UF/sub 6/ to UO/sub 2/ and to commence delivery of the oxide during the next quarter, development of fuel element fabrication and ultrasonic testing techniques, study of control rod materials, UO/sub 2/ recovery techniques, and boron analysis methods. Preliminary work on site preparation was pursued with receipt of USAEC approval for a location on the eastern slope of Warren Peak at Sundance, Wyoming. A survey of this site is underway. A preliminary Hazards Summary Report is in preparation. (For preceding period see MND-M-1812.) (auth)« less

Analytical methods in the high conversion reactor core design

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zeggel, W.; Oldekop, W.; Axmann, J.K.

High conversion reactor (HCR) design methods have been used at the Technical University of Braunschweig (TUBS) with the technological support of Kraftwerk Union (KWU). The present state and objectives of this cooperation between KWU and TUBS in the field of HCRs have been described using existing design models and current activities aimed at further development and validation of the codes. The hard physical and thermal-hydraulic boundary conditions of pressurized water reactor (PWR) cores with a high degree of fuel utilization result from the tight packing of the HCR fuel rods and the high fissionable plutonium content of the fuel. Inmore » terms of design, the problem will be solved with rod bundles whose fuel rods are adjusted by helical spacers to the proposed small rod pitches. These HCR properties require novel computational models for neutron physics, thermal hydraulics, and fuel rod design. By means of a survey of the codes, the analytical procedure for present-day HCR core design is presented. The design programs are currently under intensive development, as design tools with a solid, scientific foundation and with essential parameters that are widely valid and are required for a promising optimization of the HCR core. Design results and a survey of future HCR development are given. In this connection, the reoptimization of the PWR core in the direction of an HCR is considered a fascinating scientific task, with respect to both economic and safety aspects.« less

Apparatus for controlling molten core debris. [LMFBR

DOEpatents

Golden, M.P.; Tilbrook, R.W.; Heylmun, N.F.

1977-07-19

Disclosed is an apparatus for containing, cooling, diluting, dispersing and maintaining subcritical the molten core debris assumed to melt through the bottom of a nuclear reactor pressure vessel in the unlikely event of a core meltdown. The apparatus is basically a sacrificial bed system which includes an inverted conical funnel, a core debris receptacle including a spherical dome, a spherically layered bed of primarily magnesia bricks, a cooling system of zig-zag piping in graphite blocks about and below the bed and a cylindrical liner surrounding the graphite blocks including a steel shell surrounded by firebrick. Tantalum absorber rods are used in the receptacle and bed. 9 claims, 22 figures.

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)

Sediment Core Extrusion Method at Millimeter Resolution Using a Calibrated, Threaded-rod.

PubMed

Schwing, Patrick T; Romero, Isabel C; Larson, Rebekka A; O'Malley, Bryan J; Fridrik, Erika E; Goddard, Ethan A; Brooks, Gregg R; Hastings, David W; Rosenheim, Brad E; Hollander, David J; Grant, Guy; Mulhollan, Jim

2016-08-17

Aquatic sediment core subsampling is commonly performed at cm or half-cm resolution. Depending on the sedimentation rate and depositional environment, this resolution provides records at the annual to decadal scale, at best. An extrusion method, using a calibrated, threaded-rod is presented here, which allows for millimeter-scale subsampling of aquatic sediment cores of varying diameters. Millimeter scale subsampling allows for sub-annual to monthly analysis of the sedimentary record, an order of magnitude higher than typical sampling schemes. The extruder consists of a 2 m aluminum frame and base, two core tube clamps, a threaded-rod, and a 1 m piston. The sediment core is placed above the piston and clamped to the frame. An acrylic sampling collar is affixed to the upper 5 cm of the core tube and provides a platform from which to extract sub-samples. The piston is rotated around the threaded-rod at calibrated intervals and gently pushes the sediment out the top of the core tube. The sediment is then isolated into the sampling collar and placed into an appropriate sampling vessel (e.g., jar or bag). This method also preserves the unconsolidated samples (i.e., high pore water content) at the surface, providing a consistent sampling volume. This mm scale extrusion method was applied to cores collected in the northern Gulf of Mexico following the Deepwater Horizon submarine oil release. Evidence suggests that it is necessary to sample at the mm scale to fully characterize events that occur on the monthly time-scale for continental slope sediments.

Sediment Core Extrusion Method at Millimeter Resolution Using a Calibrated, Threaded-rod

PubMed Central

Schwing, Patrick T.; Romero, Isabel C.; Larson, Rebekka A.; O'Malley, Bryan J.; Fridrik, Erika E.; Goddard, Ethan A.; Brooks, Gregg R.; Hastings, David W.; Rosenheim, Brad E.; Hollander, David J.; Grant, Guy; Mulhollan, Jim

2016-01-01

Aquatic sediment core subsampling is commonly performed at cm or half-cm resolution. Depending on the sedimentation rate and depositional environment, this resolution provides records at the annual to decadal scale, at best. An extrusion method, using a calibrated, threaded-rod is presented here, which allows for millimeter-scale subsampling of aquatic sediment cores of varying diameters. Millimeter scale subsampling allows for sub-annual to monthly analysis of the sedimentary record, an order of magnitude higher than typical sampling schemes. The extruder consists of a 2 m aluminum frame and base, two core tube clamps, a threaded-rod, and a 1 m piston. The sediment core is placed above the piston and clamped to the frame. An acrylic sampling collar is affixed to the upper 5 cm of the core tube and provides a platform from which to extract sub-samples. The piston is rotated around the threaded-rod at calibrated intervals and gently pushes the sediment out the top of the core tube. The sediment is then isolated into the sampling collar and placed into an appropriate sampling vessel (e.g., jar or bag). This method also preserves the unconsolidated samples (i.e., high pore water content) at the surface, providing a consistent sampling volume. This mm scale extrusion method was applied to cores collected in the northern Gulf of Mexico following the Deepwater Horizon submarine oil release. Evidence suggests that it is necessary to sample at the mm scale to fully characterize events that occur on the monthly time-scale for continental slope sediments. PMID:27585268

Experimental results from the VENUS-F critical reference state for the GUINEVERE accelerator driven system project

DOE Office of Scientific and Technical Information (OSTI.GOV)

Uyttenhove, W.; Baeten, P.; Kochetkov, A.

The GUINEVERE (Generation of Uninterrupted Intense Neutron pulses at the lead Venus Reactor) project was launched in 2006 within the framework of FP6 EUROTRANS in order to validate online reactivity monitoring and subcriticality level determination in accelerator driven systems (ADS). Therefore, the VENUS reactor at SCK.CEN in Mol, Belgium, was modified towards a fast core (VENUS-F) and coupled to the GENEPI-3C accelerator built by CNRS. The accelerator can operate in both continuous and pulsed mode. The VENUS-F core is loaded with enriched Uranium and reflected with solid lead. A well-chosen critical reference state is indispensable for the validation of themore » online subcriticality monitoring methodology. Moreover, a benchmarking tool is required for nuclear data research and code validation. In this paper, the design and the importance of the critical reference state for the GUINEVERE project are motivated. The results of the first experimental phase on the critical core are presented. The control rods worth is determined by the positive period method and the application of the Modified Source Multiplication (MSM) method allows the determination of the worth of the safety rods. The results are implemented in the VENUS-F core certificate for full exploitation of the critical core. (authors)« less

Thermal breeder fuel enrichment zoning

DOEpatents

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

1992-01-01

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

SPERT I DESTRUCTIVE TEST PROGRAM SAFETY ANALYSIS REPORT

DOE Office of Scientific and Technical Information (OSTI.GOV)

Spano, A.H.; Miller, R.W.

1962-06-15

The water-moderated core used for destructive experiments is mounted in the Spent I open-type reactor vessel, which has no provision for pressurization or forced coolant flow. The core is an array of highly enriched aluminum clad, plate-type fuel assemblies, using four bladetype, gang-operated control rods. Reactor transients are initiated at ambient temperature by step-insentions of reactivity, using a control rod which can be quickly ejected from the core. Following an initial series of static measurements to determine the basic- reactor properties of the test core, a series of nondestructive, self-limiting power excursion tests was performed, which covered a reactor periodmore » range down to the point where minor fuel plate damage first occurred -approximately for a 10- msec period test. These tests provided power, temperature, and pressure data. Additional kinetic teste in the period region between 10 and 5 msec were completed to explore the region of limited core damage. Fuel plate damage results included plate distortion, cladding cracking, and fuel melting. These exploratory tests were valuable in revealing unexpected changes in the dependence of pressure, temperature, burst energy, and burst shape parameters on reactor period, although the dependence of peak power on reactor period was not significantly changed. An evaluation of hazards involved in conducting the 2- msec test, based on pessimistic assumptions regarding fission product release and weather conditions, indicates that with the procedural controls normally exercised in the conduct of any transient test at Spent and the special controls to be in effect during the destructive test series, no significant hazard to personnel or to the general public will be obtained. All nuclear operation is conducted remotely approximately 1/2 mile from the reactor building. Discussion is also given of the supervision and control of personnel during and after each destructive test, and of the plans for re-entry, cleanup, and restoration of the facility. (auth)« less

The development of a thermal hydraulic feedback mechanism with a quasi-fixed point iteration scheme for control rod position modeling for the TRIGSIMS-TH application

NASA Astrophysics Data System (ADS)

Karriem, Veronica V.

Nuclear reactor design incorporates the study and application of nuclear physics, nuclear thermal hydraulic and nuclear safety. Theoretical models and numerical methods implemented in computer programs are utilized to analyze and design nuclear reactors. The focus of this PhD study's is the development of an advanced high-fidelity multi-physics code system to perform reactor core analysis for design and safety evaluations of research TRIGA-type reactors. The fuel management and design code system TRIGSIMS was further developed to fulfill the function of a reactor design and analysis code system for the Pennsylvania State Breazeale Reactor (PSBR). TRIGSIMS, which is currently in use at the PSBR, is a fuel management tool, which incorporates the depletion code ORIGEN-S (part of SCALE system) and the Monte Carlo neutronics solver MCNP. The diffusion theory code ADMARC-H is used within TRIGSIMS to accelerate the MCNP calculations. It manages the data and fuel isotopic content and stores it for future burnup calculations. The contribution of this work is the development of an improved version of TRIGSIMS, named TRIGSIMS-TH. TRIGSIMS-TH incorporates a thermal hydraulic module based on the advanced sub-channel code COBRA-TF (CTF). CTF provides the temperature feedback needed in the multi-physics calculations as well as the thermal hydraulics modeling capability of the reactor core. The temperature feedback model is using the CTF-provided local moderator and fuel temperatures for the cross-section modeling for ADMARC-H and MCNP calculations. To perform efficient critical control rod calculations, a methodology for applying a control rod position was implemented in TRIGSIMS-TH, making this code system a modeling and design tool for future core loadings. The new TRIGSIMS-TH is a computer program that interlinks various other functional reactor analysis tools. It consists of the MCNP5, ADMARC-H, ORIGEN-S, and CTF. CTF was coupled with both MCNP and ADMARC-H to provide the heterogeneous temperature distribution throughout the core. Each of these codes is written in its own computer language performing its function and outputs a set of data. TRIGSIMS-TH provides an effective use and data manipulation and transfer between different codes. With the implementation of feedback and control- rod-position modeling methodologies, the TRIGSIMS-TH calculations are more accurate and in a better agreement with measured data. The PSBR is unique in many ways and there are no "off-the-shelf" codes, which can model this design in its entirety. In particular, PSBR has an open core design, which is cooled by natural convection. Combining several codes into a unique system brings many challenges. It also requires substantial knowledge of both operation and core design of the PSBR. This reactor is in operation decades and there is a fair amount of studies and developments in both PSBR thermal hydraulics and neutronics. Measured data is also available for various core loadings and can be used for validation activities. The previous studies and developments in PSBR modeling also aids as a guide to assess the findings of the work herein. In order to incorporate new methods and codes into exiting TRIGSIMS, a re-evaluation of various components of the code was performed to assure the accuracy and efficiency of the existing CTF/MCNP5/ADMARC-H multi-physics coupling. A new set of ADMARC-H diffusion coefficients and cross sections was generated using the SERPENT code. This was needed as the previous data was not generated with thermal hydraulic feedback and the ARO position was used as the critical rod position. The B4C was re-evaluated for this update. The data exchange between ADMARC-H and MCNP5 was modified. The basic core model is given a flexibility to allow for various changes within the core model, and this feature was implemented in TRIGSIMS-TH. The PSBR core in the new code model can be expanded and changed. This allows the new code to be used as a modeling tool for design and analyses of future code loadings.

The Shock and Vibration Bulletin. Part 3. Structural Dynamics, Machinery Dynamics and Vibration Problems

DTIC Science & Technology

1984-06-01

and to thermopile, but with a dynamically non similar control . Response limiting was accomplished by electric heat source. The test transient measuring...pulse Improvements = Final eport, Space teats were found to be reasonably simple to and Communications Group , Hughes implement and control . The time...coolant flow components, experimental studies are generally from the core is constricted by the presence r of the control rod drive line (CRDL

MODULAR CORE UNITS FOR A NEUTRONIC REACTOR

DOEpatents

Gage, J.F. Jr.; Sherer, D.B.

1964-04-01

A modular core unit for use in a nuclear reactor is described. Many identical core modules can be placed next to each other to make up a complete core. Such a module includes a cylinder of moderator material surrounding a fuel- containing re-entrant coolant channel. The re-entrant channel provides for the circulation of coolant such as liquid sodium from one end of the core unit, through the fuel region, and back out through the same end as it entered. Thermal insulation surrounds the moderator exterior wall inducing heat to travel inwardly to the coolant channel. Spaces between units may be used to accommodate control rods and support structure, which may be cooled by a secondary gas coolant, independently of the main coolant. (AEC)

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.

Evaluation of B&W UO2/ThO2 VIII experimental core: criticality and thermal disadvantage factor analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Carlo Parisi; Emanuele Negrenti

2017-02-01

In the framework of the OECD/NEA International Reactor Physics Experiment (IRPHE) Project, an evaluation of core VIII of the Babcock & Wilcox (B&W) Spectral Shift Control Reactor (SSCR) critical experiment program was performed. The SSCR concept, moderated and cooled by a variable mixture of heavy and light water, envisaged changing of the thermal neutron spectrum during the operation to encourage breeding and to sustain the core criticality. Core VIII contained 2188 fuel rods with 93% enriched UO2-ThO2 fuel in a moderator mixture of heavy and light water. The criticality experiment and measurements of the thermal disadvantage factor were evaluated.

HTR-PROTEUS PEBBLE BED EXPERIMENTAL PROGRAM CORE 4: RANDOM PACKING WITH A 1:1 MODERATOR-TO-FUEL PEBBLE RATIO

DOE Office of Scientific and Technical Information (OSTI.GOV)

John D. Bess; Leland M. Montierth

2013-03-01

In its deployment as a pebble bed reactor (PBR) critical facility from 1992 to 1996, the PROTEUS facility was designated as HTR-PROTEUS. This experimental program was performed as part of an International Atomic Energy Agency (IAEA) Coordinated Research Project (CRP) on the Validation of Safety Related Physics Calculations for Low Enriched HTGRs. Within this project, critical experiments were conducted for graphite moderated LEU systems to determine core reactivity, flux and power profiles, reaction-rate ratios, the worth of control rods, both in-core and reflector based, the worth of burnable poisons, kinetic parameters, and the effects of moisture ingress on these parameters.more » One benchmark experiment was evaluated in this report: Core 4. Core 4 represents the only configuration with random pebble packing in the HTR-PROTEUS series of experiments, and has a moderator-to-fuel pebble ratio of 1:1. Three random configurations were performed. The initial configuration, Core 4.1, was rejected because the method for pebble loading, separate delivery tubes for the moderator and fuel pebbles, may not have been completely random; this core loading was rejected by the experimenters. Cores 4.2 and 4.3 were loaded using a single delivery tube, eliminating the possibility for systematic ordering effects. The second and third cores differed slightly in the quantity of pebbles loaded (40 each of moderator and fuel pebbles), stacked height of the pebbles in the core cavity (0.02 m), withdrawn distance of the stainless steel control rods (20 mm), and withdrawn distance of the autorod (30 mm). The 34 coolant channels in the upper axial reflector and the 33 coolant channels in the lower axial reflector were open. Additionally, the axial graphite fillers used in all other HTR-PROTEUS configurations to create a 12-sided core cavity were not used in the randomly packed cores. Instead, graphite fillers were placed on the cavity floor, creating a funnel-like base, to discourage ordering effects during pebble loading. Core 4 was determined to be acceptable benchmark experiment.« less

HTR-proteus pebble bed experimental program core 4: random packing with a 1:1 moderator-to-fuel pebble ratio

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bess, John D.; Montierth, Leland M.; Sterbentz, James W.

2014-03-01

In its deployment as a pebble bed reactor (PBR) critical facility from 1992 to 1996, the PROTEUS facility was designated as HTR-PROTEUS. This experimental program was performed as part of an International Atomic Energy Agency (IAEA) Coordinated Research Project (CRP) on the Validation of Safety Related Physics Calculations for Low Enriched HTGRs. Within this project, critical experiments were conducted for graphite moderated LEU systems to determine core reactivity, flux and power profiles, reaction-rate ratios, the worth of control rods, both in-core and reflector based, the worth of burnable poisons, kinetic parameters, and the effects of moisture ingress on these parameters.more » One benchmark experiment was evaluated in this report: Core 4. Core 4 represents the only configuration with random pebble packing in the HTR-PROTEUS series of experiments, and has a moderator-to-fuel pebble ratio of 1:1. Three random configurations were performed. The initial configuration, Core 4.1, was rejected because the method for pebble loading, separate delivery tubes for the moderator and fuel pebbles, may not have been completely random; this core loading was rejected by the experimenters. Cores 4.2 and 4.3 were loaded using a single delivery tube, eliminating the possibility for systematic ordering effects. The second and third cores differed slightly in the quantity of pebbles loaded (40 each of moderator and fuel pebbles), stacked height of the pebbles in the core cavity (0.02 m), withdrawn distance of the stainless steel control rods (20 mm), and withdrawn distance of the autorod (30 mm). The 34 coolant channels in the upper axial reflector and the 33 coolant channels in the lower axial reflector were open. Additionally, the axial graphite fillers used in all other HTR-PROTEUS configurations to create a 12-sided core cavity were not used in the randomly packed cores. Instead, graphite fillers were placed on the cavity floor, creating a funnel-like base, to discourage ordering effects during pebble loading. Core 4 was determined to be acceptable benchmark experiment.« less

Core sample extractor

NASA Technical Reports Server (NTRS)

Akins, James; Cobb, Billy; Hart, Steve; Leaptrotte, Jeff; Milhollin, James; Pernik, Mark

1989-01-01

The problem of retrieving and storing core samples from a hole drilled on the lunar surface is addressed. The total depth of the hole in question is 50 meters with a maximum diameter of 100 millimeters. The core sample itself has a diameter of 60 millimeters and will be two meters in length. It is therefore necessary to retrieve and store 25 core samples per hole. The design utilizes a control system that will stop the mechanism at a certain depth, a cam-linkage system that will fracture the core, and a storage system that will save and catalogue the cores to be extracted. The Rod Changer and Storage Design Group will provide the necessary tooling to get into the hole as well as to the core. The mechanical design for the cam-linkage system as well as the conceptual design of the storage device are described.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

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 editionmore » 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.« less

Nanofocus x-ray diffraction and cathodoluminescence investigations into individual core-shell (In,Ga)N/GaN rod light-emitting diodes.

PubMed

Krause, Thilo; Hanke, Michael; Cheng, Zongzhe; Niehle, Michael; Trampert, Achim; Rosenthal, Martin; Burghammer, Manfred; Ledig, Johannes; Hartmann, Jana; Zhou, Hao; Wehmann, Hergo-Heinrich; Waag, Andreas

2016-08-12

Employing nanofocus x-ray diffraction, we investigate the local strain field induced by a five-fold (In,Ga)N multi-quantum well embedded into a GaN micro-rod in core-shell geometry. Due to an x-ray beam width of only 150 nm in diameter, we are able to distinguish between individual m-facets and to detect a significant in-plane strain gradient along the rod height. This gradient translates to a red-shift in the emitted wavelength revealed by spatially resolved cathodoluminescence measurements. We interpret the result in terms of numerically derived in-plane strain using the finite element method and subsequent kinematic scattering simulations which show that the driving parameter for this effect is an increasing indium content towards the rod tip.

Nanofocus x-ray diffraction and cathodoluminescence investigations into individual core-shell (In,Ga)N/GaN rod light-emitting diodes

NASA Astrophysics Data System (ADS)

Krause, Thilo; Hanke, Michael; Cheng, Zongzhe; Niehle, Michael; Trampert, Achim; Rosenthal, Martin; Burghammer, Manfred; Ledig, Johannes; Hartmann, Jana; Zhou, Hao; Wehmann, Hergo-Heinrich; Waag, Andreas

2016-08-01

Employing nanofocus x-ray diffraction, we investigate the local strain field induced by a five-fold (In,Ga)N multi-quantum well embedded into a GaN micro-rod in core-shell geometry. Due to an x-ray beam width of only 150 nm in diameter, we are able to distinguish between individual m-facets and to detect a significant in-plane strain gradient along the rod height. This gradient translates to a red-shift in the emitted wavelength revealed by spatially resolved cathodoluminescence measurements. We interpret the result in terms of numerically derived in-plane strain using the finite element method and subsequent kinematic scattering simulations which show that the driving parameter for this effect is an increasing indium content towards the rod tip.

Development of Optimized Core Design and Analysis Methods for High Power Density BWRs

NASA Astrophysics Data System (ADS)

Shirvan, Koroush

Increasing the economic competitiveness of nuclear energy is vital to its future. Improving the economics of BWRs is the main goal of this work, focusing on designing cores with higher power density, to reduce the BWR capital cost. Generally, the core power density in BWRs is limited by the thermal Critical Power of its assemblies, below which heat removal can be accomplished with low fuel and cladding temperatures. The present study investigates both increases in the heat transfer area between ~he fuel and coolant and changes in operating parameters to achieve higher power levels while meeting the appropriate thermal as well as materials and neutronic constraints. A scoping study is conducted under the constraints of using fuel with cylindrical geometry, traditional materials and enrichments below 5% to enhance its licensability. The reactor vessel diameter is limited to the largest proposed thus far. The BWR with High power Density (BWR-HD) is found to have a power level of 5000 MWth, equivalent to 26% uprated ABWR, resulting into 20% cheaper O&M and Capital costs. This is achieved by utilizing the same number of assemblies, but with wider 16x16 assemblies and 50% shorter active fuel than that of the ABWR. The fuel rod diameter and pitch are reduced to just over 45% of the ABWR values. Traditional cruciform form control rods are used, which restricts the assembly span to less than 1.2 times the current GE14 design due to limitation on shutdown margin. Thus, it is possible to increase the power density and specific power by 65%, while maintaining the nominal ABWR Minimum Critical Power Ratio (MCPR) margin. The plant systems outside the vessel are assumed to be the same as the ABWR-Il design, utilizing a combination of active and passive safety systems. Safety analyses applied a void reactivity coefficient calculated by SIMULA TE-3 for an equilibrium cycle core that showed a 15% less negative coefficient for the BWR-HD compared to the ABWR. The feedwater temperature was kept the same for the BWR-HD and ABWR which resulted in 4 °K cooler core inlet temperature for the BWR-HD given that its feedwater makes up a larger fraction of total core flow. The stability analysis using the STAB and S3K codes showed satisfactory results for the hot channel, coupled regional out-of-phase and coupled core-wide in-phase modes. A RELAPS model of the ABWR system was constructed and applied to six transients for the BWR-HD and ABWR. The 6MCPRs during all the transients were found to be equal or less for the new design and the core remained covered for both. The lower void coefficient along with smaller core volume proved to be advantages for the simulated transients. Helical Cruciform Fuel (HCF) rods were proposed in prior MIT studies to enhance the fuel surface to volume ratio. In this work, higher fidelity models (e.g. CFD instead of subchannel methods for the hydraulic behaviour) are used to investigate the resolution needed for accurate assessment of the HCF design. For neutronics, conserving the fuel area of cylindrical rods results in a different reactivity level with a lower void coefficient for the HCF design. In single-phase flow, for which experimental results existed, the friction factor is found to be sensitive to HCF geometry and cannot be calculated using current empirical models. A new approach for analysis of flow crisis conditions for HCF rods in the context of Departure from Nucleate Boiling (DNB) and dryout using the two phase interface tracking method was proposed and initial results are presented. It is shown that the twist of the HCF rods promotes detachment of a vapour bubble along the elbows which indicates no possibility for an early DNB for the HCF rods and in fact a potential for a higher DNB heat flux. Under annular flow conditions, it was found that the twist suppressed the liquid film thickness on the HCF rods, at the locations of the highest heat flux, which increases the possibility of reaching early dryout. It was also shown that modeling the 3D heat and stress distribution in the HCF rods is necessary for accurate steady state and transient analyses. (Abstract shortened by UMI.) (Copies available exclusively from MIT Libraries, libraries.mit.edu/docs - docs mit.edu)

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.

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.

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.

Molten core retention assembly

DOEpatents

Lampe, Robert F.

1976-06-22

Molten fuel produced in a core overheating accident is caught by a molten core retention assembly consisting of a horizontal baffle plate having a plurality of openings therein, heat exchange tubes having flow holes near the top thereof mounted in the openings, and a cylindrical, imperforate baffle attached to the plate and surrounding the tubes. The baffle assembly is supported from the core support plate of the reactor by a plurality of hanger rods which are welded to radial beams passing under the baffle plate and intermittently welded thereto. Preferably the upper end of the cylindrical baffle terminates in an outwardly facing lip to which are welded a plurality of bearings having slots therein adapted to accept the hanger rods.

Loss of control air at Browns Ferry Unit One: accident sequence analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Harrington, R.M.; Hodge, S.A.

1986-04-01

This study describes the predicted response of the Browns Ferry Nuclear Plant to a postulated complete failure of plant control air. The failure of plant control air cascades to include the loss of drywell control air at Units 1 and 2. Nevertheless, this is a benign accident unless compounded by simultaneous failures in the turbine-driven high pressure injection systems. Accident sequence calculations are presented for Loss of Control Air sequences with assumed failure upon demand of the Reactor Core Isolation Cooling (RCIC) and the High Pressure Coolant Injection (HPCI) at Unit 1. Sequences with and without operator action are considered.more » Results show that the operators can prevent core uncovery if they take action to utilize the Control Rod Drive Hydraulic System as a backup high pressure injection system.« less

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.

Utilization of thorium and U-ZrH1.6 fuels in various heterogeneous cores for TRIGA PUSPATI Reactor (RTP)

NASA Astrophysics Data System (ADS)

Damahuri, Abdul Hannan Bin; Mohamed, Hassan; Aziz Mohamed, Abdul; Idris, Faridah

2018-01-01

The use of thorium as nuclear fuel has been an appealing prospect for many years and will be great significance to nuclear power generation. There is an increasing need for more research on thorium as Malaysian government is currently active in the national Thorium Flagship Project, which was launched in 2014. The thorium project, which is still in phase 1, focuses on the research and development of the thorium extraction from mineral processing ore. Thus, the aim of the study is to investigate other alternative TRIGA PUSPATI Reactor (RTP) core designs that can fully utilize thorium. Currently, the RTP reactor has an average neutron flux of 2.797 x 1012 cm-2/s-1 and an effective multiplication factor, k eff, of 1.001. The RTP core has a circular array core configuration with six circular rings. Each ring consists of 6, 12, 18, 24, 30 or 36 U-ZrH1.6 fuel rods. There are three main type of uranium weight, namely 8.5, 12 and 20 wt.%. For this research, uranium zirconium hydride (U-ZrH1.6) fuel rods in the RTP core were replaced by thorium (ThO2) fuel rods. Seven core configurations with different thorium fuel rods placements were modelled in a 2D structure and simulated using Monte Carlo n-particle (MCNPX) code. Results show that the highest initial criticality obtained is around 1.35101. Additionally there is a significant discrepancy between results from previous study and the work because of the large estimated leakage probability of approximately 21.7% and 2D model simplification.

Molecular Design of Branched and Binary Molecules at Ordered Interfaces

DOE Office of Scientific and Technical Information (OSTI.GOV)

Genson, Kirsten Larson

2005-01-01

This study examined five different branched molecular architectures to discern the effect of design on the ability of molecules to form ordered structures at interfaces. Photochromic monodendrons formed kinked packing structures at the air-water interface due to the cross-sectional area mismatch created by varying number of alkyl tails and the hydrophilic polar head group. The lower generations formed orthorhombic unit cell with long range ordering despite the alkyl tails tilted to a large degree. Favorable interactions between liquid crystalline terminal groups and the underlying substrate were observed to compel a flexible carbosilane dendrimer core to form a compressed elliptical conformationmore » which packed stagger within lamellae domains with limited short range ordering. A twelve arm binary star polymer was observed to form two dimensional micelles at the air-water interface attributed to the higher polystyrene block composition. Linear rod-coil molecules formed a multitude of packing structures at the air-water interface due to the varying composition. Tree-like rod-coil molecules demonstrated the ability to form one-dimensional structures at the air-water interface and at the air-solvent interface caused by the preferential ordering of the rigid rod cores. The role of molecular architecture and composition was examined and the influence chemically competing fragments was shown to exert on the packing structure. The amphiphilic balance of the different molecular series exhibited control on the ordering behavior at the air-water interface and within bulk structures. The shell nature and tail type was determined to dictate the preferential ordering structure and molecular reorganization at interfaces with the core nature effect secondary.« less

Material test machine for tension-compression tests at high temperature

DOEpatents

Cioletti, Olisse C.

1988-01-01

Apparatus providing a device for testing the properties of material specimens at high temperatures and pressures in controlled water chemistries includes, inter alia, an autoclave housing the specimen which is being tested. The specimen is connected to a pull rod which couples out of the autoclave to an external assembly which includes one or more transducers, a force balance chamber and a piston type actuator. The pull rod feeds through the force balance chamber and is compensated thereby for the pressure conditions existing within the autoclave and tending to eject the pull rod therefrom. The upper end of the push rod is connected to the actuator through elements containing a transducer comprising a linear variable differential transformer (LVDT). The housing and coil assembly of the LVDT is coupled to a tube which runs through a central bore of the pull rod into the autoclave where it is connected to one side of the specimen. The movable core of the LVDT is coupled to a stem which runs through the tube where it is then connected to the other side of the specimen through a coupling member. A transducer in the form of a load cell including one or more strain gages is located on a necked-down portion of the upper part of the pull rod intermediate the LVDT and force balance chamber.

Nitrogen-polar core-shell GaN light-emitting diodes grown by selective area metalorganic vapor phase epitaxy

NASA Astrophysics Data System (ADS)

Li, Shunfeng; Wang, Xue; Fündling, Sönke; Erenburg, Milena; Ledig, Johannes; Wei, Jiandong; Wehmann, Hergo H.; Waag, Andreas; Bergbauer, Werner; Mandl, Martin; Strassburg, Martin; Trampert, Achim; Jahn, Uwe; Riechert, Henning; Jönen, Holger; Hangleiter, Andreas

2012-07-01

Homogeneous nitrogen-polar GaN core-shell light emitting diode (LED) arrays were fabricated by selective area growth on patterned substrates. Transmission electron microscopy measurements prove the core-shell structure of the rod LEDs. Depending on the growth facets, the InGaN/GaN multi-quantum wells (MQWs) show different dimensions and morphology. Cathodoluminescence (CL) measurements reveal a MQWs emission centered at about 415 nm on sidewalls and another emission at 460 nm from top surfaces. CL line scans on cleaved rod also indicate the core-shell morphology. Finally, an internal quantum efficiency of about 28% at room temperature was determined by an all-optical method on a LED array.

RAPID-L Highly Automated Fast Reactor Concept Without Any Control Rods (2) Critical experiment of lithium-6 used in LEM and LIM

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tsunoda, Hirokazu; Sato, Osamu; Okajima, Shigeaki

2002-07-01

In order to achieve fully automated reactor operation of RAPID-L reactor, innovative reactivity control systems LEM, LIM, and LRM are equipped with lithium-6 as a liquid poison. Because lithium-6 has not been used as a neutron absorbing material of conventional fast reactors, measurements of the reactivity worth of Lithium-6 were performed at the Fast Critical Assembly (FCA) of Japan Atomic Energy Research Institute (JAERI). The FCA core was composed of highly enriched uranium and stainless steel samples so as to simulate the core spectrum of RAPID-L. The samples of 95% enriched lithium-6 were inserted into the core parallel to themore » core axis for the measurement of the reactivity worth at each position. It was found that the measured reactivity worth in the core region well agreed with calculated value by the method for the core designs of RAPID-L. Bias factors for the core design method were obtained by comparing between experimental and calculated results. The factors were used to determine the number of LEM and LIM equipped in the core to achieve fully automated operation of RAPID-L. (authors)« less

CHAP-2 heat-transfer analysis of the Fort St. Vrain reactor core

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kotas, J.F.; Stroh, K.R.

1983-01-01

The Los Alamos National Laboratory is developing the Composite High-Temperature Gas-Cooled Reactor Analysis Program (CHAP) to provide advanced best-estimate predictions of postulated accidents in gas-cooled reactor plants. The CHAP-2 reactor-core model uses the finite-element method to initialize a two-dimensional temperature map of the Fort St. Vrain (FSV) core and its top and bottom reflectors. The code generates a finite-element mesh, initializes noding and boundary conditions, and solves the nonlinear Laplace heat equation using temperature-dependent thermal conductivities, variable coolant-channel-convection heat-transfer coefficients, and specified internal fuel and moderator heat-generation rates. This paper discusses this method and analyzes an FSV reactor-core accident thatmore » simulates a control-rod withdrawal at full power.« less

A COMPARISON OF EXPERIMENTS AND THREE-DIMENSIONAL ANALYSIS TECHNIQUES. PART I. UNPOISONED UNIFORM SLAB CORE WITH A PARTIALLY INSERTED HAFNIUM ROD

DOE Office of Scientific and Technical Information (OSTI.GOV)

Renzi, N.E.; Roseberry, R.J.

>The experimental measurements and nuclear analysis of a uniformly loaded, unpoisoned slab core with a partially insented hafnium rod are described. Comparisons of experimental data with calculated results of the UFO code and flux synthesis techniques are given. It was concluded that one of the flux synthesis techniques and the UFO code are able to predict flux distributions to within approximately 5% of experiment for most cases. An error of approximately 10% was found in the synthesis technique for a channel near the partially inserted rod. The various calculations were able to predict neutron pulsed shutdowns to only approximately 30%.more » (auth)« less

Multi-Fresnel lenses pumping approach for improving high-power Nd:YAG solar laser beam quality.

PubMed

Liang, Dawei; Almeida, Joana

2013-07-20

To significantly improve the present-day high-power solar laser beam quality, a three-stage multi-Fresnel lenses approach is proposed for side-pumping either a Nd:YAG single-crystal or a core-doped Sm(3+)Nd:YAG ceramic rod. Optimum pumping and laser beam parameters are found through ZEMAX and LASCAD numerical analysis. The proposed scheme offers a uniform absorption profile along the rod. 167 W laser power can be achieved, corresponding to 29.3 W/m(2) collection efficiency. High brightness figure of merit of 8.34 W is expected for the core-doped rod within a convex-concave resonator, which is 1300 times higher than that of the most-recent high-power solar laser.

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.

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.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

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

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

PLUTONIUM FUEL RODS FOR PREPARATION OF TRANSPLUTONIC ELEMENTS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bailey, W.J.

1962-02-01

Production by coextrusion of metallurgically bonded, Alclad, Al-7.35 wt% Pu alloy fuel rods with integral ends is discussed. The rods had a diameter of 0.94 in., length of, 60 in., and a nominal cladding thickness of 0.070 in. The Pu concentration was maintained at 83.3 g/rod. The coextrusion billets can be assembled with fuel cores in the as-cast condition. The casting hot-tops can be returned to the process stream. The process is useful for preparing transplutonic elements and production of high-exposure Pu. (J.R.D.)

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.

ACHILLES: Heat Transfer in PWR Core During LOCA Reflood Phase

DOE Office of Scientific and Technical Information (OSTI.GOV)

2013-11-01

1. NAME AND TITLE OF DATA LIBRARY ACHILLES -Heat Transfer in PWR Core During LOCA Reflood Phase. 2. NAME AND TITLE OF DATA RETRIEVAL PROGRAMS N/A 3. CONTRIBUTOR AEA Technology, Winfrith Technology Centre, Dorchester DT2 8DH United Kingdom through the OECD Nuclear Energy Agency Data Bank, Issy-les-Moulineaux, France. 4. DESCRIPTION OF TEST FACILITY The most important features of the Achilles rig were the shroud vessel, which contained the test section, and the downcomer. These may be thought of as representing the core barrel and the annular downcomer in the reactor pressure vessel. The test section comprises a cluster of 69more » rods in a square array within a circular shroud vessel. The rod diameter and pitch (9.5 mm and 12.6 mm) were typical of PWR dimensions. The internal diameter of the shroud vessel was 128 mm. Each rod was electrically heated over a length of 3.66 m, which is typical of the nuclear heated length in a PWR fuel rod, and each contained 6 internal thermocouples. These were arranged in one of 8 groupings which concentrated the thermocouples in different axial zones. The spacer grids were at prototypic PWR locations. Each grid had two thermocouples attached to its trailing edge at radial locations. The axial power profile along the rods was an 11 step approximation to a "chopped cosine". The shroud vessel had 5 heating zones whose power could be independently controlled. 5. DESCRIPTION OF TESTS The Achilles experiments investigated the heat transfer in the core of a Pressurized Water Reactor during the re-flood phase of a postulated large break loss of coolant accident. The results provided data to validate codes and to improve modeling. Different types of experiments were carried out which included single phase cooling, re-flood under low flow conditions, level swell and re-flood under high flow conditions. Three series of experiments were performed. The first and the third used the same test section but the second used another test section, similar in all respects except that it contained a partial blockage formed by attaching sleeves (or "balloons") to some of the rods. 6. SOURCE AND SCOPE OF DATA Phenomena Tested - Heat transfer in the core of a PWR during a re-flood phase of postulated large break LOCA. Test Designation - Achilles Rig. The programme includes the following types of experiments: - on an unballooned cluster: -- single phase air flow -- low pressure level swell -- low flooding rate re-flood -- high flooding rate re-flood - on a ballooned cluster containing 80% blockage formed by 16 balloon sleeves -- single phase air flow -- low flooding rate re-flood 7. DISCUSSION OF THE DATA RETRIEVAL PROGRAM N/A 8. DATA FORMAT AND COMPUTER Many Computers (M00019MNYCP00). 9. TYPICAL RUNNING TIME N/A 11. CONTENTS OF LIBRARY The ACHILLES package contains test data and associated data processing software as well as the documentation listed above. 12. DATE OF ABSTRACT November 2013. KEYWORDS: DATABASES, BENCHMARKS, HEAT TRANSFER, LOSS-OF-COLLANT ACCIDENT, PWR REACTORS, REFLOODING« less

Design of a magnetorheological automotive shock absorber

NASA Astrophysics Data System (ADS)

Lindler, Jason E.; Dimock, Glen A.; Wereley, Norman M.

2000-06-01

Double adjustable shock absorbers allow for independent adjustment of the yield force and post-yield damping in the force versus velocity response. To emulate the performance of a conventional double adjustable shock absorber, a magnetorheological (MR) automotive shock absorber was designed and fabricated at the University of Maryland. Located in the piston head, an applied magnetic field between the core and flux return increases the force required for a given piston rod velocity. Between the core and flux return, two different shaped gaps meet the controllable performance requirements of a double adjustable shock. A uniform gap between the core and the flux return primarily adjusts the yield force of the shock absorber, while a non-uniform gap allows for control of the post-yield damping. Force measurements from sinusoidal displacement cycles, recorded on a mechanical damper dynamometer, validate the performance of uniform and non- uniform gaps for adjustment of the yield force and post-yield damping, respectively.

Inhomogeneities and segregation behavior in strontium—barium niobate fibers grown by laser-heated pedestal growth technique. Part II

NASA Astrophysics Data System (ADS)

Erdei, S.; Galambos, L.; Tanaka, I.; Hesselink, L.; Cross, L. E.; Feigelson, R. S.; Ainger, F. W.; Kojima, H.

1996-10-01

Inhomogeneities in Ce-doped and undoped fibers grown by laser-heated pedestal growth (LHPG) along the c- or a- axis were investigated by two-dimensional scanning electron microprobe analysis (SEPMA). SEPMA data indicated that these cores are primarily connected with the source rod compositions utilized and the convection characteristics of the LHPG technique. Ba enrichment and Sr decrease were primarily detected in the cores and qualitatively described in terms of the composition-control mechanism of LHPG, the complex-segregation and a modified Burton—Prim—Slichter (BPS) equation. Certain aspects of defect structure as a complex congruency related phenomenon are also discussed in the paper giving a more complete interpretation of the origin of cores in SBN fibers.

Light-Water Breeder Reactor

DOEpatents

Beaudoin, B. R.; Cohen, J. D.; Jones, D. H.; Marier, Jr, L. J.; Raab, H. F.

1972-06-20

Described is a light-water-moderated and -cooled nuclear breeder reactor of the seed-blanket type characterized by core modules comprising loosely packed blanket zones enriched with fissile fuel and axial zoning in the seed and blanket regions within each core module. Reactivity control over lifetime is achieved by axial displacement of movable seed zones without the use of poison rods in the embodiment illustrated. The seed is further characterized by a hydrogen-to-uranium-233 atom ratio in the range 10 to 200 and a uranium-233-to-thorium-232 atom ratio ranging from 0.012 to 0.200. The seed occupies from 10 to 35 percent of the core volume in the form of one or more individual islands or annuli. (NSA 26: 55130)

Light-water breeder reactor (LWBR Development Program)

DOEpatents

Beaudoin, B.R.; Cohen, J.D.; Jones, D.H.; Marier, L.J. Jr.; Raab, H.F.

1972-06-20

Described is a light-water-moderated and -cooled nuclear breeder reactor of the seed-blanket type characterized by core modules comprising loosely packed blanket zones enriched with fissile fuel and axial zoning in the seed and blanket regions within each core module. Reactivity control over lifetime is achieved by axial displacement of movable seed zones without the use of poison rods in the embodiment illustrated. The seed is further characterized by a hydrogen-to-uranium-233 atom ratio in the range 10 to 200 and a uranium-233-to-thorium-232 atom ratio ranging from 0.012 to 0.200. The seed occupies from 10 to 35 percent of the core volume in the form of one or more individual islands or annuli. (NSA 26: 55130)

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.

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)

Peripherin-2 couples rhodopsin to the CNG channel in outer segments of rod photoreceptors.

PubMed

Becirovic, Elvir; Nguyen, O N Phuong; Paparizos, Christos; Butz, Elisabeth S; Stern-Schneider, Gabi; Wolfrum, Uwe; Hauck, Stefanie M; Ueffing, Marius; Wahl-Schott, Christian; Michalakis, Stylianos; Biel, Martin

2014-11-15

Outer segments (OSs) of rod photoreceptors are cellular compartments specialized in the conversion of light into electrical signals. This process relies on the light-triggered change in the intracellular levels of cyclic guanosine monophosphate, which in turn controls the activity of cyclic nucleotide-gated (CNG) channels in the rod OS plasma membrane. The rod CNG channel is a macromolecular complex that in its core harbors the ion-conducting CNGA1 and CNGB1a subunits. To identify additional proteins of the complex that interact with the CNGB1a core subunit, we applied affinity purification of mouse retinal proteins followed by mass spectrometry. In combination with in vitro and in vivo co-immunoprecipitation and fluorescence resonance energy transfer (FRET), we found that the tetraspanin peripherin-2 links CNGB1a to the light-detector rhodopsin. Using immunoelectron microscopy, we found that this peripherin-2/rhodopsin/CNG channel complex localizes to the contact region between the disk rims and the plasma membrane. FRET measurements revealed that the fourth transmembrane domain (TM4) of peripherin-2 is required for the interaction with rhodopsin. Quantitatively, the binding affinity of the peripherin-2/rhodopsin interaction was in a similar range as that observed for rhodopsin dimers. Finally, we demonstrate that the p.G266D retinitis pigmentosa mutation found within TM4 selectively abolishes the binding of peripherin-2 to rhodopsin. This finding suggests that the specific disruption of the rhodopsin/peripherin-2 interaction in the p.G266D mutant might contribute to the pathophysiology in affected persons. © The Author 2014. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

Effect of Osteonecrosis Intervention Rod Versus Core Decompression Using Multiple Small Drill Holes on Early Stages of Necrosis of the Femoral Head: A Prospective Study on a Series of 60 Patients with a Minimum 1-Year-Follow-Up.

PubMed

Miao, Haixiong; Ye, Dongping; Liang, Weiguo; Yao, Yicun

2015-01-01

The conventional CD used 10 mm drill holes associated with a lack of structural support. Thus, alternative methods such as a tantalum implant, small drill holes, and biological treatment were developed to prevent deterioration of the joint. The treatment of CD by multiple 3.2 mm drill holes could reduce the femoral neck fracture and partial weight bearing was allowed. This study was aimed to evaluate the effect of osteonecrosis intervention rod versus core decompression using multiple small drill holes on early stages of necrosis of the femoral head. From January 2011 to January 2012, 60 patients undergoing surgery for osteonecrosis with core decompression were randomly assigned into 2 groups based on the type of core decompression used: (1) a total of 30 osteonecrosis patients (with 16 hips on Steinburg stageⅠ,20 hips on Steinburg stageⅡ) were treated with a porous tantalum rod insertion. The diameter of the drill hole for the intervention rod was 10mm.(2) a total of 30 osteonecrosis patients (with 14 hips on Steinburg stageⅠ,20 hips on Steinburg stageⅡ) were treated with core decompression using five drill holes on the lateral femur, the diameter of the hole was 3.2 mm. The average age of the patient was 32.6 years (20-45 years) and the average time of follow-up was 25.6 months (12- 28 months) in the rod implanted group. The average age of the patient was 35.2 years (22- 43 years) and the average time of follow-up was 26.3 months (12-28 months) in the small drill holes group. The average of surgical time was 40 min, and the mean volume of blood loss was 30 ml in both surgical groups. The average of Harris score was improved from 56.2 ± 7.1 preoperative to 80.2 ± 11.4 at the last follow-up in the rod implanted group (p < 0.05). The mean Harris score was improved from 53.8 ± 6.6 preoperative to 79.7 ± 13.2 at the last follow-up in the small drill holes group (p<0. 05). No significant difference was observed in Harris score between the two groups. At the last follow-up, 28 of 36 hips were at the same radiographic stages as pre-operation, and 8 deteriorated in the rod implanted group. 26 of 34 hips were at the same radiographic stage as pre-operation, and 8 deteriorated in the small drill holes group. No significant difference was observed in radiographic stage between the two groups. There was no favourable result on the outcome of a tantalum intervention implant compared to multiple small drill holes. CD via multiple small drill holes would allow similar postoperative load-bearing and seems to result in similar or even better clinical outcome without the prolonged implantation of an expensive tantalum implant. A tantalum rod intervention and core decompression using multiple small drill holes were effective on the stage I hips rather than stage II hips.

Reactor Physics Assessment of Thick Silicon Carbide Clad PWR Fuels

DTIC Science & Technology

2013-06-01

Densities ............................................................................................................ 21 2.3 Fuel Mass (Core Total...70 7.1 Geometry, Material Density, and Mass Summary for All Cores...21 Table 3: Fuel Rod Masses for Different Clads

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gehin, Jess C; Godfrey, Andrew T; Evans, Thomas M

The Consortium for Advanced Simulation of Light Water Reactors (CASL) is developing a collection of methods and software products known as VERA, the Virtual Environment for Reactor Applications, including a core simulation capability called VERA-CS. A key milestone for this endeavor is to validate VERA against measurements from operating nuclear power reactors. The first step in validation against plant data is to determine the ability of VERA to accurately simulate the initial startup physics tests for Watts Bar Nuclear Power Station, Unit 1 (WBN1) cycle 1. VERA-CS calculations were performed with the Insilico code developed at ORNL using cross sectionmore » processing from the SCALE system and the transport capabilities within the Denovo transport code using the SPN method. The calculations were performed with ENDF/B-VII.0 cross sections in 252 groups (collapsed to 23 groups for the 3D transport solution). The key results of the comparison of calculations with measurements include initial criticality, control rod worth critical configurations, control rod worth, differential boron worth, and isothermal temperature reactivity coefficient (ITC). The VERA results for these parameters show good agreement with measurements, with the exception of the ITC, which requires additional investigation. Results are also compared to those obtained with Monte Carlo methods and a current industry core simulator.« less

The change of radial power factor distribution due to RCCA insertion at the first cycle core of AP1000

NASA Astrophysics Data System (ADS)

Susilo, J.; Suparlina, L.; Deswandri; Sunaryo, G. R.

2018-02-01

The using of a computer program for the PWR type core neutronic design parameters analysis has been carried out in some previous studies. These studies included a computer code validation on the neutronic parameters data values resulted from measurements and benchmarking calculation. In this study, the AP1000 first cycle core radial power peaking factor validation and analysis were performed using CITATION module of the SRAC2006 computer code. The computer code has been also validated with a good result to the criticality values of VERA benchmark core. The AP1000 core power distribution calculation has been done in two-dimensional X-Y geometry through ¼ section modeling. The purpose of this research is to determine the accuracy of the SRAC2006 code, and also the safety performance of the AP1000 core first cycle operating. The core calculations were carried out with the several conditions, those are without Rod Cluster Control Assembly (RCCA), by insertion of a single RCCA (AO, M1, M2, MA, MB, MC, MD) and multiple insertion RCCA (MA + MB, MA + MB + MC, MA + MB + MC + MD, and MA + MB + MC + MD + M1). The maximum power factor of the fuel rods value in the fuel assembly assumedapproximately 1.406. The calculation results analysis showed that the 2-dimensional CITATION module of SRAC2006 code is accurate in AP1000 power distribution calculation without RCCA and with MA+MB RCCA insertion.The power peaking factor on the first operating cycle of the AP1000 core without RCCA, as well as with single and multiple RCCA are still below in the safety limit values (less then about 1.798). So in terms of thermal power generated by the fuel assembly, then it can be considered that the AP100 core at the first operating cycle is safe.

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 predictions. After confirming that the new fuel thermal conductivity model in CTF worked and provided consistent results with FRAPTRAN predictions for a single fuel rod configuration, the same type of analysis was carried out for a bigger system which is the 4x4 PWR bundle consisting of 15 fuel pins and one control guide tube. Steady- state calculations at Hot Full Power (HFP) conditions for control guide tube out (unrodded) were performed using the 4x4 PWR array with CTF/TORT-TD coupled code system. Fuel centerline, surface and average temperatures predicted by CTF/TORT-TD with and without the new fuel thermal conductivity model were compared against CTF/TORT-TD/FRAPTRAN predictions to demonstrate the improvement in fuel centerline predictions when new model was used. In addition to that constant and CTF dynamic gap conductance model were used with the new thermal conductivity model to show the performance of the CTF dynamic gap conductance model and its impact on fuel centerline and surface temperatures. Finally, a Rod Ejection Accident (REA) scenario using the same 4x4 PWR array was run both at Hot Zero Power (HZP) and Hot Full Power (HFP) condition, starting at a position where half of the control rod is inserted. This scenario was run using CTF/TORT-TD coupled code system with and without the new fuel thermal conductivity model. The purpose of this transient analysis was to show the impact of thermal conductivity degradation (TCD) on feedback effects, specifically Doppler Reactivity Coefficient (DRC) and, eventually, total core reactivity.

Wireline system for multiple direct push tool usage

DOEpatents

Bratton, Wesley L.; Farrington, Stephen P.; Shinn, II, James D.; Nolet, Darren C.

2003-11-11

A tool latching and retrieval system allows the deployment and retrieval of a variety of direct push subsurface characterization tools through an embedded rod string during a single penetration without requiring withdrawal of the string from the ground. This enables the in situ interchange of different tools, as well as the rapid retrieval of soil core samples from multiple depths during a single direct push penetration. The system includes specialized rods that make up the rod string, a tool housing which is integral to the rod string, a lock assembly, and several tools which mate to the lock assembly.

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.

Preliminary posttest analysis of LOFT loss-of-coolant experiment L2-2

DOE Office of Scientific and Technical Information (OSTI.GOV)

White, J.R.; Grush, W.H.; Keeler, C.D.

A preliminary posttest analysis of Loss-of-Coolant Experiment (LOCE) L2-2, which was conducted in the Loss-of-Fluid Test (LOFT) facility, was performed to gain an understanding of the cause of the disparity between predicted and measured fuel rod cladding temperature responses in the LOFT core. LOCE L2-2 is the first experiment in the LOFT Power Ascension Series L2 (first series of LOFT nuclear experiments), which was designed to investigate the response of the LOFT nuclear core to the blowdown, refill, and reflood transients during LOCEs conducted at gradually increasing power levels. LOCE L2-2 was conducted at 50% power (25 MW, 26.38 kW/m).more » Results show that a core-wide rewet occurred early in the transient (during blowdown starting at about 7 s after rupture) which was not calculated in the pretest prediction analysis. This early core-wide rewet resulted in the peak fuel rod cladding temperatures being lower (by a mean value of 166/sup 0/K for 24 thermocouples) than had been calculated. This preliminary posttest analysis was concerned solely with determining why the early core-wide rewet was not predicted by the RELAP4/MOD6 pretest analysis and be no means is it a complete posttest analysis of LOCE L2-2 results. However, during this analysis, several errors made in the prettest analysis were found, and their impact on the predicted results is assessed. Three factors were postulated to have caused the disparity between predicted and measured fuel rod cladding temperatures for LOCE L2-2: (a) the initial fuel rod stored energy, (b) the heat transfer surface, and (c) the hydraulics calculation. These factors were examined and are discussed in this report. It was determined that core hydraulics, as influenced by the calculation of broken loop cold leg break flow, was the major factor causing the disparity.« less

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.

Controlled fragmentation of multimaterial fibres and films via polymer cold-drawing.

PubMed

Shabahang, Soroush; Tao, Guangming; Kaufman, Joshua J; Qiao, Yangyang; Wei, Lei; Bouchenot, Thomas; Gordon, Ali P; Fink, Yoel; Bai, Yuanli; Hoy, Robert S; Abouraddy, Ayman F

2016-06-23

Polymer cold-drawing is a process in which tensile stress reduces the diameter of a drawn fibre (or thickness of a drawn film) and orients the polymeric chains. Cold-drawing has long been used in industrial applications, including the production of flexible fibres with high tensile strength such as polyester and nylon. However, cold-drawing of a composite structure has been less studied. Here we show that in a multimaterial fibre composed of a brittle core embedded in a ductile polymer cladding, cold-drawing results in a surprising phenomenon: controllable and sequential fragmentation of the core to produce uniformly sized rods along metres of fibre, rather than the expected random or chaotic fragmentation. These embedded structures arise from mechanical-geometric instabilities associated with 'neck' propagation. Embedded, structured multimaterial threads with complex transverse geometry are thus fragmented into a periodic train of rods held stationary in the polymer cladding. These rods can then be easily extracted via selective dissolution of the cladding, or can self-heal by thermal restoration to re-form the brittle thread. Our method is also applicable to composites with flat rather than cylindrical geometries, in which case cold-drawing leads to the break-up of an embedded or coated brittle film into narrow parallel strips that are aligned normally to the drawing axis. A range of materials was explored to establish the universality of this effect, including silicon, germanium, gold, glasses, silk, polystyrene, biodegradable polymers and ice. We observe, and verify through nonlinear finite-element simulations, a linear relationship between the smallest transverse scale and the longitudinal break-up period. These results may lead to the development of dynamical and thermoreversible camouflaging via a nanoscale Venetian-blind effect, and the fabrication of large-area structured surfaces that facilitate high-sensitivity bio-detection.

Controlled fragmentation of multimaterial fibres and films via polymer cold-drawing

NASA Astrophysics Data System (ADS)

Shabahang, Soroush; Tao, Guangming; Kaufman, Joshua J.; Qiao, Yangyang; Wei, Lei; Bouchenot, Thomas; Gordon, Ali P.; Fink, Yoel; Bai, Yuanli; Hoy, Robert S.; Abouraddy, Ayman F.

2016-06-01

Polymer cold-drawing is a process in which tensile stress reduces the diameter of a drawn fibre (or thickness of a drawn film) and orients the polymeric chains. Cold-drawing has long been used in industrial applications, including the production of flexible fibres with high tensile strength such as polyester and nylon. However, cold-drawing of a composite structure has been less studied. Here we show that in a multimaterial fibre composed of a brittle core embedded in a ductile polymer cladding, cold-drawing results in a surprising phenomenon: controllable and sequential fragmentation of the core to produce uniformly sized rods along metres of fibre, rather than the expected random or chaotic fragmentation. These embedded structures arise from mechanical-geometric instabilities associated with ‘neck’ propagation. Embedded, structured multimaterial threads with complex transverse geometry are thus fragmented into a periodic train of rods held stationary in the polymer cladding. These rods can then be easily extracted via selective dissolution of the cladding, or can self-heal by thermal restoration to re-form the brittle thread. Our method is also applicable to composites with flat rather than cylindrical geometries, in which case cold-drawing leads to the break-up of an embedded or coated brittle film into narrow parallel strips that are aligned normally to the drawing axis. A range of materials was explored to establish the universality of this effect, including silicon, germanium, gold, glasses, silk, polystyrene, biodegradable polymers and ice. We observe, and verify through nonlinear finite-element simulations, a linear relationship between the smallest transverse scale and the longitudinal break-up period. These results may lead to the development of dynamical and thermoreversible camouflaging via a nanoscale Venetian-blind effect, and the fabrication of large-area structured surfaces that facilitate high-sensitivity bio-detection.

The R&D PERFROI Project on Thermal Mechanical and Thermal Hydraulics Behaviors of a Fuel Rod Assembly during a Loss of Coolant Accident

DOE Office of Scientific and Technical Information (OSTI.GOV)

Repetto, G.; Dominguez, C.; Durville, B.

The safety principle in case of a LOCA is to preserve the short and long term coolability of the core. The associated safety requirements are to ensure the resistance of the fuel rods upon quench and post-quench loads and to maintain a coolable geometry in the core. An R&D program has been launched by IRSN with the support of EDF, to perform both experimental and modeling activities in the frame of the LOCA transient, on technical issues such as: - flow blockage within a fuel rods bundle and its potential impact on coolability, - fuel fragment relocation in the balloonedmore » areas: its potential impact on cladding PCT (Peak Cladding Temperature) and on the maximum oxidation rate, - potential loss of cladding integrity upon quench and post-quench loads. The PERFROI project (2014-2019) focusing on the first above issue, is structured in two axes: 1. axis 1: thermal mechanical behavior of deformation and rupture of cladding taking into account the contact between fuel rods; specific research at LaMCoS laboratory focus on the hydrogen behavior in cladding alloys and its impact on the mechanical behavior of the rod; and, 2. axis 2: thermal hydraulics study of a partially blocked region of the core (ballooned area taking into account the fuel relocation with local over power), during cooling phase by water injection; More detailed activities foreseen in collaboration with LEMTA laboratory will focus on the characterization of two phase flows with heat transfer in deformed structures.« less

Single-Mode WGM Resonators Fabricated by Diamond Turning

NASA Technical Reports Server (NTRS)

Grudinin, Ivan; Maleki, Lute; Savchenkov, Anatoliy; Matsko, Andrewy; Strekalov, Dmitry; Iltchenko, Vladimir

2008-01-01

A diamond turning process has made possible a significant advance in the art of whispering-gallery-mode (WGM) optical resonators. By use of this process, it is possible to fashion crystalline materials into WGM resonators that have ultrahigh resonance quality factors (high Q values), are compact (ranging in size from millimeters down to tens of microns), and support single electromagnetic modes. This development combines and extends the developments reported in "Few- Mode Whispering-Gallery-Mode Resonators" (NPO-41256), NASA Tech Briefs, Vol. 30, No. 1 (January 2006), page 16a and "Fabrication of Submillimeter Axisymmetric Optical Components" (NPO-42056), NASA Tech Briefs, Vol. 31, No. 5 (May 2007), page 10a. To recapitulate from the first cited prior article: A WGM resonator of this special type consists of a rod, made of a suitable transparent material, from which protrudes a thin circumferential belt of the same material. The belt is integral with the rest of the rod and acts as a circumferential waveguide. If the depth and width of the belt are made appropriately small, then the belt acts as though it were the core of a single-mode optical fiber: the belt and the rod material adjacent to it support a single, circumferentially propagating mode or family of modes. To recapitulate from the second cited prior article: A major step in the fabrication of a WGM resonator of this special type is diamond turning or computer numerically controlled machining of a rod of a suitable transparent crystalline material on an ultrahigh-precision lathe. During the rotation of a spindle in which the rod is mounted, a diamond tool is used to cut the rod. A computer program is used to control stepping motors that move the diamond tool, thereby controlling the shape cut by the tool. Because the shape can be controlled via software, it is possible to choose a shape designed to optimize a resonator spectrum, including, if desired, to limit the resonator to supporting a single mode. After diamond turning, a resonator can be polished to increase its Q. By virtue of its largely automated, computer-controlled nature, the process is suitable for mass production of nominally identical single-mode WGM resonators. In a demonstration of the capabilities afforded by this development, a number of WGM resonators of various designs were fabricated side by side on the surface of a single CaF2 rod (see figure).

Diversity and evolution of phycobilisomes in marine Synechococcus spp.: a comparative genomics study.

PubMed

Six, Christophe; Thomas, Jean-Claude; Garczarek, Laurence; Ostrowski, Martin; Dufresne, Alexis; Blot, Nicolas; Scanlan, David J; Partensky, Frédéric

2007-01-01

Marine Synechococcus owe their specific vivid color (ranging from blue-green to orange) to their large extrinsic antenna complexes called phycobilisomes, comprising a central allophycocyanin core and rods of variable phycobiliprotein composition. Three major pigment types can be defined depending on the major phycobiliprotein found in the rods (phycocyanin, phycoerythrin I or phycoerythrin II). Among strains containing both phycoerythrins I and II, four subtypes can be distinguished based on the ratio of the two chromophores bound to these phycobiliproteins. Genomes of eleven marine Synechococcus strains recently became available with one to four strains per pigment type or subtype, allowing an unprecedented comparative genomics study of genes involved in phycobilisome metabolism. By carefully comparing the Synechococcus genomes, we have retrieved candidate genes potentially required for the synthesis of phycobiliproteins in each pigment type. This includes linker polypeptides, phycobilin lyases and a number of novel genes of uncharacterized function. Interestingly, strains belonging to a given pigment type have similar phycobilisome gene complements and organization, independent of the core genome phylogeny (as assessed using concatenated ribosomal proteins). While phylogenetic trees based on concatenated allophycocyanin protein sequences are congruent with the latter, those based on phycocyanin and phycoerythrin notably differ and match the Synechococcus pigment types. We conclude that the phycobilisome core has likely evolved together with the core genome, while rods must have evolved independently, possibly by lateral transfer of phycobilisome rod genes or gene clusters between Synechococcus strains, either via viruses or by natural transformation, allowing rapid adaptation to a variety of light niches.

Lead Coolant Test Facility Systems Design, Thermal Hydraulic Analysis and Cost Estimate

DOE Office of Scientific and Technical Information (OSTI.GOV)

Soli Khericha; Edwin Harvego; John Svoboda

2012-01-01

The Idaho National Laboratory prepared a preliminary technical and functional requirements (T&FR), thermal hydraulic design and cost estimate for a lead coolant test facility. The purpose of this small scale facility is to simulate lead coolant fast reactor (LFR) coolant flow in an open lattice geometry core using seven electrical rods and liquid lead or lead-bismuth eutectic coolant. Based on review of current world lead or lead-bismuth test facilities and research needs listed in the Generation IV Roadmap, five broad areas of requirements were identified as listed: (1) Develop and Demonstrate Feasibility of Submerged Heat Exchanger; (2) Develop and Demonstratemore » Open-lattice Flow in Electrically Heated Core; (3) Develop and Demonstrate Chemistry Control; (4) Demonstrate Safe Operation; and (5) Provision for Future Testing. This paper discusses the preliminary design of systems, thermal hydraulic analysis, and simplified cost estimate. The facility thermal hydraulic design is based on the maximum simulated core power using seven electrical heater rods of 420 kW; average linear heat generation rate of 300 W/cm. The core inlet temperature for liquid lead or Pb/Bi eutectic is 4200 C. The design includes approximately seventy-five data measurements such as pressure, temperature, and flow rates. The preliminary estimated cost of construction of the facility is $3.7M (in 2006 $). It is also estimated that the facility will require two years to be constructed and ready for operation.« less

Evaluation of the 3D Finite Element Method Using a Tantalum Rod for Osteonecrosis of the Femoral Head

PubMed Central

Shi, Jingsheng; Chen, Jie; Wu, Jianguo; Chen, Feiyan; Huang, Gangyong; Wang, Zhan; Zhao, Guanglei; Wei, Yibing; Wang, Siqun

2014-01-01

Background The aim of this study was to contrast the collapse values of the postoperative weight-bearing areas of different tantalum rod implant positions, fibula implantation, and core decompression model and to investigate the advantages and disadvantages of tantalum rod implantation in different ranges of osteonecrosis in comparison with other methods. Material/Methods The 3D finite element method was used to establish the 3D finite element model of normal upper femur, 3D finite element model after tantalum rod implantation into different positions of the upper femur in different osteonecrosis ranges, and other 3D finite element models for simulating fibula implant and core decompression. Results The collapse values in the weight-bearing area of the femoral head of the tantalum rod implant model inside the osteonecrosis area, implant model in the middle of the osteonecrosis area, fibula implant model, and shortening implant model exhibited no statistically significant differences (p>0.05) when the osteonecrosis range was small (60°). The stress values on the artificial bone surface for the tantalum rod implant model inside the osteonecrosis area and the shortening implant model exhibited statistical significance (p<0.01). Conclusions Tantalum rod implantation into the osteonecrosis area can reduce the collapse values in the weight-bearing area when osteonecrosis of the femoral head (ONFH) was in a certain range, thereby obtaining better clinical effects. When ONFH was in a large range (120°), the tantalum rod implantation inside the osteonecrosis area, shortening implant or fibula implant can reduce the collapse values of the femoral head, as assessed by other methods. PMID:25479830

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.

Axially shaped channel and integral flow trippers

DOE Office of Scientific and Technical Information (OSTI.GOV)

Crowther, R.L.; Johansson, E.B.; Matzner, B.

1988-06-07

A fuel assembly is described comprising fuel rods positioned in spaced array by upper and lower tie-plates, an open ended flow channel surrounding the array for conducting coolant upward between a lower support plate having coolant communicated thereto to an upper support grid having a steam/water outlet communicated thereto. The flow channel surrounds the array for conducting coolant about the fuel rods. The open ended channel has a polygon shaped cross section with the channel constituting a closed conduit with flat side sections connected at corners to form the enclosed conduit; means separate from the channel for connecting the uppermore » and lower tie-plates together and maintaining the fuel rods in spaced array independent of the flow channel. The improvement in the flow channel comprises tapered side walls. The tapered side walls extend from an average thick cross section adjacent the lower support plate to an average thin cross section adjacent the upper core grid whereby the channel is reduced in thickness adjacent the upper core grid to correspond with the reduced pressure adjacent the upper core grid.« less

Axially shaped channel and integral flow trippers

DOE Office of Scientific and Technical Information (OSTI.GOV)

Crowther, R.L. Jr.; Johansson, E.B.; Matzner, B.

1992-02-11

This patent describes a fuel assembly. It comprises: fuel rods positioned in spaced array by upper and lower tie-plates, and open ended flow channel surrounding the array for conducting coolant upward between a lower support plate having coolant communicated thereto to an upper support grid having a steam/water outlet communicated thereto. The flow channel surrounding the array for conducting coolant about the fuel rods; the open ended channel having a polygon shaped cross section with the channel constituting a closed conduit with flat side sections connected at corners to form the enclosed conduit; means separate from the channel for connectingmore » the upper and lower tie-plates together and maintaining the fuel rods in spaced array independent of the flow channel, the improvement in the flow channel comprising tapered side walls, the tapered side walls extending from an average thick cross section adjacent the lower support plate to an average thin cross section adjacent the upper core grid whereby the channel is reduced in thickness adjacent the upper core grid to correspond with the reduced pressure adjacent the upper core grid.« less

Oriented clonal cell dynamics enables accurate growth and shaping of vertebrate cartilage.

PubMed

Kaucka, Marketa; Zikmund, Tomas; Tesarova, Marketa; Gyllborg, Daniel; Hellander, Andreas; Jaros, Josef; Kaiser, Jozef; Petersen, Julian; Szarowska, Bara; Newton, Phillip T; Dyachuk, Vyacheslav; Li, Lei; Qian, Hong; Johansson, Anne-Sofie; Mishina, Yuji; Currie, Joshua D; Tanaka, Elly M; Erickson, Alek; Dudley, Andrew; Brismar, Hjalmar; Southam, Paul; Coen, Enrico; Chen, Min; Weinstein, Lee S; Hampl, Ales; Arenas, Ernest; Chagin, Andrei S; Fried, Kaj; Adameyko, Igor

2017-04-17

Cartilaginous structures are at the core of embryo growth and shaping before the bone forms. Here we report a novel principle of vertebrate cartilage growth that is based on introducing transversally-oriented clones into pre-existing cartilage. This mechanism of growth uncouples the lateral expansion of curved cartilaginous sheets from the control of cartilage thickness, a process which might be the evolutionary mechanism underlying adaptations of facial shape. In rod-shaped cartilage structures (Meckel, ribs and skeletal elements in developing limbs), the transverse integration of clonal columns determines the well-defined diameter and resulting rod-like morphology. We were able to alter cartilage shape by experimentally manipulating clonal geometries. Using in silico modeling, we discovered that anisotropic proliferation might explain cartilage bending and groove formation at the macro-scale.

Open cycle gas core nuclear rockets

NASA Technical Reports Server (NTRS)

Ragsdale, Robert

1991-01-01

The open cycle gas core engine is a nuclear propulsion device. Propulsion is provided by hot hydrogen which is heated directly by thermal radiation from the nuclear fuel. Critical mass is sustained in the uranium plasma in the center. It has typically 30 to 50 kg of fuel. It is a thermal reactor in the sense that fissions are caused by absorption of thermal neutrons. The fast neutrons go out to an external moderator/reflector material and, by collision, slow down to thermal energy levels, and then come back in and cause fission. The hydrogen propellant is stored in a tank. The advantage of the concept is very high specific impulse because you can take the plasma to any temperature desired by increasing the fission level by withdrawing or turning control rods or control drums.

[Comparison of core decompression with stem cell transplantation and tantalum rod implanting in treating stage II non-traumatic osteonecrosis of femoral head].

PubMed

He, Bang-Jian; Li, Ju; Lyu, Yi; Tong, Pei-Jian

2016-12-25

To compare clinical effects of core decompression with stem cell transplantation and tantalum rod implanting in treating stage II non-traumatic osteonecrosis of femoral head. From March 2012 to September 2012, 45 patients(55 hips)with stage ARCO II non-traumatic osteonecrosis of femoral head were treated and divided into core decompression with stem cell transplantation group(group A) and tantalum rod implanting group(group B) according to number table. In group A, there were 23 cases(28 hips) , including 12 males and 11 females aged from 23 to 51 years old with an average of (36.87±9.52) years, the courses of disease ranged from 2 to 28 months with an average of (17.13±7.74) months, preoperative Harris score was for 35 to 70 with an average of(54.74±11.81), treated with core decompression with stem cell transplantation. In group B, there were 22 cases(27 hips), including 11 males and 11 females aged from 26 to 46 years old with an average of (35.59±7.39) years, the courses of disease ranged from 3 to 26 months with an average of(16.00±7.46) months, preoperative Harris score was for 35 to 76 with an average of (57.18±12.95), treated with core tantalum rod implanting. Operative time, blood loss, hospital stays, hospitalization expenses were observed and compared after treatment between two groups, the clinical effects were evaluated according to Harris criteria. All patients were followed up from 6 to 12 months with an average of 10.8 months. There were significant difference in hospitalization expenses between two groups( P <0.05), while there was no significant statistical difference in blood loss and hospital stay ( P >0.05). At the final following-up, Harris score in group A was(83.04±8.97), 6 cases obtained excellent results, 14 good, 2 good and 1 poor;while Harris score in group A was(84.41±9.94), and 9 cases obtained excellent results, 9 good, 3 good and 1 poor; there was no statistical meaning differences between two groups( P >0.05). Core decompression with stem cell transplantation and tantalum rod implanting could both improve function of hip joint, while core decompression with stem cell transplantation had advantages of shorter operation time, less cost, and higher potency ratio. It is suitable for stage ARCO II non-traumatic femoral head necrosis.

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.

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.

Geotechnical Diver Tools Operation and Maintenance Manual

DTIC Science & Technology

1988-10-01

taking a core with the least amount of 4. Screw piston onto bottom of piston rod; disturbance. The piston is unscrewed from lubricate U-packing seal on...corer head onto piston road. 3. Screw piston onto piston rods check U-pecking seal for proper direction (see manual). 4. Slide aore tube over piston and

Fabrication of Nanosized Island-Like CdO Crystallites-Decorated TiO₂ Rod Nanocomposites via a Combinational Methodology and Their Low-Concentration NO₂ Gas-Sensing Behavior.

PubMed

Liang, Yuan-Chang; Xu, Nian-Cih; Wang, Chein-Chung; Wei, Da-Hua

2017-07-10

TiO₂-CdO composite rods were synthesized through a hydrothermal method and sputtering thin-film deposition. The hydrothermally derived TiO₂ rods exhibited a rectangular cross-sectional crystal feature with a smooth surface, and the as-synthesized CdO thin film exhibited a rounded granular surface feature. Structural analyses revealed that the CdO thin film sputtered onto the surfaces of the TiO₂ rods formed a discontinuous shell layer comprising many island-like CdO crystallites. The TiO₂-CdO composite rods were highly crystalline, and their surfaces were rugged. A comparison of the NO₂ gas-sensing properties of the CdO thin film, TiO₂ rods, and TiO₂-CdO composite rods revealed that the composite rods exhibited superior gas-sensing responses to NO₂ gas than did the CdO thin film and TiO 2 rods, which can be attributed to the microstructural differences and the formation of heterojunctions between the TiO₂ core and CdO crystallites.

MOX fuel assembly design

DOE Office of Scientific and Technical Information (OSTI.GOV)

Reese, A.P.; Crowther, R.L. Jr.

1992-02-18

This patent describes improvement in a boiling water reactor core having a plurality of vertically upstanding fuel bundles; each fuel bundle containing longitudinally extending sealed rods with fissile material therein; the improvement comprises the fissile material including a mixture of uranium and recovered plutonium in rods of the fuel bundle at locations other than the corners of the fuel bundle; and, neutron absorbing material being located in rods of the fuel bundle at rod locations adjacent the corners of the fuel bundles whereby the neutron absorbing material has decreased shielding from the plutonium and maximum exposure to thermal neutrons formore » shaping the cold reactivity shutdown zone in the fuel bundle.« less

Lunar deep drill apparatus

NASA Technical Reports Server (NTRS)

Harvey, Jill (Editor)

1989-01-01

A self contained, mobile drilling and coring system was designed to operate on the Lunar surface and be controlled remotely from earth. The system uses SKITTER (Spatial Kinematic Inertial Translatory Tripod Extremity Robot) as its foundation and produces Lunar core samples two meters long and fifty millimeters in diameter. The drill bit used for this is composed of 30 per carat diamonds in a sintered tungsten carbide matrix. To drill up to 50 m depths, the bit assembly will be attached to a drill string made from 2 m rods which will be carried in racks on SKITTER. Rotary power for drilling will be supplied by a Curvo-Synchronous motor. SKITTER is to support this system through a hexagonal shaped structure which will contain the drill motor and the power supply. A micro-coring drill will be used to remove a preliminary sample 5 mm in diameter and 20 mm long from the side of the core. This whole system is to be controlled from earth. This is carried out by a continuously monitoring PLC onboard the drill rig. A touch screen control console allows the operator on earth to monitor the progress of the operation and intervene if necessary.

PRELIMINARY HAZARDS SUMMARY REPORT FOR THE VALLECITOS SUPERHEAT REACTOR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Murray, J.L.

1961-02-01

BS>The Vallecitos Superheat Reactor (VSR) is a light-watermoderated, thermal-spectrum reactor, cooled by a combination of moderator boiling and forced convection cooling with saturated steam. The reactor core consists of 32 fuel hurdles containing 5300 lb of UO/sub 2/ enriched in U/sub 235/ to 3.6%. The fuel elements are arranged in individual process tubes that direct the cooling steam flow and separate the steam from the water moderator. The reactor vessel is designed for 1250 psig and operates at 960 to 1000 psig. With the reactor operating at 12.5 Mw(t), the maximum fuel cladding temperature is 1250 deg F and themore » cooling steam is superheated to an average temperature of about 810 deg F at 905 psig. Nu clear operation of the reactor is controlled by 12 control rods, actuated by drives mounted on the bottom of the reactor vessel. The water moderator recirculates inside the reactor vessel and through the core region by natural convection. Inherent safety features of the reactor include the negative core reactivity effects upon heating the UO/sub 2/ fuel (Doppler effect), upon increasing the temperature or void content of the moderator in the operating condition, and upon unflooding the fuel process tubes in the hot condition. Snfety features designed into the reactor and plant systems include a system of sensors and devices to detect petentially unsafe operating conditions and to initiate automatically the appropriate countermeasures, a set of fast and reliable control rods for scramming the reactor if a potentially unsafe condition occurs, a manually-actuated liquid neutron poison system, and an emergency cooling system to provide continued steam flow through the reactor core in the event the reactor becomes isolated from either its normal source of steam supply or discharge. The release of radioactivity to unrestricted areas is maintained within permissible limits by monitoring the radioactivity of wastes and controlling their release. The reactor and many of its auxiliaries are housed within a high-integrity essentially leak-tight containment vessel. (auth)« less

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.

Design of cladding rods-assisted depressed-core few-mode fibers with improved modal spacing

NASA Astrophysics Data System (ADS)

Han, Jiawei; Zhang, Jie

2018-03-01

This paper investigates the design details of cladding rods-assisted (CRA) depressed-core (DC) few-mode fibers (FMFs) that feature more equally spaced linearly polarized (LP) modal effective indices, suitable for high-spatial-density weakly-coupled mode-division multiplexing systems. The influences of the index profile of cladding rods on LP mode-resolved effective index, bending sensitivity, and effective area Aeff, are numerically described. Based on the design considerations of LP modal Aeff-dependent spatial efficiency and LP modal bending loss-dependent robustness, the small LP21-LP02 and LP22-LP03 modal spacing limitations, encountered in state-of-the-art weakly-coupled step-index FMFs, have been substantially improved by at least 25%. In addition, the proposed CRA DC FMFs also show sufficiently large effective areas (in excess of 110 μm2) for all guided LP modes, which are expected to exhibit good nonlinear performance.

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.

Analytical analyses of startup measurements associated with the first use of LEU fuel in Romania`s 14-MW TRIGA reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bretscher, M.M.; Snelgrove, J.L.; Ciocanescu, M.

1992-12-01

The 14-MW TRIGA steady state reactor (SSR) is located in Pitesti, Romania. Beginning with an HEU core (10 wt% U), the reactor first went critical in November 1979 but was shut down ten years later because of insufficient excess reactivity. Last November the Institute for Nuclear Research (INR), which operates the SSR, received from the ANL RERTR program a shipment of 125 LEU pins fabricated by General Atomics and of the same geometry as the original fuel but with an enrichment of 19.7% 235U and a loading of 45 wt% U. Using 100 of these pins, four LEU clusters, eachmore » containing a 5 x 5 square array of fuel rods, were assembled. These four LEU clusters replaced the four most highly burned HEU elements in the SSR. The reactor resumed operations last February with a 35-element mixed HEU/LEU core configuration. In preparation for full power operation of the SSR with this mixed HEU/LEU core, a number of measurements were made. These included control rod calibrations, excess reactivity determinations, worths of experiment facilities, reaction rate distributions, and themocouple measurements of fuel temperatures as a function of reactor power. This paper deals with a comparison of some of these measured reactor parameters with corresponding analytical calculations.« less

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)

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)

Mechanical design of a light water breeder reactor

DOEpatents

Fauth, Jr., William L.; Jones, Daniel S.; Kolsun, George J.; Erbes, John G.; Brennan, John J.; Weissburg, James A.; Sharbaugh, John E.

1976-01-01

In a light water reactor system using the thorium-232 -- uranium-233 fuel system in a seed-blanket modular core configuration having the modules arranged in a symmetrical array surrounded by a reflector blanket region, the seed regions are disposed for a longitudinal movement between the fixed or stationary blanket region which surrounds each seed region. Control of the reactor is obtained by moving the inner seed region thus changing the geometry of the reactor, and thereby changing the leakage of neutrons from the relatively small seed region into the blanket region. The mechanical design of the Light Water Breeder Reactor (LWBR) core includes means for axially positioning of movable fuel assemblies to achieve the neutron economy required of a breeder reactor, a structure necessary to adequately support the fuel modules without imposing penalties on the breeding capability, a structure necessary to support fuel rods in a closely packed array and a structure necessary to direct and control the flow of coolant to regions in the core in accordance with the heat transfer requirements.

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)

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.

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.

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.

Verification of Three Dimensional Triangular Prismatic Discrete Ordinates Transport Code ENSEMBLE-TRIZ by Comparison with Monte Carlo Code GMVP

NASA Astrophysics Data System (ADS)

Homma, Yuto; Moriwaki, Hiroyuki; Ohki, Shigeo; Ikeda, Kazumi

2014-06-01

This paper deals with verification of three dimensional triangular prismatic discrete ordinates transport calculation code ENSEMBLE-TRIZ by comparison with multi-group Monte Carlo calculation code GMVP in a large fast breeder reactor. The reactor is a 750 MWe electric power sodium cooled reactor. Nuclear characteristics are calculated at beginning of cycle of an initial core and at beginning and end of cycle of equilibrium core. According to the calculations, the differences between the two methodologies are smaller than 0.0002 Δk in the multi-plication factor, relatively about 1% in the control rod reactivity, and 1% in the sodium void reactivity.

Minimizing or eliminating refueling of nuclear reactor

DOEpatents

Doncals, Richard A.; Paik, Nam-Chin; Andre, Sandra V.; Porter, Charles A.; Rathbun, Roy W.; Schwallie, Ambrose L.; Petras, Diane S.

1989-01-01

Demand for refueling of a liquid metal fast nuclear reactor having a life of 30 years is eliminated or reduced to intervals of at least 10 years by operating the reactor at a low linear-power density, typically 2.5 kw/ft of fuel rod, rather than 7.5 or 15 kw/ft, which is the prior art practice. So that power of the same magnitude as for prior art reactors is produced, the volume of the core is increased. In addition, the height of the core and it diameter are dimensioned so that the ratio of the height to the diameter approximates 1 to the extent practicable considering the requirement of control and that the pressure drop in the coolant shall not be excessive. The surface area of a cylinder of given volume is a minimum if the ratio of the height to the diameter is 1. By minimizing the surface area, the leakage of neutrons is reduced. By reducing the linear-power density, increasing core volume, reducing fissile enrichment and optimizing core geometry, internal-core breeding of fissionable fuel is substantially enhanced. As a result, core operational life, limited by control worth requirements and fuel burnup capability, is extended up to 30 years of continuous power operation.

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.

Development of advanced strain diagnostic techniques for reactor environments.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fleming, Darryn D.; Holschuh, Thomas Vernon,; Miller, Timothy 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.more » 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.« less

CRDM with separate SCRAM latch engagement and locking

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dodd, Christopher D.; DeSantis, Paul K.; Stambaugh, Kevin J.

A control rod drive mechanism (CRDM) configured to latch onto the lifting rod of a control rod assembly and including separate latch engagement and latch holding mechanisms. A CRDM configured to latch onto the lifting rod of a control rod assembly and including a four-bar linkage closing the latch, wherein the four-bar linkage biases the latch closed under force of gravity.

30. Engine controls and valve gear, looking aft on main ...

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

30. Engine controls and valve gear, looking aft on main (promenade) deck level. Threaded admission valve lift rods (two at immediate left of chronometer) permit adjustment of valve timing in lower and upper admission valves of cylinder (left rod controls lower valve, right rod upper valve). Valve rods are lifted by jaw-like "wipers" during operation. Exhaust valve lift rods and wipers are located to right of chronometer. Crank at extreme right drives valve wiper shaft when engaged to end of eccentric rod, shown under "Crank Indicator" dial. Pair of handles to immediate left of admission valve rods control condenser water valves; handles to right of exhaust valve rods control feedwater flow to boilers from pumps. Gauges indicate boiler pressure (left) and condenser vacuum (right); "Crank Indicator" on wall aids engineer in keeping engine crank off "dead-center" at stop so that engine may be easily restarted. - Steamboat TICONDEROGA, Shelburne Museum Route 7, Shelburne, Chittenden County, VT

A COMPARISON OF EXPERIMENTS AND THREE-DIMENSIONAL ANALYSIS TECHNIQUES. PART II. UNPOISONED UNIFORM SLAB CORE WITH A PARTIALLY INSERTED HAFNIUM ROD AND A PARTIALLY INSERTED WATER GAP

DOE Office of Scientific and Technical Information (OSTI.GOV)

Roseberry, R.J.

The experimental measurements and nuclear analysis of a uniformly loaded, unpoisoned slab core with a partially inserted hafnium rod and/or a partially inserted water gap are described. Comparisons of experimental data with calculated results of the UFO core and flux synthesis techniques are given. It is concluded that one of the flux synthesis techniques and the UFO code are able to predict flux distributions to within approximately -5% of experiment for most cases, with a maximum error of approximately -10% for a channel at the core- reflector boundary. The second synthesis technique failed to give comparable agreement with experiment evenmore » when various refinements were used, e.g. increasing the number of mesh points, performing the flux synthesis technique of iteration, and spectrum-weighting the appropriate calculated fluxes through the use of the SWAKRAUM code. These results are comparable to those reported in Part I of this study. (auth)« less

Impact of thermal spectrum small modular reactors on performance of once-through nuclear fuel cycles with low-enriched uranium

DOE PAGES

Brown, Nicholas R.; Worrall, Andrew; Todosow, Michael

2016-11-18

Small modular reactors (SMRs) offer potential benefits, such as enhanced operational flexibility. However, it is vital to understand the holistic impact of SMRs on nuclear fuel cycle performance. The focus of this paper is the fuel cycle impacts of light water SMRs in a once-through fuel cycle with low-enriched uranium fuel. A key objective of this paper is to describe preliminary example reactor core physics and fuel cycle analyses conducted in support of the U.S. Department of Energy, Office of Nuclear Energy, Fuel Cycle Options Campaign. The hypothetical light water SMR example case considered in these preliminary scoping studies ismore » a cartridge type one-batch core with slightly less than 5.0% enrichment. Challenges associated with SMRs include increased neutron leakage, fewer assemblies in the core (and therefore fewer degrees of freedom in the core design), complex enrichment and burnable absorber loadings, full power operation with inserted control rods, the potential for frequent load-following operation, and shortened core height. Each of these will impact the achievable discharge burnup in the reactor and the fuel cycle performance. This paper summarizes a list of the factors relevant to SMR fuel, core, and operation that will impact fuel cycle performance. The high-level issues identified and preliminary scoping calculations in this paper are intended to inform on potential fuel cycle impacts of one-batch thermal spectrum SMRs. In particular, this paper highlights the impact of increased neutron leakage and reduced number of batches on the achievable burnup of the reactor. Fuel cycle performance metrics for a hypothetical example SMR are compared with those for a conventional three-batch light water reactor in the following areas: nuclear waste management, environmental impact, and resource utilization. The metrics performance for such an SMR is degraded for the mass of spent nuclear fuel and high-level waste disposed of, mass of depleted uranium disposed of, land use per energy generated, and carbon emissions per energy generated. Finally, it is noted that the features of some SMR designs impact three main aspects of fuel cycle performance: (1) small cores which means high leakage (there is a radial and axial component), (2) no boron which means heterogeneous core and extensive use of control rods and BPs, and (3) single batch cores. But not all of the SMR designs have all of these traits. As a result, the approach used in this study is therefore a bounding case and not all SMRs may be affected to the same extent.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brown, Nicholas R.; Worrall, Andrew; Todosow, Michael

Small modular reactors (SMRs) offer potential benefits, such as enhanced operational flexibility. However, it is vital to understand the holistic impact of SMRs on nuclear fuel cycle performance. The focus of this paper is the fuel cycle impacts of light water SMRs in a once-through fuel cycle with low-enriched uranium fuel. A key objective of this paper is to describe preliminary example reactor core physics and fuel cycle analyses conducted in support of the U.S. Department of Energy, Office of Nuclear Energy, Fuel Cycle Options Campaign. The hypothetical light water SMR example case considered in these preliminary scoping studies ismore » a cartridge type one-batch core with slightly less than 5.0% enrichment. Challenges associated with SMRs include increased neutron leakage, fewer assemblies in the core (and therefore fewer degrees of freedom in the core design), complex enrichment and burnable absorber loadings, full power operation with inserted control rods, the potential for frequent load-following operation, and shortened core height. Each of these will impact the achievable discharge burnup in the reactor and the fuel cycle performance. This paper summarizes a list of the factors relevant to SMR fuel, core, and operation that will impact fuel cycle performance. The high-level issues identified and preliminary scoping calculations in this paper are intended to inform on potential fuel cycle impacts of one-batch thermal spectrum SMRs. In particular, this paper highlights the impact of increased neutron leakage and reduced number of batches on the achievable burnup of the reactor. Fuel cycle performance metrics for a hypothetical example SMR are compared with those for a conventional three-batch light water reactor in the following areas: nuclear waste management, environmental impact, and resource utilization. The metrics performance for such an SMR is degraded for the mass of spent nuclear fuel and high-level waste disposed of, mass of depleted uranium disposed of, land use per energy generated, and carbon emissions per energy generated. Finally, it is noted that the features of some SMR designs impact three main aspects of fuel cycle performance: (1) small cores which means high leakage (there is a radial and axial component), (2) no boron which means heterogeneous core and extensive use of control rods and BPs, and (3) single batch cores. But not all of the SMR designs have all of these traits. As a result, the approach used in this study is therefore a bounding case and not all SMRs may be affected to the same extent.« less

Yb3+-doped rod-type amplifiers with local adiabatic tapers for peak power scaling and beam quality improvement

NASA Astrophysics Data System (ADS)

Zhu, Yuan; Eschrich, Tina; Leich, Martin; Grimm, Stephan; Kobelke, Jens; Lorenz, Martin; Bartelt, Hartmut; Jäger, Matthias

2017-10-01

The use of short local tapers in large mode area fiber amplifiers is proposed for peak power scaling while maintaining good beam quality. To avoid modal distortions, the powder-sintering (REPUSIL) method was employed to obtain core materials with excellent refractive index homogeneity. First experiments with Yb3+-doped rod-type amplifiers delivered 2 ns pulses with peak powers of 540 kW and energies of 1.4 mJ for the untapered rod and 230 kW for the tapered rod (limited by facet damage). The beam quality improved from an M 2 value of approximately 10 to 3.5. The investigation of the taper structure indicates room for further improvement.

Parameter study on the influence of prepressurization on PWR fuel rod behavior during normal operation and hypothetical LOCAs

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brzoska, B.; Depisch, F.; Fuchs, H.P.

To analyze the influence of prepressurization on fuel rod behavior, a parametric study has been performed that considers the effects of as-fabricated fuel rod internal prepressure on the normal operation and postulated loss-of-coolant accident (LOCA) rod behavior of a 1300-MW(electric) Kraftwerk Union (KWU) standard pressurized water reactor nuclear power plant. A variation of the prepressure in the range from 15 to 35 bars has only a slight influence on normal operation behavior. Considering the LOCA behavior, only a small temperature increase results from prepressure reduction, while the core-wide straining behavior is improved significantly. The KWU prepressurization takes both conditions intomore » account.« less

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.

Experimental verification of a theoretical model of an active cladding optical fiber fluorosensor

NASA Technical Reports Server (NTRS)

Albin, Sacharia; Briant, Alvin L.; Egalon, Claudio O.; Rogowski, Robert S.; Nankung, Juock S.

1993-01-01

Experiments were conducted to verify a theoretical model on the injection efficiency of sources in the cladding of an optical fiber. The theoretical results predicted an increase in the injection efficiency for higher differences in refractive indices between the core and cladding. The experimental apparatus used consisted of a glass rod 50 cm long, coated at one end with a thin film of fluorescent substance. The fluorescent substance was excited with side illumination, perpendicular to the rod axis, using a 476 nm Argon-ion laser. Part of the excited fluorescence was injected into the core and guided to a detector. The signal was measured for several different cladding refractive indices. The cladding consisted of sugar dissolved in water and the refractive index was changed by varying the sugar concentration in the solution. The results indicate that the power injected into the rod, due to evanescent wave injection, increases with the difference in refractive index which is in qualitative agreement with theory.

Calculation of the Phenix end-of-life test 'Control Rod Withdrawal' with the ERANOS code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tiberi, V.

2012-07-01

The Inst. of Radiological Protection and Nuclear Safety (IRSN) acts as technical support to French public authorities. As such, IRSN is in charge of safety assessment of operating and under construction reactors, as well as future projects. In this framework, one current objective of IRSN is to evaluate the ability and accuracy of numerical tools to foresee consequences of accidents. Neutronic studies step in the safety assessment from different points of view among which the core design and its protection system. They are necessary to evaluate the core behavior in case of accident in order to assess the integrity ofmore » the first barrier and the absence of a prompt criticality risk. To reach this objective one main physical quantity has to be evaluated accurately: the neutronic power distribution in core during whole reactor lifetime. Phenix end of life tests, carried out in 2009, aim at increasing the experience feedback on sodium cooled fast reactors. These experiments have been done in the framework of the development of the 4. generation of nuclear reactors. Ten tests have been carried out: 6 on neutronic and fuel aspects, 2 on thermal hydraulics and 2 for the emergency shutdown. Two of them have been chosen for an international exercise on thermal hydraulics and neutronics in the frame of an IAEA Coordinated Research Project. Concerning neutronics, the Control Rod Withdrawal test is relevant for safety because it allows evaluating the capability of calculation tools to compute the radial power distribution on fast reactors core configurations in which the flux field is very deformed. IRSN participated to this benchmark with the ERANOS code developed by CEA for fast reactors studies. This paper presents the results obtained in the framework of the benchmark activity. A relatively good agreement was found with available measures considering the approximations done in the modeling. The work underlines the importance of burn-up calculations in order to have a fine core concentrations mesh for the calculation of the power distribution. (authors)« less

Modified rod-in-tube for high-NA tellurite glass fiber fabrication: materials and technologies.

PubMed

Chen, Qiuling; Wang, Hui; Wang, Qingwei; Chen, Qiuping; Hao, Yinlei

2015-02-01

In this paper, we report the whole fabrication process for high-numerical aperture (NA) tellurite glass fibers from material preparation to preform fabrication, and eventually, fiber drawing. A tellurite-based high-NA (0.9) magneto-optical glass fiber was drawn successfully and characterized. First, matchable core and cladding glasses were fabricated and matched in terms of physical properties. Second, a uniform bubble-free preform was fabricated by means of a modified rod-in-tube technique. Finally, the fiber drawing process was studied and optimized. The high-NA fibers (∅(core), 40-50 μm and ∅(cladding), 120-130 μm) so obtained were characterized for their geometrical and optical properties.

Oriented clonal cell dynamics enables accurate growth and shaping of vertebrate cartilage

PubMed Central

Kaucka, Marketa; Zikmund, Tomas; Tesarova, Marketa; Gyllborg, Daniel; Hellander, Andreas; Jaros, Josef; Kaiser, Jozef; Petersen, Julian; Szarowska, Bara; Newton, Phillip T; Dyachuk, Vyacheslav; Li, Lei; Qian, Hong; Johansson, Anne-Sofie; Mishina, Yuji; Currie, Joshua D; Tanaka, Elly M; Erickson, Alek; Dudley, Andrew; Brismar, Hjalmar; Southam, Paul; Coen, Enrico; Chen, Min; Weinstein, Lee S; Hampl, Ales; Arenas, Ernest; Chagin, Andrei S; Fried, Kaj; Adameyko, Igor

2017-01-01

Cartilaginous structures are at the core of embryo growth and shaping before the bone forms. Here we report a novel principle of vertebrate cartilage growth that is based on introducing transversally-oriented clones into pre-existing cartilage. This mechanism of growth uncouples the lateral expansion of curved cartilaginous sheets from the control of cartilage thickness, a process which might be the evolutionary mechanism underlying adaptations of facial shape. In rod-shaped cartilage structures (Meckel, ribs and skeletal elements in developing limbs), the transverse integration of clonal columns determines the well-defined diameter and resulting rod-like morphology. We were able to alter cartilage shape by experimentally manipulating clonal geometries. Using in silico modeling, we discovered that anisotropic proliferation might explain cartilage bending and groove formation at the macro-scale. DOI: http://dx.doi.org/10.7554/eLife.25902.001 PMID:28414273

Strike action electromagnetic machine for immersion of rod elements into ground

NASA Astrophysics Data System (ADS)

Usanov, K. M.; Volgin, A. V.; Chetverikov, E. A.; Kargin, V. A.; Moiseev, A. P.; Ivanova, Z. I.

2017-10-01

During construction, survey work, and drilling shallow wells by striking, operations associated with dipping and removing the rod elements are the most common. At the same time relatively long, with small diameter, elements, in which the ratio of length to diameter l/d is 100 or more, constitute a significant proportion. At present, the application of power pulse linear electromagnetic motors to drive drum machines is recognized to be highly effective. However, the mechanical method of transmission of shocks does not allow dipping long longitudinally unstable core elements. In this case, mechanical energy must be transferred from the motor to the rod through its side surface. The design of the strike action electromagnetic machine with a through axial channel for non-mechanical end striking of the pile of long, longitudinally unstable metal rods is proposed. Electromagnetic striking machine for non-mechanical end striking rod elements provides operations characterized by ecological compatibility, safety and high quality.

A solid reactor core thermal model for nuclear thermal rockets

NASA Astrophysics Data System (ADS)

Rider, William J.; Cappiello, Michael W.; Liles, Dennis R.

1991-01-01

A Helium/Hydrogen Cooled Reactor Analysis (HERA) computer code has been developed. HERA has the ability to model arbitrary geometries in three dimensions, which allows the user to easily analyze reactor cores constructed of prismatic graphite elements. The code accounts for heat generation in the fuel, control rods, and other structures; conduction and radiation across gaps; convection to the coolant; and a variety of boundary conditions. The numerical solution scheme has been optimized for vector computers, making long transient analyses economical. Time integration is either explicit or implicit, which allows the use of the model to accurately calculate both short- or long-term transients with an efficient use of computer time. Both the basic spatial and temporal integration schemes have been benchmarked against analytical solutions.

MC21 analysis of the MIT PWR benchmark: Hot zero power results

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kelly Iii, D. J.; Aviles, B. N.; Herman, B. R.

2013-07-01

MC21 Monte Carlo results have been compared with hot zero power measurements from an operating pressurized water reactor (PWR), as specified in a new full core PWR performance benchmark from the MIT Computational Reactor Physics Group. Included in the comparisons are axially integrated full core detector measurements, axial detector profiles, control rod bank worths, and temperature coefficients. Power depressions from grid spacers are seen clearly in the MC21 results. Application of Coarse Mesh Finite Difference (CMFD) acceleration within MC21 has been accomplished, resulting in a significant reduction of inactive batches necessary to converge the fission source. CMFD acceleration has alsomore » been shown to work seamlessly with the Uniform Fission Site (UFS) variance reduction method. (authors)« less

POSITIONING DEVICE

DOEpatents

McCorkle, W.H.

1959-07-14

A positioner for a control rod for a nuclear reactor is described. The positioner includes a spur gear and rack for adjusting the control rod slowly and in small ainounts as well as a piston and cylinder for moving the control rod rapidly thrcugh larger distances. The positioner also has associsted with it a worm wheel and gear for rotating it out of engagement with the control rod.

Best estimate plus uncertainty analysis of departure from nucleate boiling limiting case with CASL core simulator VERA-CS in response to PWR main steam line break event

DOE PAGES

Brown, Cameron S.; Zhang, Hongbin; Kucukboyaci, Vefa; ...

2016-09-07

VERA-CS (Virtual Environment for Reactor Applications, Core Simulator) is a coupled neutron transport and thermal-hydraulics subchannel code under development by the Consortium for Advanced Simulation of Light Water Reactors (CASL). VERA-CS was used to simulate a typical pressurized water reactor (PWR) full core response with 17x17 fuel assemblies for a main steam line break (MSLB) accident scenario with the most reactive rod cluster control assembly stuck out of the core. The accident scenario was initiated at the hot zero power (HZP) at the end of the first fuel cycle with return to power state points that were determined by amore » system analysis code and the most limiting state point was chosen for core analysis. The best estimate plus uncertainty (BEPU) analysis method was applied using Wilks’ nonparametric statistical approach. In this way, 59 full core simulations were performed to provide the minimum departure from nucleate boiling ratio (MDNBR) at the 95/95 (95% probability with 95% confidence level) tolerance limit. The results show that this typical PWR core remains within MDNBR safety limits for the MSLB accident.« less

Best estimate plus uncertainty analysis of departure from nucleate boiling limiting case with CASL core simulator VERA-CS in response to PWR main steam line break event

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brown, Cameron S.; Zhang, Hongbin; Kucukboyaci, Vefa

VERA-CS (Virtual Environment for Reactor Applications, Core Simulator) is a coupled neutron transport and thermal-hydraulics subchannel code under development by the Consortium for Advanced Simulation of Light Water Reactors (CASL). VERA-CS was used to simulate a typical pressurized water reactor (PWR) full core response with 17x17 fuel assemblies for a main steam line break (MSLB) accident scenario with the most reactive rod cluster control assembly stuck out of the core. The accident scenario was initiated at the hot zero power (HZP) at the end of the first fuel cycle with return to power state points that were determined by amore » system analysis code and the most limiting state point was chosen for core analysis. The best estimate plus uncertainty (BEPU) analysis method was applied using Wilks’ nonparametric statistical approach. In this way, 59 full core simulations were performed to provide the minimum departure from nucleate boiling ratio (MDNBR) at the 95/95 (95% probability with 95% confidence level) tolerance limit. The results show that this typical PWR core remains within MDNBR safety limits for the MSLB accident.« less

Synthesis and Microstructural Characterization of Manganese Oxide Electrodes for Application as Electrochemical Supercapacitors

NASA Astrophysics Data System (ADS)

Babakhani, Banafsheh

The aim of this thesis work was to synthesize Mn-based oxide electrodes with high surface area structures by anodic electrodeposition for application as electrochemical capacitors. Rod-like structures provide large surface areas leading to high specific capacitances. Since templated electrosynthesis of rods is not easy to use in practical applications, it is more desirable to form rod-like structures without using any templates. In this work, Mn oxide electrodes with rod-like structures (˜1.5 µm in diameter) were synthesized from a solution of 0.01 M Mn acetate under galvanostatic control without any templates, on Au coated Si substrates. The electrochemical properties of the synthesized nanocrystalline electrodes were investigated to determine the effect of morphology, chemistry and crystal structure on the corresponding electrochemical behavior of Mn oxide electrodes. Mn oxides prepared at different current densities showed a defective antifluoritetype crystal structure. The rod-like Mn oxide electrodes synthesized at low current densities (5 mAcm.2) exhibited a high specific capacitance due to their large surface areas. Also, specific capacity retention after 250 cycles in an aqueous solution of 0.5 M Na2SO4 at 100 mVs -1 was about 78% of the initial capacity (203 Fg-1 ). To improve the electrochemical capacitive behavior of Mn oxide electrodes, a sequential approach and a one-step method were adopted to synthesize Mn oxide/PEDOT electrodes through anodic deposition on Au coated Si substrates from aqueous solutions. In the former case, free standing Mn oxide rods (about 10 µm long and less than 1.5 µm in diameter) were first synthesized, then coated by electro-polymerization of a conducting polymer (PEDOT) giving coaxial rods. The one-step, co-electrodeposition method produced agglomerated Mn oxide/PEDOT particles. The electrochemical behavior of the deposits depended on the morphology and crystal structure of the fabricated electrodes, which were affected by the composition and pH of the electrolyte, temperature, current density and polymer deposition time. Mn oxide/PEDOT coaxial core/shell rods consisted of MnO2 with an antifluorite-type structure coated with amorphous PEDOT. The Mn oxide/PEDOT coaxial core/shell electrodes prepared by the sequential method showed significantly better specific capacity and redox performance properties relative to both uncoated Mn oxide rods and co- electrodeposited Mn oxide/PEDOT electrodes. The best specific capacitance for Mn oxide/PEDOT rods produced sequentially was ˜295 F g-1 with ˜92% retention after 250 cycles in 0.5 M Na2SO4 at 100 mV s-1. To further improve the electrochemical capacitive behavior of Mn oxide electrodes, Co-doped and Fe-doped Mn oxide electrodes with a rod-like morphology and antifluorite-type crystal structure were synthesized by anodic electrodeposition, on Au coated Si substrates, from dilute solutions of Mn acetate and Co sulphate and Mn acetate and Fe chloride. Also, Mn-Co oxide/PEDOT coaxial core/shell rods were synthesized by applying a shell of PEDOT on Mn-Co oxide electrodes. Mn-Co oxide/PEDOT electrodes consisted of MnO2, with partial Co 2+ and Co3+ ion substitution for Mn4+, and amorphous PEDOT. Mn-Fe oxide electrodes consisted of MnO2, with partial Fe2+ and Fe3+ ion substitution for Mn4+. Electrochemical analysis showed that the capacitance values for all deposits increased with increasing scan rate to 100 mVs -1, and then decreased after 100 mVs-1. The Mn-Co oxide/PEDOT electrodes showed improved specific capacity and electrochemical cyclability relative to uncoated Mn-Co oxides and Mn-Fe oxides. Mn-Co oxide/PEDOT electrodes with rod-like structures had high capacitances (up to 310 Fg -1) at a scan rate of 100 mVs-1 and maintained their capacitance after 500 cycles in 0.5 M Na2SO4 (91% retention). Capacitance reduction for the deposits was mainly due to the loss of Mn ions by dissolution in the electrolyte solution. To better understand the nucleation and growth mechanisms of Mn oxide electrodes, the effects of supersaturation ratio on the morphology and crystal structure of electrodeposited Mn oxide were studied. By changing deposition parameters, including deposition current density, electrolyte composition, pH and temperature, a series of nanocrystalline Mn oxide electrodes with various morphologies (continuous coatings, rod-like structures, aggregated rods and thin sheets) and an antifluorite-type crystal structure was obtained. Mn oxide thin sheets showed instantaneous nucleation and single crystalline growth; rods had a mix of instantaneous/progressive nucleation and polycrystalline growth and continuous coatings formed by progressive nucleation and polycrystalline growth. Electrochemical analysis revealed the best capacitance behaviour obtained for Mn oxide thin sheets followed by Mn oxide rods, with dimensions on the microscale, and then continuous coatings. The highest specific capacitance (˜230 Fg-1) and capacitance retention rates (˜88%) were obtained for Mn oxide thin sheets after 250 cycles in 0.5 M Na2 SO4 at 20 mVs-1.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sen, Ramazan Sonat; Hummel, Andrew John; Hiruta, Hikaru

The deterministic full core simulators require homogenized group constants covering the operating and transient conditions over the entire lifetime. Traditionally, the homogenized group constants are generated using lattice physics code over an assembly or block in the case of prismatic high temperature reactors (HTR). For the case of strong absorbers that causes strong local depressions on the flux profile require special techniques during homogenization over a large volume. Fuel blocks with burnable poisons or control rod blocks are example of such cases. Over past several decades, there have been a tremendous number of studies performed for improving the accuracy ofmore » full-core calculations through the homogenization procedure. However, those studies were mostly performed for light water reactor (LWR) analyses, thus, may not be directly applicable to advanced thermal reactors such as HTRs. This report presents the application of SuPer-Homogenization correction method to a hypothetical HTR core.« less

Updates to the Generation of Physics Data Inputs for MAMMOTH Simulations of the Transient Reactor Test Facility - FY2016

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ortensi, Javier; Baker, Benjamin Allen; Schunert, Sebastian

The INL is currently evolving the modeling and simulation (M&S) capability that will enable improved core operation as well as design and analysis of TREAT experiments. This M&S capability primarily uses MAMMOTH, a reactor physics application being developed under Multi-physics Object Oriented Simulation Environment (MOOSE) framework. MAMMOTH allows the coupling of a number of other MOOSE-based applications. This second year of work has been devoted to the generation of a deterministic reference solution for the full core, the preparation of anisotropic diffusion coefficients, the testing of the SPH equivalence method, and the improvement of the control rod modeling. In addition,more » this report includes the progress made in the modeling of the M8 core configuration and experiment vehicle since January of this year.« less

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

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 cladmore » melting at 1370/sup 0/C.« less

Transition from HEU to LEU fuel in Romania`s 14-MW TRIGA reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bretscher, M.M.; Snelgrove, J.L.

1991-12-31

The 14-MW TRIGA steady state reactor (SSR) located in Pitesti, Romania, first went critical in the fall of 1979. Initially, the core configuration for full power operation used 29 fuel clusters each containing a 5 {times} 5 square array of HEU (10 wt%) -- ZrH -- Er (2.8 wt%) fuel-moderator rods (1.295 cm o.d.) clad in Incology. With a total inventory of 35 HEU fuel clusters, burnup considerations required a gradual expansion of the core from 29 to 32 and finally to 35 clusters before the reactor was shut down because of insufficient excess reactivity. At this time each ofmore » the original 29 fuel clusters had an overage {sup 235}U burnup in the range from 50 to 62%. Because of the US policy regarding the export of highly enriched uranium, fresh HEU TRIGA replacement fuel is not available. After a number of safety-related measurements, the SSR is expected to resume full power operation in the near future using a mixed core containing five LEU TRIGA clusters of the same geometry as the original fuel but with fuel-moderator rods containing 45 wt% U (19.7% {sup 235}U enrichment) and 1.1 wt% Er. Rods for 14 additional LEU fuel clusters will be fabricated by General Atomics. In support of the SSR mixed core operation numerous neutronic calculations have been performed. This paper presents some of the results of those calculations.« less

Structure and Composition of Isolated Core-Shell (In ,Ga )N /GaN Rods Based on Nanofocus X-Ray Diffraction and Scanning Transmission Electron Microscopy

NASA Astrophysics Data System (ADS)

Krause, Thilo; Hanke, Michael; Nicolai, Lars; Cheng, Zongzhe; Niehle, Michael; Trampert, Achim; Kahnt, Maik; Falkenberg, Gerald; Schroer, Christian G.; Hartmann, Jana; Zhou, Hao; Wehmann, Hergo-Heinrich; Waag, Andreas

2017-02-01

Nanofocus x-ray diffraction is used to investigate the structure and local strain field of an isolated (In ,Ga )N /GaN core-shell microrod. Because the high spatial resolution of the x-ray beam is only 80 ×90 nm2, we are able to investigate several distinct volumes on one individual side facet. Here, we find a drastic increase in thickness of the outer GaN shell along the rod height. Additionally, we performed high-angle annular dark-field scanning-transmission-electron-microscopy measurements on several rods from the same sample showing that (In,Ga)N double-quantum-well and GaN barrier thicknesses also increase strongly along the height. Moreover, plastic relaxation is observed in the top part of the rod. Based on the experimentally obtained structural parameters, we simulate the strain-induced deformation using the finite-element method, which serves as the input for subsequent kinematic scattering simulations. The simulations reveal a significant increase of elastic in-plane relaxation along the rod height. However, at a certain height, the occurrence of plastic relaxation yields a decrease of the elastic strain. Because of the experimentally obtained structural input for the finite-element simulations, we can exclude unknown structural influences on the strain distribution, and we are able to translate the elastic relaxation into an indium concentration which increases by a factor of 4 from the bottom to the height where plastic relaxation occurs.

Depletion optimization of lumped burnable poisons in pressurized water reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kodah, Z.H.

1982-01-01

Techniques were developed to construct a set of basic poison depletion curves which deplete in a monotonical manner. These curves were combined to match a required optimized depletion profile by utilizing either linear or non-linear programming methods. Three computer codes, LEOPARD, XSDRN, and EXTERMINATOR-2 were used in the analyses. A depletion routine was developed and incorporated into the XSDRN code to allow the depletion of fuel, fission products, and burnable poisons. The Three Mile Island Unit-1 reactor core was used in this work as a typical PWR core. Two fundamental burnable poison rod designs were studied. They are a solidmore » cylindrical poison rod and an annular cylindrical poison rod with water filling the central region.These two designs have either a uniform mixture of burnable poisons or lumped spheroids of burnable poisons in the poison region. Boron and gadolinium are the two burnable poisons which were investigated in this project. Thermal self-shielding factor calculations for solid and annular poison rods were conducted. Also expressions for overall thermal self-shielding factors for one or more than one size group of poison spheroids inside solid and annular poison rods were derived and studied. Poison spheroids deplete at a slower rate than the poison mixture because each spheroid exhibits some self-shielding effects of its own. The larger the spheroid, the higher the self-shielding effects due to the increase in poison concentration.« less

This group view shows propellant preparation buidling 4241/E42, 4242/E43, and ...

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

This group view shows propellant preparation buidling 4241/E-42, 4242/E-43, and northwest (314 degrees). Note warning lights at the extreme left of the view, and the use of lightning rods on structures. Building 4241/E-42 housed solid rocket motors after they were cast and awaiting curing. Building 4241/E-42 was the Preparation Control center which housed remote controls for operations in the other two buildings. Building 4243/E-44 housed a remotely controlled mandrel puller for pulling mandrels (casting cores) from cured grain, and a vertical lathe for trimming grain to shape and size. - Jet Propulsion Laboratory Edwards Facility, Edwards Air Force Base, Boron, Kern County, CA

HTR-PROTEUS pebble bed experimental program cores 9 & 10: columnar hexagonal point-on-point packing with a 1:1 moderator-to-fuel pebble ratio

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bess, John D.

2014-03-01

PROTEUS is a zero-power research reactor based on a cylindrical graphite annulus with a central cylindrical cavity. The graphite annulus remains basically the same for all experimental programs, but the contents of the central cavity are changed according to the type of reactor being investigated. Through most of its service history, PROTEUS has represented light-water reactors, but from 1992 to 1996 PROTEUS was configured as a pebble-bed reactor (PBR) critical facility and designated as HTR-PROTEUS. The nomenclature was used to indicate that this series consisted of High Temperature Reactor experiments performed in the PROTEUS assembly. During this period, seventeen criticalmore » configurations were assembled and various reactor physics experiments were conducted. These experiments included measurements of criticality, differential and integral control rod and safety rod worths, kinetics, reaction rates, water ingress effects, and small sample reactivity effects (Ref. 3). HTR-PROTEUS was constructed, and the experimental program was conducted, for the purpose of providing experimental benchmark data for assessment of reactor physics computer codes. Considerable effort was devoted to benchmark calculations as a part of the HTR-PROTEUS program. References 1 and 2 provide detailed data for use in constructing models for codes to be assessed. Reference 3 is a comprehensive summary of the HTR-PROTEUS experiments and the associated benchmark program. This document draws freely from these references. Only Cores 9 and 10 are evaluated in this benchmark report due to similarities in their construction. The other core configurations of the HTR-PROTEUS program are evaluated in their respective reports as outlined in Section 1.0. Cores 9 and 10 were evaluated and determined to be acceptable benchmark experiments.« less

HTR-PROTEUS PEBBLE BED EXPERIMENTAL PROGRAM CORES 5, 6, 7, & 8: COLUMNAR HEXAGONAL POINT-ON-POINT PACKING WITH A 1:2 MODERATOR-TO-FUEL PEBBLE RATIO

DOE Office of Scientific and Technical Information (OSTI.GOV)

John D. Bess

2013-03-01

PROTEUS is a zero-power research reactor based on a cylindrical graphite annulus with a central cylindrical cavity. The graphite annulus remains basically the same for all experimental programs, but the contents of the central cavity are changed according to the type of reactor being investigated. Through most of its service history, PROTEUS has represented light-water reactors, but from 1992 to 1996 PROTEUS was configured as a pebble-bed reactor (PBR) critical facility and designated as HTR-PROTEUS. The nomenclature was used to indicate that this series consisted of High Temperature Reactor experiments performed in the PROTEUS assembly. During this period, seventeen criticalmore » configurations were assembled and various reactor physics experiments were conducted. These experiments included measurements of criticality, differential and integral control rod and safety rod worths, kinetics, reaction rates, water ingress effects, and small sample reactivity effects (Ref. 3). HTR-PROTEUS was constructed, and the experimental program was conducted, for the purpose of providing experimental benchmark data for assessment of reactor physics computer codes. Considerable effort was devoted to benchmark calculations as a part of the HTR-PROTEUS program. References 1 and 2 provide detailed data for use in constructing models for codes to be assessed. Reference 3 is a comprehensive summary of the HTR-PROTEUS experiments and the associated benchmark program. This document draws freely from these references. Only Cores 9 and 10 are evaluated in this benchmark report due to similarities in their construction. The other core configurations of the HTR-PROTEUS program are evaluated in their respective reports as outlined in Section 1.0. Cores 9 and 10 were evaluated and determined to be acceptable benchmark experiments.« less

HTR-PROTEUS PEBBLE BED EXPERIMENTAL PROGRAM CORES 9 & 10: COLUMNAR HEXAGONAL POINT-ON-POINT PACKING WITH A 1:1 MODERATOR-TO-FUEL PEBBLE RATIO

DOE Office of Scientific and Technical Information (OSTI.GOV)

John D. Bess

2013-03-01

PROTEUS is a zero-power research reactor based on a cylindrical graphite annulus with a central cylindrical cavity. The graphite annulus remains basically the same for all experimental programs, but the contents of the central cavity are changed according to the type of reactor being investigated. Through most of its service history, PROTEUS has represented light-water reactors, but from 1992 to 1996 PROTEUS was configured as a pebble-bed reactor (PBR) critical facility and designated as HTR-PROTEUS. The nomenclature was used to indicate that this series consisted of High Temperature Reactor experiments performed in the PROTEUS assembly. During this period, seventeen criticalmore » configurations were assembled and various reactor physics experiments were conducted. These experiments included measurements of criticality, differential and integral control rod and safety rod worths, kinetics, reaction rates, water ingress effects, and small sample reactivity effects (Ref. 3). HTR-PROTEUS was constructed, and the experimental program was conducted, for the purpose of providing experimental benchmark data for assessment of reactor physics computer codes. Considerable effort was devoted to benchmark calculations as a part of the HTR-PROTEUS program. References 1 and 2 provide detailed data for use in constructing models for codes to be assessed. Reference 3 is a comprehensive summary of the HTR-PROTEUS experiments and the associated benchmark program. This document draws freely from these references. Only Cores 9 and 10 are evaluated in this benchmark report due to similarities in their construction. The other core configurations of the HTR-PROTEUS program are evaluated in their respective reports as outlined in Section 1.0. Cores 9 and 10 were evaluated and determined to be acceptable benchmark experiments.« less

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.

Experience in estimating neutron poison worths

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chiang, R.T.; Congdon, S.P.

1989-01-01

Gadolinia, {sup 135}Xe, {sup 149}Sm, control rod, and soluble boron are five neutron poisons that may appear in light water reactor assemblies. Reliable neutron poison worth estimation is useful for evaluating core operating strategies, fuel cycle economics, and reactor safety design. Based on physical presence, neutron poisons can be divided into two categories: local poisons and global poisons. Gadolinia and control rod are local poisons, and {sup 135}Xe, {sup 149}Sm, and soluble boron are global poisons. The first-order perturbation method is commonly used to estimate nuclide worths in fuel assemblies. It is well known, however, that the first-order perturbation methodmore » was developed for small perturbations, such as the perturbation due to weak absorbers, and that neutron poisons are not weak absorbers. The authors have developed an improved method to replace the first-order perturbation method, which yields very poor results, for estimating local poison worths. It has also been shown that the first-order perturbation method seems adequate to estimate worths for global poisons caused by flux compensation.« less

The improvement of the method of equivalent cross section in HTR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Guo, J.; Li, F.

The Method of Equivalence Cross-Sections (MECS) is a combined transport-diffusion method. By appropriately adjusting the diffusion coefficient of homogenized absorber region, the diffusion theory could yield satisfactory results for the full core model with strong neutron absorber material, for example the control rod in High temperature gas cooled reactor (HTR). Original implementation of MECS based on 1-D cell transport model has some limitation on accuracy and applicability, a new implementation of MECS based on 2-D transport model are proposed and tested in this paper. This improvement can extend the MECS to the calculation of twin small absorber ball system whichmore » have a non-circular boring in graphite reflector and different radial position. A least-square algorithm for the calculation of equivalent diffusion coefficient is adopted, and special treatment for diffusion coefficient for higher energy group is proposed in the case that absorber is absent. Numerical results to adopt MECS into control rod calculation in HTR are encouraging. However, there are some problems left. (authors)« less

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

ATR LEU Fuel and Burnable Absorber Neutronics Performance Optimization by Fuel Meat Thickness Variation

DOE Office of Scientific and Technical Information (OSTI.GOV)

G. S. Chang

2007-09-01

The Advanced Test Reactor (ATR) is a high power density and high neutron flux research reactor operating in the United States. Powered with highly enriched uranium (HEU), the ATR has a maximum thermal power rating of 250 MWth. Because of the large test volumes located in high flux areas, the ATR is an ideal candidate for assessing the feasibility of converting an HEU driven reactor to a low-enriched core. The present work investigates the necessary modifications and evaluates the subsequent operating effects of this conversion. A detailed plate-by-plate MCNP ATR 1/8th core model was developed and validated for a fuelmore » cycle burnup comparison analysis. Using the current HEU U 235 enrichment of 93.0 % as a baseline, an analysis can be performed to determine the low-enriched uranium (LEU) density and U-235 enrichment required in the fuel meat to yield an equivalent K-eff between the HEU core and the LEU core versus effective full power days (EFPD). The MCNP ATR 1/8th core model will be used to optimize the U-235 loading in the LEU core, such that the differences in K-eff and heat flux profile between the HEU and LEU core can be minimized. The depletion methodology MCWO was used to calculate K-eff versus EFPDs in this paper. The MCWO-calculated results for the LEU cases with foil (U-10Mo) types demonstrated adequate excess reactivity such that the K-eff versus EFPDs plot is similar to the reference ATR HEU case. Each HEU fuel element contains 19 fuel plates with a fuel meat thickness of 0.508 mm. In this work, the proposed LEU (U-10Mo) core conversion case with a nominal fuel meat thickness of 0.508 mm and the same U-235 enrichment (15.5 wt%) can be used to optimize the radial heat flux profile by varying the fuel plate thickness from 0.254 to 0.457 mm at the inner 4 fuel plates (1-4) and outer 4 fuel plates (16-19). In addition, a 0.7g of burnable absorber Boron-10 was added in the inner and outer plates to reduce the initial excess reactivity, and the inner/outer heat flux more effectively. The optimized LEU relative radial fission heat flux profile is bounded by the reference ATR HEU case. However, to demonstrate that the LEU core fuel cycle performance can meet the Updated Final Safety Analysis Report (UFSAR) safety requirements, additional studies will be necessary to evaluate and compare safety parameters such as void reactivity and Doppler coefficients, control components worth (outer shim control cylinders, safety rods and regulating rod), and shutdown margins between the HEU and LEU cores.« less

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.

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)

77 FR 46940 - Airworthiness Directives; Glasflugel Gliders

Federal Register 2010, 2011, 2012, 2013, 2014

2012-08-07

... condition as corrosion damage to the elevator control rod that could lead to failure of the elevator control... into the elevator control rod through a control bore hole and resulted in corrosion damage. The investigation concluded as well that the corrosion cannot be detected from outside the elevator control rod...

Extensions of the MCNP5 and TRIPOLI4 Monte Carlo Codes for Transient Reactor Analysis

NASA Astrophysics Data System (ADS)

Hoogenboom, J. Eduard; Sjenitzer, Bart L.

2014-06-01

To simulate reactor transients for safety analysis with the Monte Carlo method the generation and decay of delayed neutron precursors is implemented in the MCNP5 and TRIPOLI4 general purpose Monte Carlo codes. Important new variance reduction techniques like forced decay of precursors in each time interval and the branchless collision method are included to obtain reasonable statistics for the power production per time interval. For simulation of practical reactor transients also the feedback effect from the thermal-hydraulics must be included. This requires coupling of the Monte Carlo code with a thermal-hydraulics (TH) code, providing the temperature distribution in the reactor, which affects the neutron transport via the cross section data. The TH code also provides the coolant density distribution in the reactor, directly influencing the neutron transport. Different techniques for this coupling are discussed. As a demonstration a 3x3 mini fuel assembly with a moving control rod is considered for MCNP5 and a mini core existing of 3x3 PWR fuel assemblies with control rods and burnable poisons for TRIPOLI4. Results are shown for reactor transients due to control rod movement or withdrawal. The TRIPOLI4 transient calculation is started at low power and includes thermal-hydraulic feedback. The power rises about 10 decades and finally stabilises the reactor power at a much higher level than initial. The examples demonstrate that the modified Monte Carlo codes are capable of performing correct transient calculations, taking into account all geometrical and cross section detail.

Percolation in suspensions of hard nanoparticles: From spheres to needles

NASA Astrophysics Data System (ADS)

Schilling, Tanja; Miller, Mark A.; van der Schoot, Paul

2015-09-01

We investigate geometric percolation and scaling relations in suspensions of nanorods, covering the entire range of aspect ratios from spheres to extremely slender needles. A new version of connectedness percolation theory is introduced and tested against specialised Monte Carlo simulations. The theory accurately predicts percolation thresholds for aspect ratios of rod length to width as low as 10. The percolation threshold for rod-like particles of aspect ratios below 1000 deviates significantly from the inverse aspect ratio scaling prediction, thought to be valid in the limit of infinitely slender rods and often used as a rule of thumb for nanofibres in composite materials. Hence, most fibres that are currently used as fillers in composite materials cannot be regarded as practically infinitely slender for the purposes of percolation theory. Comparing percolation thresholds of hard rods and new benchmark results for ideal rods, we find that i) for large aspect ratios, they differ by a factor that is inversely proportional to the connectivity distance between the hard cores, and ii) they approach the slender rod limit differently.

Benchmark tests of JENDL-3.2 for thermal and fast reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Takano, Hideki; Akie, Hiroshi; Kikuchi, Yasuyuki

1994-12-31

Benchmark calculations for a variety of thermal and fast reactors have been performed by using the newly evaluated JENDL-3 Version-2 (JENDL-3.2) file. In the thermal reactor calculations for the uranium and plutonium fueled cores of TRX and TCA, the k{sub eff} and lattice parameters were well predicted. The fast reactor calculations for ZPPR-9 and FCA assemblies showed that the k{sub eff} reactivity worths of Doppler, sodium void and control rod, and reaction rate distribution were in a very good agreement with the experiments.

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.

Dispersion-compensating photonic crystal fiber with wavelength tunability based on a modified dual concentric core structure

NASA Astrophysics Data System (ADS)

Chen, Nan; Zhang, Xuedian; Nie, Fukun; Lu, Xinglian; Chang, Min

2018-07-01

We present a 5-layer air-hole dispersion-compensating photonic crystal fiber (PCF) with a modified dual concentric core structure, based on central rod doping. The finite element method (FEM) was used to investigate the structure numerically. If the structural parameters remain unchanged, a high degree of linear correlation between the central rod refractive index and the operating wavelength can be achieved in the wavelength range of 1.5457-1.5857 μm, which suggests that the operating wavelength can be determined by the refractive index of the centre rod. A negative dispersion coefficient between -5765.2 ps/km/nm and -6115.8 ps/km/nm was obtained by calculation and within the bandwidth of 108 nm (1.515-1.623 μm) around 1.55 μm, a dispersion coefficient of -3000 ps/km/nm can be ensured for compensation. In addition, this proposed PCF also has the advantage of low confinement loss, between 0.00011 and 0.00012 dB/m, and ease of fabrication with existing technology. The proposed PCF has good prospects in dispersion-compensating applications.

DOE Office of Scientific and Technical Information (OSTI.GOV)

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

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

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.

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)

JPRS Report, Science & Technology, China: Energy

DTIC Science & Technology

1988-06-29

capacity. There are currently two types of HTGR reactor designs: the particle-bed core , which uses spherical fuel elements, and the rod type core , in...and trial operating experience with the HTGR reactor. Its main design features are as follows. 1. A particle-bed core , continuous fueling and...Favorable for Development of Small-Scale HTGR (Xu Jiming; HE DONGLI GONGCHENG, Feb 88) 47 ERRATUM: In JPRS-CEN-88-003 of 25 April 1988 in article

TRAC-PD2 posttest analysis of CCTF Test C1-16 (Run 025). [Cylindrical Core Test Facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sugimoto, J.

The TRAC-PD2 code version was used to analyze CCTF Test C1-16 (Run 025). The results indicate that the core heater rod temperatures, the liquid mass in the vessel, and differential pressures in the primary loop are predicted well, but the void fraction distribution in the core and water accumulation in the upper plenum are not in good agreement with the data.

Flexible growing rods: a pilot study to determine if polymer rod constructs may provide stability to skeletally immature spines

PubMed Central

2015-01-01

Background Surgical treatments for early onset scoliosis (EOS), including growing rod constructs, involve many complications. Some are due to biomechanical factors. A construct that is more flexible than current instrumentation systems may reduce complications. The purpose of this preliminary study was to determine spine range of motion (ROM) after implantation of simulated growing rod constructs with a range of clinically relevant structural properties. The hypothesis was that ROM of spines instrumented with polyetheretherketone (PEEK) rods would be greater than metal rods and lower than noninstrumented controls. Further, adjacent segment motion was expected to be lower with polymer rods compared to conventional systems. Methods Biomechanical tests were conducted on 6 skeletally immature porcine thoracic spines (domestic swine, 35-40 kg). Spines were harvested after death from swine that had been utilized for other studies (IACUC approved) which had not involved the spine. Paired pedicle screws were used as anchors at proximal and distal levels. Specimens were tested under the following conditions: control, then dual rods of PEEK (6.25 mm), titanium (4 mm), and CoCr (5 mm) alloy. Lateral bending (LB) and flexion-extension (FE) moments of ±5 Nm were applied. Vertebral rotations were measured using video. Differences were determined by two-tailed t-tests and Bonferroni correction with four primary comparisons: PEEK vs control and PEEK vs CoCr, in LB and FE (α=0.05/4). Results In LB, ROM of specimens with PEEK rods was lower than control at each instrumented level. ROM was greater for PEEK rods than both Ti and CoCr at every instrumented level. Mean ROM at proximal and distal noninstrumented levels was lower for PEEK than for Ti and CoCr. In FE, mean ROM at proximal and distal noninstrumented levels was lower for PEEK than for metal. Combining treated levels, in LB, ROM for PEEK rods was 35% of control (p<0.0001) and 270% of CoCr rods (p<0.01). In FE, ROM with PEEK was 27% of control (p<0.001) and 180% of CoCr (p<0.01). Conclusions PEEK rods decreased flexibility versus noninstumented controls, and increased flexibility versus metal rods. Smaller increases in ROM at proximal and distal adjacent motion segments occurred with PEEK compared to metal rods, which may help decrease junctional kyphosis. Flexible growing rods may eventually help improve treatment options for young patients with severe deformity. PMID:25810752

A wireline piston core barrel for sampling cohesionless sand and gravel below the water table

USGS Publications Warehouse

Zapico, Michael M.; Vales, Samuel; Cherry, John A.

1987-01-01

A coring device has been developed to obtain long and minimally disturbed samples of saturated cohesionless sand and gravel. The coring device, which includes a wireline and piston, was developed specifically for use during hollow-stem auger drilling but it also offers possibilities for cable tool and rotary drilling. The core barrel consists of an inner liner made of inexpensive aluminum or plastic tubing, a piston for core recovery, and an exterior steel housing that protects the liner when the core barrel is driven into the aquifer. The core barrel, which is approximately 1.6m (5.6 feet) long, is advanced ahead of the lead auger by hammering at the surface on drill rods that are attached to the core barrel. After the sampler has been driven 1.5m (5 feet), the drill rods are detached and a wireline is used to hoist the core barrel, with the sample contained in the aluminum or plastic liner, to the surface. A vacuum developed by the piston during the coring operation provides good recovery of both the sediment and aquifer fluids contained in the sediment. In the field the sample tubes can be easily split along their length for on-site inspection or they can be capped with the pore water fluids inside and transported to the laboratory. The cores are 5cm (2 inches) in diameter by 1.5m (5 feet) long. Core acquisition to depths of 35m (115 feet), with a recovery greater than 90 percent, has become routine in University of Waterloo aquifer studies. A large diameter (12.7cm [5 inch]) version has also been used successfully. Nearly continuous sample sequences from sand and gravel aquifers have been obtained for studies of sedimentology, hydraulic conductivity, hydrogeochemistry and microbiology.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bess, John D.; Sterbentz, James W.; Snoj, Luka

PROTEUS is a zero-power research reactor based on a cylindrical graphite annulus with a central cylindrical cavity. The graphite annulus remains basically the same for all experimental programs, but the contents of the central cavity are changed according to the type of reactor being investigated. Through most of its service history, PROTEUS has represented light-water reactors, but from 1992 to 1996 PROTEUS was configured as a pebble-bed reactor (PBR) critical facility and designated as HTR-PROTEUS. The nomenclature was used to indicate that this series consisted of High Temperature Reactor experiments performed in the PROTEUS assembly. During this period, seventeen criticalmore » configurations were assembled and various reactor physics experiments were conducted. These experiments included measurements of criticality, differential and integral control rod and safety rod worths, kinetics, reaction rates, water ingress effects, and small sample reactivity effects (Ref. 3). HTR-PROTEUS was constructed, and the experimental program was conducted, for the purpose of providing experimental benchmark data for assessment of reactor physics computer codes. Considerable effort was devoted to benchmark calculations as a part of the HTR-PROTEUS program. References 1 and 2 provide detailed data for use in constructing models for codes to be assessed. Reference 3 is a comprehensive summary of the HTR-PROTEUS experiments and the associated benchmark program. This document draws freely from these references. Only Cores 9 and 10 are evaluated in this benchmark report due to similarities in their construction. The other core configurations of the HTR-PROTEUS program are evaluated in their respective reports as outlined in Section 1.0. Cores 9 and 10 were evaluated and determined to be acceptable benchmark experiments.« less

Fail-safe reactivity compensation method for a nuclear reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nygaard, Erik T.; Angelo, Peter L.; Aase, Scott B.

The present invention relates generally to the field of compensation methods for nuclear reactors and, in particular to a method for fail-safe reactivity compensation in solution-type nuclear reactors. In one embodiment, the fail-safe reactivity compensation method of the present invention augments other control methods for a nuclear reactor. In still another embodiment, the fail-safe reactivity compensation method of the present invention permits one to control a nuclear reaction in a nuclear reactor through a method that does not rely on moving components into or out of a reactor core, nor does the method of the present invention rely on themore » constant repositioning of control rods within a nuclear reactor in order to maintain a critical state.« less

Electrical servo actuator bracket. [fuel control valves on jet engines

NASA Technical Reports Server (NTRS)

Sawyer, R. V. (Inventor)

1981-01-01

An electrical servo actuator is mounted on a support arm which is allowed to pivot on a bolt through a fixed mounting bracket. The actuator is pivotally connected to the end of the support arm by a bolt which has an extension allowed to pass through a slot in the fixed mounting bracket. An actuator rod extends from the servo actuator to a crank arm which turns a control shaft. A short linear thrust of the rod pivots the crank arm through about 90 for full-on control with the rod contracted into the servo actuator, and full-off control when the rod is extended from the actuator. A spring moves the servo actuator and actuator rod toward the control crank arm once the actuator rod is fully extended in the full-off position. This assures the turning of the control shaft to a full-off position. A stop bolt and slot are provided to limit pivot motion. Once fully extended, the spring pivots the motion.

Sheath-Core Graphite/Silk Fiber Made by Dry-Meyer-Rod-Coating for Wearable Strain Sensors.

PubMed

Zhang, Mingchao; Wang, Chunya; Wang, Qi; Jian, Muqiang; Zhang, Yingying

2016-08-17

Recent years have witnessed the explosive development of flexible strain sensors. Nanomaterials have been widely utilized to fabricate flexible strain sensors, because of their high flexibility and electrical conductivity. However, the fabrication processes for nanomaterials and the subsequent strain sensors are generally complicated and are manufactured at high cost. In this work, we developed a facile dry-Meyer-rod-coating process to fabricate sheath-core-structured single-fiber strain sensors using ultrafine graphite flakes as the sheath and silk fibers as the core by virtue of their flexibility, high production, and low cost. The fabricated strain sensor exhibits a high sensitivity with a gauge factor of 14.5 within wide workable strain range up to 15%, and outstanding stability (up to 3000 cycles). The single-fiber-based strain sensors could be attached to a human body to detect joint motions or easily integrated into the multidirectional strain sensor for monitoring multiaxial strain, showing great potential applications as wearable strain sensors.

Generation of 180 W average green power from a frequency-doubled picosecond rod fiber amplifier

DOE PAGES

Zhao, Zhi; Sheehy, Brian; Minty, Michiko

2017-03-29

Here, we report on the generation of 180 W average green power from a frequency-doubled picosecond rod fiber amplifier. In an Yb-doped fiber master-oscillator-power-amplifier system, 2.3-ps 704 MHz pulses are first amplified in small-core fibers and then in large-mode-area rod fibers to produce 270 W average infrared power with a high polarization extinction ratio and diffraction-limited beam quality. By carrying out frequency doubling in a lithium triborate (LBO) crystal, 180 W average green power is generated. To the best of our knowledge, this is the highest average green power achieved in fiber-based laser systems.

Interim status report on lead-cooled fast reactor (LFR) research and development.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tzanos, C. P.; Sienicki, J. J.; Moisseytsev, A.

2008-03-31

This report discusses the status of Lead-Cooled Fast Reactor (LFR) research and development carried out during the first half of FY 2008 under the U.S. Department of Energy Generation IV Nuclear Energy Systems Initiative. Lead-Cooled Fast Reactor research and development has recently been transferred from Generation IV to the Reactor Campaign of the Global Nuclear Energy Partnership (GNEP). Another status report shall be issued at the end of FY 2008 covering all of the LFR activities carried out in FY 2008 for both Generation IV and GNEP. The focus of research and development in FY 2008 is an initial investigationmore » of a concept for a LFR Advanced Recycling Reactor (ARR) Technology Pilot Plant (TPP)/demonstration test reactor (demo) incorporating features and operating conditions of the European Lead-cooled SYstem (ELSY) {approx} 600 MWe lead (Pb)-cooled LFR preconceptual design for the transmutation of waste and central station power generation, and which would enable irradiation testing of advanced fuels and structural materials. Initial scoping core concept development analyses have been carried out for a 100 MWt core composed of sixteen open-lattice 20 by 20 fuel assemblies largely similar to those of the ELSY preconceptual fuel assembly design incorporating fuel pins with mixed oxide (MOX) fuel, central control rods in each fuel assembly, and cooled with Pb coolant. For a cycle length of three years, the core is calculated to have a conversion ratio of 0.79, an average discharge burnup of 108 MWd/kg of heavy metal, and a burnup reactivity swing of about 13 dollars. With a control rod in each fuel assembly, the reactivity worth of an individual rod would need to be significantly greater than one dollar which is undesirable for postulated rod withdrawal reactivity insertion events. A peak neutron fast flux of 2.0 x 10{sup 15} (n/cm{sup 2}-s) is calculated. For comparison, the 400 MWt Fast Flux Test Facility (FFTF) achieved a peak neutron fast flux of 7.2 x 10{sup 15} (n/cm{sup 2}-s) and the initially 563 MWt PHENIX reactor attained 2.0 x 10{sup 15} (n/cm{sup 2}-s) before one of three intermediate cooling loops was shut down due to concerns about potential steam generator tube failures. The calculations do not assume a test assembly location for advanced fuels and materials irradiation in place of a fuel assembly (e.g., at the center of the core); the calculations have not examined whether it would be feasible to replace the central assembly by a test assembly location. However, having only fifteen driver assemblies implies a significant effect due to perturbations introduced by the test assembly. The peak neutron fast flux is low compared with the fast fluxes previously achieved in FFTF and PHENIX. Furthermore, the peak neutron fluence is only about half of the limiting value (4 x 10{sup 23} n/cm{sup 2}) typically used for ferritic steels. The results thus suggest that a larger power level (e.g., 400 MWt) and a larger core would be better for a TPP based upon the ELSY fuel assembly design and which can also perform irradiation testing of advanced fuels and materials. In particular, a core having a higher power level and larger dimensions would achieve a suitable average discharge burnup, peak fast flux, peak fluence, and would support the inclusion of one or more test assembly locations. Participation in the Generation IV International Forum Provisional System Steering Committee for the LFR is being maintained throughout FY 2008. Results from the analysis of samples previously exposed to flowing lead-bismuth eutectic (LBE) in the DELTA loop are summarized and a model for the oxidation/corrosion kinetics of steels in heavy liquid metal coolants was applied to systematically compare the calculated long-term (i.e., following several years of growth) oxide layer thicknesses of several steels.« less

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

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

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

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

CONTROL IN NUCLEAR REACTORS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bell, F.R.

1963-01-16

An arrangement was described for scramming a reactor in an emergency. Control rods were position adjusted by an electric motor and transmission. A locking system kept the control rods in position but was arranged to be released in an emergency to allow the rods to drop into their shutdown position. (C.E.S.)

HWCTR CONTROL ROD AND SAFETY ROD DRIVE SYSTEMS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kale, S.H.

1963-07-01

The Heavy Water Components Test Reactor (HWCTR) is a pressurized, D/sub 2/O reactor designed for operation up to 70 Mw at 1500 psig and 3l5 deg C. It has 18 control rods and six safety rods, each driven by an electric motor through a rack and pinion gear train. Racks, pinions, and bearings are located inside individual pressure housings that are penetrated by means of floating ring labyrinth seals. The drives are mounted on the reactor vessel top head. Safety rods have electromagnetic clutches and fall into the reactor when scrammed. The reliability and performance of the rod drives aremore » very good. Seal leakage is well within design limits. Recent inspections of seals and control rod plants showed no evidence of crud buildup or stress corrosion cracking of type 17- 4PH'' stainless steel components. (auth)« less

Assessment of the Neutronic and Fuel Cycle Performance of the Transatomic Power Molten Salt Reactor Design

DOE Office of Scientific and Technical Information (OSTI.GOV)

Robertson, Sean; Dewan, Leslie; Massie, Mark

This report presents results from a collaboration between Transatomic Power Corporation (TAP) and Oak Ridge National Laboratory (ORNL) to provide neutronic and fuel cycle analysis of the TAP core design through the Department of Energy Gateway for Accelerated Innovation in Nuclear (GAIN) Nuclear Energy Voucher program. The TAP concept is a molten salt reactor using configurable zirconium hydride moderator rod assemblies to shift the neutron spectrum in the core from mostly epithermal at beginning of life to thermal at end of life. Additional developments in the ChemTriton modeling and simulation tool provide the critical moderator-to-fuel ratio searches and time-dependent parametersmore » necessary to simulate the continuously changing physics in this complex system. The implementation of continuous-energy Monte Carlo transport and depletion tools in ChemTriton provide for full-core three-dimensional modeling and simulation. Results from simulations with these tools show agreement with TAP-calculated performance metrics for core lifetime, discharge burnup, and salt volume fraction, verifying the viability of reducing actinide waste production with this concept. Additional analyses of mass feed rates and enrichments, isotopic removals, tritium generation, core power distribution, core vessel helium generation, moderator rod heat deposition, and reactivity coeffcients provide additional information to make informed design decisions. This work demonstrates capabilities of ORNL modeling and simulation tools for neutronic and fuel cycle analysis of molten salt reactor concepts.« less

The analysis and compensation of errors of precise simple harmonic motion control under high speed and large load conditions based on servo electric cylinder

NASA Astrophysics Data System (ADS)

Ma, Chen-xi; Ding, Guo-qing

2017-10-01

Simple harmonic waves and synthesized simple harmonic waves are widely used in the test of instruments. However, because of the errors caused by clearance of gear and time-delay error of FPGA, it is difficult to control servo electric cylinder in precise simple harmonic motion under high speed, high frequency and large load conditions. To solve the problem, a method of error compensation is proposed in this paper. In the method, a displacement sensor is fitted on the piston rod of the electric cylinder. By using the displacement sensor, the real-time displacement of the piston rod is obtained and fed back to the input of servo motor, then a closed loop control is realized. There is compensation of pulses in the next period of the synthetic waves. This paper uses FPGA as the processing core. The software mainly comprises a waveform generator, an Ethernet module, a memory module, a pulse generator, a pulse selector, a protection module, an error compensation module. A durability of shock absorbers is used as the testing platform. The durability mainly comprises a single electric cylinder, a servo motor for driving the electric cylinder, and the servo motor driver.

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

NASA Astrophysics Data System (ADS)

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

2015-09-01

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

Ultrasound control of magnet growing rod distraction in early onset scoliosis.

PubMed

Pérez Cervera, T; Lirola Criado, J F; Farrington Rueda, D M

2016-01-01

The growing rod technique is currently one of the most common procedures used in the management of early onset scoliosis. However, in order to preserve spine growth and control the deformity it requires frequent surgeries to distract the rods. Magnetically driven growing rods have recently been introduced with same treatment goal, but without the inconvenience of repeated surgical distractions. One of the limitations of this technical advance is an increase in radiation exposure due to the increase in distraction frequency compared to conventional growing rods. An improvement of the original technique is presented, proposing a solution to the inconvenience of multiple radiation exposure using ultrasound technology to control the distraction process of magnetically driven growing rods. Copyright © 2014 SECOT. Publicado por Elsevier España, S.L.U. All rights reserved.

Reactivity Initiated Accident Simulation to Inform Transient Testing of Candidate Advanced Cladding

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brown, Nicholas R; Wysocki, Aaron J; Terrani, Kurt A

2016-01-01

Abstract. Advanced cladding materials with potentially enhanced accident tolerance will yield different light water reactor performance and safety characteristics than the present zirconium-based cladding alloys. These differences are due to different cladding material properties and responses to the transient, and to some extent, reactor physics, thermal, and hydraulic characteristics. Some of the differences in reactors physics characteristics will be driven by the fundamental properties (e.g., absorption in iron for an iron-based cladding) and others will be driven by design modifications necessitated by the candidate cladding materials (e.g., a larger fuel pellet to compensate for parasitic absorption). Potential changes in thermalmore » hydraulic limits after transition from the current zirconium-based cladding to the advanced materials will also affect the transient response of the integral fuel. This paper leverages three-dimensional reactor core simulation capabilities to inform on appropriate experimental test conditions for candidate advanced cladding materials in a control rod ejection event. These test conditions are using three-dimensional nodal kinetics simulations of a reactivity initiated accident (RIA) in a representative state-of-the-art pressurized water reactor with both nuclear-grade iron-chromium-aluminum (FeCrAl) and silicon carbide based (SiC-SiC) cladding materials. The effort yields boundary conditions for experimental mechanical tests, specifically peak cladding strain during the power pulse following the rod ejection. The impact of candidate cladding materials on the reactor kinetics behavior of RIA progression versus reference zirconium cladding is predominantly due to differences in: (1) fuel mass/volume/specific power density, (2) spectral effects due to parasitic neutron absorption, (3) control rod worth due to hardened (or softened) spectrum, and (4) initial conditions due to power peaking and neutron transport cross sections in the equilibrium cycle cores due to hardened (or softened) spectrum. This study shows minimal impact of SiC-based cladding configurations on the transient response versus reference zirconium-based cladding. However, the FeCrAl cladding response indicates similar energy deposition, but with significantly shorter pulses of higher magnitude. Therefore the FeCrAl-based cases have a more rapid fuel thermal expansion rate and the resultant pellet-cladding interaction occurs more rapidly.« less

ATR LEU fuel and burnable absorber neutronics performance optimization by fuel meat thickness variation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chang, G.S.

2008-07-15

The Advanced Test Reactor (ATR) is a high power density and high neutron flux research reactor operating in the United States. Powered with highly enriched uranium (HEU), the ATR has a maximum thermal power rating of 250 MWth. Because of the large test volumes located in high flux areas, the ATR is an ideal candidate for assessing the feasibility of converting an HEU driven reactor to a low-enriched core. The present work investigates the necessary modifications and evaluates the subsequent operating effects of this conversion. A detailed plate-by-plate MCNP ATR 1/8th core model was developed and validated for a fuelmore » cycle burnup comparison analysis. Using the current HEU U-235 enrichment of 93.0 % as a baseline, an analysis can be performed to determine the low-enriched uranium (LEU) density and U-235 enrichment required in the fuel meat to yield an equivalent K-eff between the HEU core th and the LEU core versus effective full power days (EFPD). The MCNP ATR 1/8th core model will be used to optimize the U-235 loading in the LEU core, such that the differences in K-eff and heat flux profile between the HEU and LEU core can be minimized. The depletion methodology MCWO was used to calculate K-eff versus EFPDs in this paper. The MCWO-calculated results for the LEU cases with foil (U-10Mo) types demonstrated adequate excess reactivity such that the K-eff versus EFPDs plot is similar to the reference ATR HEU case. Each HEU fuel element contains 19 fuel plates with a fuel meat thickness of 0.508 mm. In this work, the proposed LEU (U-10Mo) core conversion case with a nominal fuel meat thickness of 0.381 mm and the same U-235 enrichment (19.7 wt%) can be used to optimize the radial heat flux profile by varying the fuel meat thickness from 0.191 mm (7.5 mil) to 0.343 mm (13.5 mil) at the inner 4 fuel plates (1-4) and outer 4 fuel plates (16-19). In addition, 0.8g of a burnable absorber, Boron-10, was added in the inner and outer plates to reduce the initial excess reactivity, and the inner/outer heat flux more effectively. The optimized LEU relative radial fission heat flux profile is bounded by the reference ATR HEU case. However, to demonstrate that the LEU core fuel cycle performance can meet the Updated Final Safety Analysis Report (UFSAR) safety requirements, additional studies will be necessary to evaluate and compare safety parameters such as void reactivity and Doppler coefficients, control components worth (outer shim control cylinders, safety rods and regulating rod), and shutdown margins between the HEU and LEU cores. (author)« less

Few-Mode Whispering-Gallery-Mode Resonators

NASA Technical Reports Server (NTRS)

Savchenkov, Anatoliy; Strekalov, Dmitry; Matsko, Andrey; Iltchenko, Vladimir; Maleki, Lute

2006-01-01

Whispering-gallery-mode (WGM) optical resonators of a type now under development are designed to support few well-defined waveguide modes. In the simplest case, a resonator of this type would support one equatorial family of WGMs; in a more complex case, such a resonator would be made to support two, three, or some other specified finite number of modes. Such a resonator can be made of almost any transparent material commonly used in optics. The nature of the supported modes does not depend on which material is used, and the geometrical dispersion of this resonator is much smaller than that of a typical prior WGM resonator. Moreover, in principle, many such resonators could be fabricated as integral parts of a single chip. Basically, a resonator of this type consists of a rod, made of a suitable transparent material, from which protrudes a thin circumferential belt of the same material. The belt is integral with the rest of the rod (see figure) and acts as a circumferential waveguide. If the depth (d) and width (w) of the belt are made appropriately small, then the belt acts as though it were the core of a single-mode optical fiber: the belt and its adjacent supporting rod material support a single, circumferentially propagating mode or family of modes. It has been shown theoretically that the fiber-optic-like behavior of the belton- rod resonator structure can be summarized, in part, by the difference, Dn, between (1) an effective index of refraction of an imaginary fiber core and (2) the index of refraction (n) of the transparent rod/belt material. It has also been shown theoretically that for a given required value of Dn, the required depth of the belt can be estimated as d R Dn, where R is the radius of the rod. It must be emphasized that this estimated depth is independent of n and, hence, is independent of the choice of rod material. As in the cases of prior WGM resonators, input/output optical coupling involves utilization of evanescent fields. In the present case, there are two evanescent fields: one at the belt/air interface and one in the boundary region between the belt and the rest of the rod.

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.

Performance study of highly efficient 520 W average power long pulse ceramic Nd:YAG rod laser

NASA Astrophysics Data System (ADS)

Choubey, Ambar; Vishwakarma, S. C.; Ali, Sabir; Jain, R. K.; Upadhyaya, B. N.; Oak, S. M.

2013-10-01

We report the performance study of a 2% atomic doped ceramic Nd:YAG rod for long pulse laser operation in the millisecond regime with pulse duration in the range of 0.5-20 ms. A maximum average output power of 520 W with 180 J maximum pulse energy has been achieved with a slope efficiency of 5.4% using a dual rod configuration, which is the highest for typical lamp pumped ceramic Nd:YAG lasers. The laser output characteristics of the ceramic Nd:YAG rod were revealed to be nearly equivalent or superior to those of high-quality single crystal Nd:YAG rod. The laser pump chamber and resonator were designed and optimized to achieve a high efficiency and good beam quality with a beam parameter product of 16 mm mrad (M2˜47). The laser output beam was efficiently coupled through a 400 μm core diameter optical fiber with 90% overall transmission efficiency. This ceramic Nd:YAG laser will be useful for various material processing applications in industry.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Garnier, Ch.; Mailhe, P.; Sontheimer, F.

2007-07-01

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

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.

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.

Optimal Navigation of Self-Propelled Colloids in Microstructured Mazes

NASA Astrophysics Data System (ADS)

Yang, Yuguang; Bevan, Michael

Controlling navigation of self-propelled microscopic `robots' subject to random Brownian motion in complex microstructured environments (e.g., porous media, tumor vasculature) is important to many emerging applications (e.g., enhanced oil recovery, drug delivery). In this work, we design an optimal feedback policy to navigate an active self-propelled colloidal rod in complex mazes with various obstacle types. Actuation of the rods is modelled based on a light-controlled osmotic flow mechanism, which produces different propulsion velocities along the rod's long axis. Actuator-parameterized Langevin equations, with soft rod-obstacle repulsive interactions, are developed to describe the system dynamics. A Markov decision process (MDP) framework is used for optimal policy calculations with design goals of colloidal rods reaching target end points in minimum time. Simulations show that optimal MDP-based policies are able to control rod trajectories to reach target regions order-of-magnitudes faster than uncontrolled rods, which diverges as maze complexity increases. An efficient multi-graph based implementation for MDP is also presented, which scales linearly with the maze dimension.

Position-controlled MOVPE growth and electro-optical characterization of core-shell InGaN/GaN microrod LEDs

NASA Astrophysics Data System (ADS)

Schimpke, Tilman; Lugauer, H.-J.; Avramescu, A.; Varghese, T.; Koller, A.; Hartmann, J.; Ledig, J.; Waag, A.; Strassburg, M.

2016-03-01

Today's InGaN-based white LEDs still suffer from a significant efficiency reduction at elevated current densities, the so-called "Droop". Core-shell microrods, with quantum wells (QWs) covering their entire surface, enable a tremendous increase in active area scaling with the rod's aspect ratio. Enlarging the active area on a given footprint area is a viable and cost effective route to mitigate the droop by effectively reducing the local current density. Microrods were grown in a large volume metal-organic vapor phase epitaxy (MOVPE) reactor on GaN-on-sapphire substrates with a thin, patterned SiO2 mask for position control. Out of the mask openings, pencil-shaped n-doped GaN microrod cores were grown under conditions favoring 3D growth. In a second growth step, these cores are covered with a shell containing a quantum well and a p-n junction to form LED structures. The emission from the QWs on the different facets was studied using resonant temperature-dependent photoluminescence (PL) and cathodoluminescence (CL) measurements. The crystal quality of the structures was investigated by transmission electron microscopy (TEM) showing the absence of extended defects like threading dislocations in the 3D core. In order to fabricate LED chips, dedicated processes were developed to accommodate for the special requirements of the 3D geometry. The electrical and optical properties of ensembles of tens of thousands microrods connected in parallel are discussed.

Final report, PT IP-535-C: Test of smaller VSR`s in DR reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vaughn, A.D.

1963-04-17

Because of rod-sticking problems at DR Reactor, a knuckle rod of B Reactor design was installed in vertical safety channel number 28. The substitute VSR, which has a smaller diameter than the original DR rod, has been tested for its operational feasibility including both drop time and reactivity effect. The reactivity effect of the rod was estimated by comparison of the reactivity transient caused by insertion of the specific B-type rod after scramming into the pile, with similar transients caused by normal vertical safety rod in an adjacent channel. This document lists the indicated relative control strength of the rodmore » as an empirical basis for future safety calculations. Results indicate that the B-type knuckel rod in DR reactor is about 80% as strong as a normal DR vertical safety rod if used in equivalent flux distribution and location; this makes it reasonable to assume that the local control strength of the B-type knuckel rod is 98 {mu}b.« less

Advanced gray rod control assembly

DOE Office of Scientific and Technical Information (OSTI.GOV)

Drudy, Keith J; Carlson, William R; Conner, Michael E

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 tomore » enhance neutron absorption, absorber depletion, assembly weight, and assembly heat transfer characteristics.« less

Analysis of fuel options for the breakeven core configuration of the Advanced Recycling Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stauff, N.E.; Klim, T.K.; Taiwo, T.A.

2013-07-01

A trade-off study is performed to determine the impacts of various fuel forms on the core design and core physics characteristics of the sodium-cooled Toshiba- Westinghouse Advanced Recycling Reactor (ARR). The fuel forms include oxide, nitride, and metallic forms of U and Th. The ARR core configuration is redesigned with driver and blanket regions in order to achieve breakeven fissile breeding performance with the various fuel types. State-of-the-art core physics tools are used for the analyses. In addition, a quasi-static reactivity balance approach is used for a preliminary comparison of the inherent safety performances of the various fuel options. Thorium-fueledmore » cores exhibit lower breeding ratios and require larger blankets compared to the U-fueled cores, which is detrimental to core compactness and increases reprocessing and manufacturing requirements. The Th cores also exhibit higher reactivity swings through each cycle, which penalizes reactivity control and increases the number of control rods required. On the other hand, using Th leads to drastic reductions in void and coolant expansion coefficients of reactivity, with the potential for enhancing inherent core safety. Among the U-fueled ARR cores, metallic and nitride fuels result in higher breeding ratios due to their higher heavy metal densities. On the other hand, oxide fuels provide a softer spectrum, which increases the Doppler effect and reduces the positive sodium void worth. A lower fuel temperature is obtained with the metallic and nitride fuels due to their higher thermal conductivities and compatibility with sodium bonds. This is especially beneficial from an inherent safety point of view since it facilitates the reactor cool-down during loss of power removal transients. The advantages in terms of inherent safety of nitride and metallic fuels are maintained when using Th fuel. However, there is a lower relative increase in heavy metal density and in breeding ratio going from oxide to metallic or nitride Th fuels relative to the U counterpart fuels. (authors)« less

Method for producing solid or hollow spherical particles of chosen chemical composition and of uniform size

DOEpatents

Hendricks, Charles D.

1988-01-01

A method is provided for producing commercially large quantities of high melting temperature solid or hollow spherical particles of a predetermined chemical composition and having a uniform and controlled size distribution. An end (18, 50, 90) of a solid or hollow rod (20, 48, 88) of the material is rendered molten by a laser beam (14, 44, 82). Because of this, there is no possibility of the molten rod material becoming contaminated with extraneous material. In various aspects of the invention, an electric field is applied to the molten rod end (18, 90), and/or the molten rod end (50, 90) is vibrated. In a further aspect of the invention, a high-frequency component is added to the electric field applied to the molten end of the rod (90). By controlling the internal pressure of the rod, the rate at which the rod is introduced into the laser beam, the environment of the process, the vibration amplitude and frequency of the molten rod end, the electric field intensity applied to the molten rod end, and the frequency and intensity of the component added to the electric field, the uniformity and size distribution of the solid or hollow spherical particles (122) produced by the inventive method is controlled. The polarity of the electric field applied to the molten rod end can be chosen to eliminate backstreaming electrons, which tend to produce run-away heating in the rod, from the process.

Measurement of Diffusion in Entangled Rod-Coil Triblock Copolymers

NASA Astrophysics Data System (ADS)

Olsen, B. D.; Wang, M.

2012-02-01

Although rod-coil block copolymers have attracted increasing attention for functional nanomaterials, their dynamics relevant to self-assembly and processing have not been widely investigated. Because the rod and coil blocks have different reptation behavior and persistence lengths, the mechanism by which block copolymers will diffuse is unclear. In order to understand the effect of the rigid block on reptation, tracer diffusion of a coil-rod-coil block copolymer through an entangled coil polymer matrix was experimentally measured. A monodisperse, high molecular weight coil-rod-coil triblock was synthesized using artificial protein engineering to prepare the helical rod and bioconjugaiton of poly(ethylene glycol) coils to produce the final triblock. Diffusion measurements were performed using Forced Rayleigh scattering (FRS), at varying ratios of the rod length to entanglement length, where genetic engineering is used to control the protein rod length and the polymer matrix concentration controls the entanglement length. As compared to PEO homopolymer tracers, the coil-rod-coil triblocks show markedly slower diffusion, suggesting that the mismatch between rod and coil reptation mechanisms results in hindered diffusion of these molecules in the entangled state.

Design of a fuel element for a lead-cooled fast reactor

NASA Astrophysics Data System (ADS)

Sobolev, V.; Malambu, E.; Abderrahim, H. Aït

2009-03-01

The options of a lead-cooled fast reactor (LFR) of the fourth generation (GEN-IV) reactor with the electric power of 600 MW are investigated in the ELSY Project. The fuel selection, design and optimization are important steps of the project. Three types of fuel are considered as candidates: highly enriched Pu-U mixed oxide (MOX) fuel for the first core, the MOX containing between 2.5% and 5.0% of the minor actinides (MA) for next core and Pu-U-MA nitride fuel as an advanced option. Reference fuel rods with claddings made of T91 ferrite-martensitic steel and two alternative fuel assembly designs (one uses a closed hexagonal wrapper and the other is an open square variant without wrapper) have been assessed. This study focuses on the core variant with the closed hexagonal fuel assemblies. Based on the neutronic parameters provided by Monte-Carlo modeling with MCNP5 and ALEPH codes, simulations have been carried out to assess the long-term thermal-mechanical behaviour of the hottest fuel rods. A modified version of the fuel performance code FEMAXI-SCK-1, adapted for fast neutron spectrum, new fuels, cladding materials and coolant, was utilized for these calculations. The obtained results show that the fuel rods can withstand more than four effective full power years under the normal operation conditions without pellet-cladding mechanical interaction (PCMI). In a variant with solid fuel pellets, a mild PCMI can appear during the fifth year, however, it remains at an acceptable level up to the end of operation when the peak fuel pellet burnup ∼80 MW d kg-1 of heavy metal (HM) and the maximum clad damage of about 82 displacements per atom (dpa) are reached. Annular pellets permit to delay PCMI for about 1 year. Based on the results of this simulation, further steps are envisioned for the optimization of the fuel rod design, aiming at achieving the fuel burnup of 100 MW d kg-1 of HM.

POSITIONING DEVICE

DOEpatents

Wall, R.R.; Peterson, D.L.

1959-09-15

A positioner is described for a vertical reactor-control rod. The positioner comprises four grooved friction rotatable members that engage the control rod on all sides and shift it longitudinally. The four friction members are drivingly interconnected for conjoint rotation and comprise two pairs of coaxial members. The members of each pair are urged toward one another by hydraulic or pneumatic pressure and thus grip the control rod so as to hold it in any position or adjust it. Release of the by-draulic or pneumatic pressure permits springs between the friction members of each pair to force them apart, whereby the control rod moves quickly by gravity into the reactor.

QUICK RELEASABLE DRIVE

DOEpatents

Dickson, J.J.

1958-07-01

A quick releasable mechanical drive system suitable for use in a nuclear reactor is described. A small reversible motor positions a control rod by means of a worm and gear speed reducer, a magnetic torque clutch, and a bell crank. As the control rod is raised to the operating position, a heavy coil spring is compressed. In the event of an emergency indicated by either a''scram'' signal or a power failure, the current to the magnetic clutch is cut off, thereby freeing the coil spring and the bell crank positioner from the motor and speed reduction gearing. The coil spring will immediately act upon the bell crank to cause the insertion of the control rod. This arrangement will allow the slow, accurate positioning of the control rod during reactor operation, while providing an independent force to rapidly insert the rod in the event of an emergency.

Efficient Design and Analysis of Lightweight Reinforced Core Sandwich and PRSEUS Structures

NASA Technical Reports Server (NTRS)

Bednarcyk, Brett A.; Yarrington, Phillip W.; Lucking, Ryan C.; Collier, Craig S.; Ainsworth, James J.; Toubia, Elias A.

2012-01-01

Design, analysis, and sizing methods for two novel structural panel concepts have been developed and incorporated into the HyperSizer Structural Sizing Software. Reinforced Core Sandwich (RCS) panels consist of a foam core with reinforcing composite webs connecting composite facesheets. Boeing s Pultruded Rod Stitched Efficient Unitized Structure (PRSEUS) panels use a pultruded unidirectional composite rod to provide axial stiffness along with integrated transverse frames and stitching. Both of these structural concepts are ovencured and have shown great promise applications in lightweight structures, but have suffered from the lack of efficient sizing capabilities similar to those that exist for honeycomb sandwich, foam sandwich, hat stiffened, and other, more traditional concepts. Now, with accurate design methods for RCS and PRSEUS panels available in HyperSizer, these concepts can be traded and used in designs as is done with the more traditional structural concepts. The methods developed to enable sizing of RCS and PRSEUS are outlined, as are results showing the validity and utility of the methods. Applications include several large NASA heavy lift launch vehicle structures.

Benchmark Evaluation of HTR-PROTEUS Pebble Bed Experimental Program

DOE PAGES

Bess, John D.; Montierth, Leland; Köberl, Oliver; ...

2014-10-09

Benchmark models were developed to evaluate 11 critical core configurations of the HTR-PROTEUS pebble bed experimental program. Various additional reactor physics measurements were performed as part of this program; currently only a total of 37 absorber rod worth measurements have been evaluated as acceptable benchmark experiments for Cores 4, 9, and 10. Dominant uncertainties in the experimental keff for all core configurations come from uncertainties in the ²³⁵U enrichment of the fuel, impurities in the moderator pebbles, and the density and impurity content of the radial reflector. Calculations of k eff with MCNP5 and ENDF/B-VII.0 neutron nuclear data aremore » greater than the benchmark values but within 1% and also within the 3σ uncertainty, except for Core 4, which is the only randomly packed pebble configuration. Repeated calculations of k eff with MCNP6.1 and ENDF/B-VII.1 are lower than the benchmark values and within 1% (~3σ) except for Cores 5 and 9, which calculate lower than the benchmark eigenvalues within 4σ. The primary difference between the two nuclear data libraries is the adjustment of the absorption cross section of graphite. Simulations of the absorber rod worth measurements are within 3σ of the benchmark experiment values. The complete benchmark evaluation details are available in the 2014 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments.« less

Experimental evaluation of a magnetic torquer rod using an innovative test system

NASA Astrophysics Data System (ADS)

Fakhari Mehrjardi, Mohamad; Mirshams, Mehran

2010-03-01

In today's world satellites have an immense and profound role in a country's financial, social and military development and having the technology of creation and launching satellites is a yard stick to a country's progress. Each satellite, like any other advanced machine is consisted of many subsystems in order to do its mission, among those, the attitude Control subsystem has the duty of stabilizing and orientation. Depending on the type of stabilization and control laws, different actuators like momentum wheels, reaction wheels, magnetic torquers and etcetera are used. Due to its smaller shape and weight, lower cost and minimal power consumption, the magnetic torquer is frequently used in low-earth orbit satellites. A magnetic torquer is consisted of a winding wire and a magnetic core that with the current of electricity passing through the winding wire, a magnetic dipole moment is produced. In reaction to the earth's magnetic field, this moment produces the required torque. Thus, having a broader understanding of the specification of the magnetic torquer before using it in the satellite is quite necessary. As a result, in this paper we try to show how to make such system in the laboratory. A magnetorquer is manufactured that the main idea is to estimate the magnetic dipole moment from the magnetic field measurement by this magnetic torquer. To achieve this, first we talk about the theories of creating such device and test system, then we will delve into the more technical aspects of designing such subsystem. In the end, from the output results, the performance curve of the magnetic torquer is produced and the linear areas and scale coefficients are determined. This paper presents test methodology, experimental setup and test results of manufacturing a torque rod with CK30 ferromagnetic alloy core.

Experimental evaluation of a magnetic torquer rod using an innovative test system

NASA Astrophysics Data System (ADS)

Fakhari Mehrjardi, Mohamad; Mirshams, Mehran

2009-12-01

In today's world satellites have an immense and profound role in a country's financial, social and military development and having the technology of creation and launching satellites is a yard stick to a country's progress. Each satellite, like any other advanced machine is consisted of many subsystems in order to do its mission, among those, the attitude Control subsystem has the duty of stabilizing and orientation. Depending on the type of stabilization and control laws, different actuators like momentum wheels, reaction wheels, magnetic torquers and etcetera are used. Due to its smaller shape and weight, lower cost and minimal power consumption, the magnetic torquer is frequently used in low-earth orbit satellites. A magnetic torquer is consisted of a winding wire and a magnetic core that with the current of electricity passing through the winding wire, a magnetic dipole moment is produced. In reaction to the earth's magnetic field, this moment produces the required torque. Thus, having a broader understanding of the specification of the magnetic torquer before using it in the satellite is quite necessary. As a result, in this paper we try to show how to make such system in the laboratory. A magnetorquer is manufactured that the main idea is to estimate the magnetic dipole moment from the magnetic field measurement by this magnetic torquer. To achieve this, first we talk about the theories of creating such device and test system, then we will delve into the more technical aspects of designing such subsystem. In the end, from the output results, the performance curve of the magnetic torquer is produced and the linear areas and scale coefficients are determined. This paper presents test methodology, experimental setup and test results of manufacturing a torque rod with CK30 ferromagnetic alloy core.

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.

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. 2010 Elsevier Ltd. All rights reserved.

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.

EVALUATION OF DATA OBTAINED ON "MANUFACTURING PROCESS" DEVELOPMENT BUNDLES PD 1 THROUGH 5 PRIOR TO MACHINING OPERATIONS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Frankhouser, W.L.; Eyler, J.H.

1956-07-24

Five reference fuel rod bundles were welded and evaluated dimensionally. Dimensional data are presented for the as-welded condition and for the annealed bundle with spacer strips removed (prior to the final machining operations). The welding sequence developed for Core Manufacturing should provide A'' boundles in respect to rod spacing measurements. It will probably not be possible to meet the same requirements for water channel averages, because the design tolerances are not consistent with some factors inherent to the production process. A method to improve this situation is presented. The data presented were evaluated in a fashion similar to that whichmore » would be used in the proposed scheme. Rods tended to bow resulting in a slightly barrel-shaped'' boundle. It is believed this condition can be overcome by providing special bundle peripheral clamps during annealing. Rod distortion should also be reduced by a redesign and relocation of strip spacers. The new design is proposed. (auth)« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stimpson, Shane G.

In this report, load-follow simulations using VERA-CS with one-way coupling to standalone BISON has been demonstrated including both a single rod with a full cycle of load-follow operations and a quarter-core model with a single month of load-follow.

FY 2016 Status Report on the Modeling of the M8 Calibration Series using MAMMOTH

DOE Office of Scientific and Technical Information (OSTI.GOV)

Baker, Benjamin Allen; Ortensi, Javier; DeHart, Mark David

2016-09-01

This report provides a summary of the progress made towards validating the multi-physics reactor analysis application MAMMOTH using data from measurements performed at the Transient Reactor Test facility, TREAT. The work completed consists of a series of comparisons of TREAT element types (standard and control rod assemblies) in small geometries as well as slotted mini-cores to reference Monte Carlo simulations to ascertain the accuracy of cross section preparation techniques. After the successful completion of these smaller problems, a full core model of the half slotted core used in the M8 Calibration series was assembled. Full core MAMMOTH simulations were comparedmore » to Serpent reference calculations to assess the cross section preparation process for this larger configuration. As part of the validation process the M8 Calibration series included a steady state wire irradiation experiment and coupling factors for the experiment region. The shape of the power distribution obtained from the MAMMOTH simulation shows excellent agreement with the experiment. Larger differences were encountered in the calculation of the coupling factors, but there is also great uncertainty on how the experimental values were obtained. Future work will focus on resolving some of these differences.« less

Evaluation of Neutron Radiography Reactor LEU-Core Start-Up Measurements

DOE PAGES

Bess, John D.; Maddock, Thomas L.; Smolinski, Andrew T.; ...

2014-11-04

Benchmark models were developed to evaluate the cold-critical start-up measurements performed during the fresh core reload of the Neutron Radiography (NRAD) reactor with Low Enriched Uranium (LEU) fuel. Experiments include criticality, control-rod worth measurements, shutdown margin, and excess reactivity for four core loadings with 56, 60, 62, and 64 fuel elements. The worth of four graphite reflector block assemblies and an empty dry tube used for experiment irradiations were also measured and evaluated for the 60-fuel-element core configuration. Dominant uncertainties in the experimental k eff come from uncertainties in the manganese content and impurities in the stainless steel fuel claddingmore » as well as the 236U and erbium poison content in the fuel matrix. Calculations with MCNP5 and ENDF/B-VII.0 neutron nuclear data are approximately 1.4% (9σ) greater than the benchmark model eigenvalues, which is commonly seen in Monte Carlo simulations of other TRIGA reactors. Simulations of the worth measurements are within the 2σ uncertainty for most of the benchmark experiment worth values. The complete benchmark evaluation details are available in the 2014 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments.« less

Evaluation of Neutron Radiography Reactor LEU-Core Start-Up Measurements

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bess, John D.; Maddock, Thomas L.; Smolinski, Andrew T.

Benchmark models were developed to evaluate the cold-critical start-up measurements performed during the fresh core reload of the Neutron Radiography (NRAD) reactor with Low Enriched Uranium (LEU) fuel. Experiments include criticality, control-rod worth measurements, shutdown margin, and excess reactivity for four core loadings with 56, 60, 62, and 64 fuel elements. The worth of four graphite reflector block assemblies and an empty dry tube used for experiment irradiations were also measured and evaluated for the 60-fuel-element core configuration. Dominant uncertainties in the experimental k eff come from uncertainties in the manganese content and impurities in the stainless steel fuel claddingmore » as well as the 236U and erbium poison content in the fuel matrix. Calculations with MCNP5 and ENDF/B-VII.0 neutron nuclear data are approximately 1.4% (9σ) greater than the benchmark model eigenvalues, which is commonly seen in Monte Carlo simulations of other TRIGA reactors. Simulations of the worth measurements are within the 2σ uncertainty for most of the benchmark experiment worth values. The complete benchmark evaluation details are available in the 2014 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments.« less

Neutronic reactor

DOEpatents

Wende, Charles W. J.; Babcock, Dale F.; Menegus, Robert L.

1983-01-01

A nuclear reactor includes an active portion with fissionable fuel and neutron moderating material surrounded by neutron reflecting material. A control element in the active portion includes a group of movable rods constructed of neutron-absorbing material. Each rod is movable with respect to the other rods to vary the absorption of neutrons and effect control over neutron flux.

Controllability of semi-infinite rod heating by a point source

NASA Astrophysics Data System (ADS)

Khurshudyan, A.

2018-04-01

The possibility of control over heating of a semi-infinite thin rod by a point source concentrated at an inner point of the rod, is studied. Quadratic and piecewise constant solutions of the problem are derived, and the possibilities of solving appropriate problems of optimal control are indicated. Determining of the parameters of the piecewise constant solution is reduced to a problem of nonlinear programming. Numerical examples are considered.

Environment-based pin-power reconstruction method for homogeneous core calculations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Leroyer, H.; Brosselard, C.; Girardi, E.

2012-07-01

Core calculation schemes are usually based on a classical two-step approach associated with assembly and core calculations. During the first step, infinite lattice assemblies calculations relying on a fundamental mode approach are used to generate cross-sections libraries for PWRs core calculations. This fundamental mode hypothesis may be questioned when dealing with loading patterns involving several types of assemblies (UOX, MOX), burnable poisons, control rods and burn-up gradients. This paper proposes a calculation method able to take into account the heterogeneous environment of the assemblies when using homogeneous core calculations and an appropriate pin-power reconstruction. This methodology is applied to MOXmore » assemblies, computed within an environment of UOX assemblies. The new environment-based pin-power reconstruction is then used on various clusters of 3x3 assemblies showing burn-up gradients and UOX/MOX interfaces, and compared to reference calculations performed with APOLLO-2. The results show that UOX/MOX interfaces are much better calculated with the environment-based calculation scheme when compared to the usual pin-power reconstruction method. The power peak is always better located and calculated with the environment-based pin-power reconstruction method on every cluster configuration studied. This study shows that taking into account the environment in transport calculations can significantly improve the pin-power reconstruction so far as it is consistent with the core loading pattern. (authors)« less

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.

The slightly-enriched spectral shift control reactor. Final report, September 30, 1988--September 30, 1991

DOE Office of Scientific and Technical Information (OSTI.GOV)

Martin, W.R.; Lee, J.C.; Larsen, E.W.

1991-11-01

An advanced converter reactor design utilizing mechanical spectral shift control rods in a conventional pressurized water reactor configuration is under investigation. The design is based on the principle that a harder spectrum during the early part of the fuel cycle will result in large neutron captures in fertile {sup 238}U, which can then be burned in situ in a softer spectrum later in the cycle. Preliminary design calculations performed during FY 89 showed that the slightly-enriched spectral shift reactor design offers the benefit of substantially increased fuel resource utilization with the proven safety characteristics of the pressurized water reactor technologymore » retained. Optimization of the fuel design and development of fuel management strategies were carried out in FY 90, along with effort to develop and validate neutronic methodology for tight-lattice configurations with hard spectra. During FY 91, the final year of the grant, the final Slightly-Enriched Spectral Shift Reactor (SESSR) design was determined, and reference design analyses were performed for the assemblies as well as the global core configuration, both at the beginning of cycle (BOC) and with depletion. The final SESSR design results in approximately a 20% increase in the utilization of uranium resources, based on equilibrium fuel cycle analyses. Acceptable pin power peaking is obtained with the final core design, with assembly peaking factors equal to less than 1.04 for spectral shift control rods both inserted and withdrawn, and global peaking factors at BOC predicted to be 1.4. In addition, a negative Moderation Temperature Coefficient (MTC) is maintained for BOC, which is difficult to achieve with conventional advanced converter designs based on a closed fuel cycle. The SESSR design avoids the need for burnable poison absorber, although they could be added if desired to increase the cycle length while maintaining a negative MTC.« less

The slightly-enriched spectral shift control reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Martin, W.R.; Lee, J.C.; Larsen, E.W.

1991-11-01

An advanced converter reactor design utilizing mechanical spectral shift control rods in a conventional pressurized water reactor configuration is under investigation. The design is based on the principle that a harder spectrum during the early part of the fuel cycle will result in large neutron captures in fertile {sup 238}U, which can then be burned in situ in a softer spectrum later in the cycle. Preliminary design calculations performed during FY 89 showed that the slightly-enriched spectral shift reactor design offers the benefit of substantially increased fuel resource utilization with the proven safety characteristics of the pressurized water reactor technologymore » retained. Optimization of the fuel design and development of fuel management strategies were carried out in FY 90, along with effort to develop and validate neutronic methodology for tight-lattice configurations with hard spectra. During FY 91, the final year of the grant, the final Slightly-Enriched Spectral Shift Reactor (SESSR) design was determined, and reference design analyses were performed for the assemblies as well as the global core configuration, both at the beginning of cycle (BOC) and with depletion. The final SESSR design results in approximately a 20% increase in the utilization of uranium resources, based on equilibrium fuel cycle analyses. Acceptable pin power peaking is obtained with the final core design, with assembly peaking factors equal to less than 1.04 for spectral shift control rods both inserted and withdrawn, and global peaking factors at BOC predicted to be 1.4. In addition, a negative Moderation Temperature Coefficient (MTC) is maintained for BOC, which is difficult to achieve with conventional advanced converter designs based on a closed fuel cycle. The SESSR design avoids the need for burnable poison absorber, although they could be added if desired to increase the cycle length while maintaining a negative MTC.« less

Layered reactive particles with controlled geometries, energies, and reactivities, and methods for making the same

DOEpatents

Fritz, Gregory M.; Weihs, Timothy P.; Grzyb, Justin A.

2016-07-05

An energetic composite having a plurality of reactive particles each having a reactive multilayer construction formed by successively depositing reactive layers on a rod-shaped substrate having a longitudinal axis, dividing the reactive-layer-deposited rod-shaped substrate into a plurality of substantially uniform longitudinal segments, and removing the rod-shaped substrate from the longitudinal segments, so that the reactive particles have a controlled, substantially uniform, cylindrically curved or otherwise rod-contoured geometry which facilitates handling and improves its packing fraction, while the reactant multilayer construction controls the stability, reactivity and energy density of the energetic composite.

Layered reactive particles with controlled geometries, energies, and reactivities, and methods for making the same

DOEpatents

Fritz, Gregory M; Knepper, Robert Allen; Weihs, Timothy P; Gash, Alexander E; Sze, John S

2013-04-30

An energetic composite having a plurality of reactive particles each having a reactive multilayer construction formed by successively depositing reactive layers on a rod-shaped substrate having a longitudinal axis, dividing the reactive-layer-deposited rod-shaped substrate into a plurality of substantially uniform longitudinal segments, and removing the rod-shaped substrate from the longitudinal segments, so that the reactive particles have a controlled, substantially uniform, cylindrically curved or otherwise rod-contoured geometry which facilitates handling and improves its packing fraction, while the reactant multilayer construction controls the stability, reactivity and energy density of the energetic composite.

Deterministic Modeling of the High Temperature Test Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ortensi, J.; Cogliati, J. J.; Pope, M. A.

2010-06-01

Idaho National Laboratory (INL) is tasked with the development of reactor physics analysis capability of the Next Generation Nuclear Power (NGNP) project. In order to examine INL’s current prismatic reactor deterministic analysis tools, the project is conducting a benchmark exercise based on modeling the High Temperature Test Reactor (HTTR). This exercise entails the development of a model for the initial criticality, a 19 column thin annular core, and the fully loaded core critical condition with 30 columns. Special emphasis is devoted to the annular core modeling, which shares more characteristics with the NGNP base design. The DRAGON code is usedmore » in this study because it offers significant ease and versatility in modeling prismatic designs. Despite some geometric limitations, the code performs quite well compared to other lattice physics codes. DRAGON can generate transport solutions via collision probability (CP), method of characteristics (MOC), and discrete ordinates (Sn). A fine group cross section library based on the SHEM 281 energy structure is used in the DRAGON calculations. HEXPEDITE is the hexagonal z full core solver used in this study and is based on the Green’s Function solution of the transverse integrated equations. In addition, two Monte Carlo (MC) based codes, MCNP5 and PSG2/SERPENT, provide benchmarking capability for the DRAGON and the nodal diffusion solver codes. The results from this study show a consistent bias of 2–3% for the core multiplication factor. This systematic error has also been observed in other HTTR benchmark efforts and is well documented in the literature. The ENDF/B VII graphite and U235 cross sections appear to be the main source of the error. The isothermal temperature coefficients calculated with the fully loaded core configuration agree well with other benchmark participants but are 40% higher than the experimental values. This discrepancy with the measurement stems from the fact that during the experiments the control rods were adjusted to maintain criticality, whereas in the model, the rod positions were fixed. In addition, this work includes a brief study of a cross section generation approach that seeks to decouple the domain in order to account for neighbor effects. This spectral interpenetration is a dominant effect in annular HTR physics. This analysis methodology should be further explored in order to reduce the error that is systematically propagated in the traditional generation of cross sections.« less

Managing Geothermal Exploratory Drilling Risks Drilling Geothermal Exploration and Delineation Wells with Small-Footprint Highly Portable Diamond Core Drills

NASA Astrophysics Data System (ADS)

Tuttle, J.; Listi, R.; Combs, J.; Welch, V.; Reilly, S.

2012-12-01

Small hydraulic core rigs are highly portable (truck or scow-mounted), and have recently been used for geothermal exploration in areas such as Nevada, California, the Caribbean Islands, Central and South America and elsewhere. Drilling with slim diameter core rod below 7,000' is common, with continuous core recovery providing native-state geological information to aid in identifying the resource characteristics and boundaries; this is a highly cost-effective process. Benefits associated with this innovative exploration and delineation technology includes the following: Low initial Capital Equipment Cost and consumables costs Small Footprint, reducing location and road construction, and cleanup costs Supporting drill rod (10'/3meter) and tools are relatively low weight and easily shipped Speed of Mobilization and rig up Reduced requirements for support equipment (cranes, backhoes, personnel, etc) Small mud systems and cementing requirements Continuous, simplified coring capability Depth ratings comparable to that of large rotary rigs (up to ~10,000'+) Remote/small-location accessible (flown into remote areas or shipped in overseas containers) Can be scow or truck-mounted This technical presentation's primary goal is to share the technology of utilizing small, highly portable hydraulic coring rigs to provide exploratory drilling (and in some cases, production drilling) for geothermal projects. Significant cost and operational benefits are possible for the Geothermal Operator, especially for those who are pursuing projects in remote locations or countries, or in areas that are either inaccessible or in which a small footprint is required. John D. Tuttle Sinclair Well Products jtuttle@sinclairwp.com

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ellis, Ronald James

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

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

Optimizing the sensing performance of a single-rod fluxgate magnetometer using thin magnetic wires

NASA Astrophysics Data System (ADS)

Can, Hava; Svec, Peter, Jr.; Tanrıseven, Sercan; Bydzovsky, Jan; Birlikseven, Cengiz; Sözeri, Hüseyin; Svec, Peter, Sr.; Topal, Uğur

2015-11-01

This paper presents the optimal conditions for the design of a single-rod fluxgate magnetometer using Co-based amorphous magnetic wires with reduced geometrical dimensions of 100 μm in diameter. In order to enhance the performance of the current sensor (i.e. the noise level, the sensitivity, the dynamical range, the scaling factor, etc), the core materials were subjected to annealing at different annealing temperatures in a longitudinal magnetic field ranging from 0 to 0.5 T. The B-H measurements have shown that the heat treatments significantly change the magnetic parameters of the cores (the saturation field, the initial and apparent permeabilities). For instance, the initial permeability μ i attains values of between 3500 and 4700 depending on the treatment conditions. These magnetic parameters were subsequently correlated with the sensor performance by using the principles of the fluxgate physics. Consequently, the enhanced fluxgate effect with improved sensing characteristics has been obtained by annealing the wire core at 250 °C (B  =  0 T). It is shown that this magnetic wire with a sensing area of 0.00785 mm2 is suitable as a sensor core for the nondestructive testing of metallic objects and the surfaces of magnetic cards. The sensor signal shows perfect linear dependence to dc or low frequency fields up to ~1 Oe. The fitting parameters R 2 of 0.9998 could be achieved in a dc field interval of  -1.0 Oe and 1.0 Oe (when R 2  =1.0, all points lie exactly on the curve with no scatter). Such linearity has not been seen in such a large dynamical range until now in the rod-type single-core fluxgates. It is also shown that there is no hysteresis on the V 2f -H dc graphs (the V 2f is the sensor signal) even after applying fields as high as 100 Oe. Besides, the cross-field effect is almost zero due to the geometry of the long-thin wire.

Parallel Operation of Multiple Closely Spaced Small Aspect Ratio Rod Pinches

DOE PAGES

Harper-Slaboszewicz, Victor J.; Leckbee, Joshua; Bennett, Nichelle; ...

2014-12-10

A series of simulations and experiments to resolve questions about the operation of arrays of closely spaced small aspect ratio rod pinches has been performed. Design and post-shot analysis of the experimental results are supported by 3D particle-in-cell simulations. Both simulations and experiments support these conclusions. Penetration of current to the interior of the array appears to be efficient, as the current on the center rods is essentially equal to the current on the outer rods. Current loss in the feed due to the formation of magnetic nulls was avoided in these experiments by design of the feed surface ofmore » the cathode and control of the gap to keep the electric fields on the cathode below the emission threshold. Some asymmetry in the electron flow to the rod was observed, but the flow appeared to symmetrize as it reached the end of the rod. Interaction between the rod pinches can be controlled to allow the stable and consistent operation of arrays of rod pinches.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Paulauskas, Tadas; Buurma, Christopher; Colegrove, Eric

Dislocation cores have long dominated the electronic and optical behaviors of semiconductor devices and detailed atomic characterization is required to further explore their effects. Miniaturization of semiconductor devices to nanometre scale also puts emphasis on a material's mechanical properties to withstand failure due to processing or operational stresses. Sessile junctions of dislocations provide barriers to propagation of mobile dislocations and may lead to work-hardening. The sessile Lomer–Cottrell and Hirth lock dislocations, two stable lowest elastic energy stair-rods, are studied in this paper. More specifically, using atomic resolution high-angle annular dark-field imaging and atomic-column-resolved X-ray spectrum imaging in an aberration-corrected scanningmore » transmission electron microscope, dislocation core structures are examined in zinc-blende CdTe. A procedure is outlined for atomic scale analysis of dislocation junctions which allows determination of their identity with specially tailored Burgers circuits and also formation mechanisms of the polar core structures based on Thompson's tetrahedron adapted to reactions of polar dislocations as they appear in CdTe and other zinc-blende solids. Strain fields associated with the dislocations calculatedviageometric phase analysis are found to be diffuse and free of `hot spots' that reflect compact structures and low elastic energy of the pure-edge stair-rods.« less

COMPOSITE CONTROL ROD

DOEpatents

Rock, H.R.

1963-12-24

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

Core Dynamics Analysis for Reactivity Insertion and Loss of Coolant Flow Tests Using the High Temperature Engineering Test Reactor

NASA Astrophysics Data System (ADS)

Takamatsu, Kuniyoshi; Nakagawa, Shigeaki; Takeda, Tetsuaki

Safety demonstration tests using the High Temperature Engineering Test Reactor (HTTR) are in progress to verify its inherent safety features and improve the safety technology and design methodology for High-temperature Gas-cooled Reactors (HTGRs). The reactivity insertion test is one of the safety demonstration tests for the HTTR. This test simulates the rapid increase in the reactor power by withdrawing the control rod without operating the reactor power control system. In addition, the loss of coolant flow tests has been conducted to simulate the rapid decrease in the reactor power by tripping one, two or all out of three gas circulators. The experimental results have revealed the inherent safety features of HTGRs, such as the negative reactivity feedback effect. The numerical analysis code, which was named-ACCORD-, was developed to analyze the reactor dynamics including the flow behavior in the HTTR core. We have modified this code to use a model with four parallel channels and twenty temperature coefficients. Furthermore, we added another analytical model of the core for calculating the heat conduction between the fuel channels and the core in the case of the loss of coolant flow tests. This paper describes the validation results for the newly developed code using the experimental results. Moreover, the effect of the model is formulated quantitatively with our proposed equation. Finally, the pre-analytical result of the loss of coolant flow test by tripping all gas circulators is also discussed.

VIEW OF CABLES AND TAPES ASSOCIATED WITH ADRIVE CONTROL ROD ...

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

VIEW OF CABLES AND TAPES ASSOCIATED WITH A-DRIVE CONTROL ROD SYSTEM, AT LEVEL +15’, DIRECTLY ABOVE PDP CONTROL ROOM, LOOKING NORTH - Physics Assembly Laboratory, Area A/M, Savannah River Site, Aiken, Aiken County, SC

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.

Shaping of the axial power density distribution in the core to minimize the vapor volume fraction at the outlet of the VVER-1200 fuel assemblies

DOE Office of Scientific and Technical Information (OSTI.GOV)

Savander, V. I.; Shumskiy, B. E., E-mail: borisshumskij@yandex.ru; Pinegin, A. A.

The possibility of decreasing the vapor fraction at the VVER-1200 fuel assembly outlet by shaping the axial power density field is considered. The power density field was shaped by axial redistribution of the concentration of the burnable gadolinium poison in the Gd-containing fuel rods. The mathematical modeling of the VVER-1200 core was performed using the NOSTRA computer code.

Tailoring Nd3+ emission spectrum by a neodymium-doped tellurite all-solid photonic bandgap fiber

NASA Astrophysics Data System (ADS)

Tong, Hoang Tuan; Demichi, Daisuke; Suzuki, Takenobu; Ohishi, Yasutake

2018-02-01

A tellurite all-solid photonic bandgap fiber (ASPBF) whose cladding consists of 60 high-index rods arranged periodically around a central core was successfully fabricated. The diameter of high-index rod was about 5.0 μm and the distance between the center of two adjacent high-index rods was approximately 8.0 μm. The high-index rod was made of the TeO2-Li2O-WO3-MoO3-Nb2O5 (TLWMN) glass, the cladding was made of the TeO2-ZnO-Na2O-La2O3 (TZNL) glass as the background glass material and the central core was made of TZNL glass doped with 0.5 wt% of Nd2O3. A supercontinuum light from 0.6 to 2.4 μm was coupled into the core of fiber which is 2.2 cm long to measure its transmission spectrum. High transmission bands were obtained in the vicinity of 0.75 and 1.3 μm but the transmission was suppressed in the wavelength range from 1.0 to 1.06 μm. When a titanium∶Sapphire laser source at 0.75 μm was used, the emission spectrum was obtained with two peaks at 1.06 and 1.33 μm which are attributed to the 4F3/2->4I11/2 and 4F3/2->4I13/2 transitions of Nd3+ ion, respectively. The intensities of those emission peaks were compared with those obtained from a bulk glass having the same doping concentration of Nd3+. The results showed that by using tellurite ASPBF, the intensity of the 1.06-μm emission was suppressed by one-twelfth but the intensity of the 1.33-μm emission was maintained. This feature is very advantageous to filter out the 1.06-μm emission of Nd3+ ion in order to realize practical amplifier devices at 1.3 μm.

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.

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.

Two Novel Phycoerythrin-Associated Linker Proteins in the Marine Cyanobacterium Synechococcus sp. Strain WH8102

PubMed Central

Six, Christophe; Thomas, Jean-Claude; Thion, Laurent; Lemoine, Yves; Zal, Frank; Partensky, Frédéric

2005-01-01

The recent availability of the whole genome of Synechococcus sp. strain WH8102 allows us to have a global view of the complex structure of the phycobilisomes of this marine picocyanobacterium. Genomic analyses revealed several new characteristics of these phycobilisomes, consisting of an allophycocyanin core and rods made of one type of phycocyanin and two types of phycoerythrins (I and II). Although the allophycocyanin appears to be similar to that found commonly in freshwater cyanobacteria, the phycocyanin is simpler since it possesses only one complete set of α and β subunits and two rod-core linkers (CpcG1 and CpcG2). It is therefore probably made of a single hexameric disk per rod. In contrast, we have found two novel putative phycoerythrin-associated linker polypeptides that appear to be specific for marine Synechococcus spp. The first one (SYNW2000) is unusually long (548 residues) and apparently results from the fusion of a paralog of MpeC, a phycoerythrin II linker, and of CpeD, a phycoerythrin-I linker. The second one (SYNW1989) has a more classical size (300 residues) and is also an MpeC paralog. A biochemical analysis revealed that, like MpeC, these two novel linkers were both chromophorylated with phycourobilin. Our data suggest that they are both associated (partly or totally) with phycoerythrin II, and we propose to name SYNW2000 and SYNW1989 MpeD and MpeE, respectively. We further show that acclimation of phycobilisomes to high light leads to a dramatic reduction of MpeC, whereas the two novel linkers are not significantly affected. Models for the organization of the rods are proposed. PMID:15716439

Two novel phycoerythrin-associated linker proteins in the marine cyanobacterium Synechococcus sp. strain WH8102.

PubMed

Six, Christophe; Thomas, Jean-Claude; Thion, Laurent; Lemoine, Yves; Zal, Frank; Partensky, Frédéric

2005-03-01

The recent availability of the whole genome of Synechococcus sp. strain WH8102 allows us to have a global view of the complex structure of the phycobilisomes of this marine picocyanobacterium. Genomic analyses revealed several new characteristics of these phycobilisomes, consisting of an allophycocyanin core and rods made of one type of phycocyanin and two types of phycoerythrins (I and II). Although the allophycocyanin appears to be similar to that found commonly in freshwater cyanobacteria, the phycocyanin is simpler since it possesses only one complete set of alpha and beta subunits and two rod-core linkers (CpcG1 and CpcG2). It is therefore probably made of a single hexameric disk per rod. In contrast, we have found two novel putative phycoerythrin-associated linker polypeptides that appear to be specific for marine Synechococcus spp. The first one (SYNW2000) is unusually long (548 residues) and apparently results from the fusion of a paralog of MpeC, a phycoerythrin II linker, and of CpeD, a phycoerythrin-I linker. The second one (SYNW1989) has a more classical size (300 residues) and is also an MpeC paralog. A biochemical analysis revealed that, like MpeC, these two novel linkers were both chromophorylated with phycourobilin. Our data suggest that they are both associated (partly or totally) with phycoerythrin II, and we propose to name SYNW2000 and SYNW1989 MpeD and MpeE, respectively. We further show that acclimation of phycobilisomes to high light leads to a dramatic reduction of MpeC, whereas the two novel linkers are not significantly affected. Models for the organization of the rods are proposed.

Polarization-resolved micro-photoluminescence investigation of InGaN/GaN core-shell microrods

NASA Astrophysics Data System (ADS)

Mounir, Christian; Schimpke, Tilman; Rossbach, Georg; Avramescu, Adrian; Strassburg, Martin; Schwarz, Ulrich T.

2017-01-01

We investigate the optical emission properties of the active InGaN shell of high aspect-ratio InGaN/GaN core-shell microrods (μRods) by confocal quasi-resonant polarization-resolved and excitation density dependent micro-photoluminescence (μPL). The active shell, multiple thin InGaN/GaN quantum wells (MQWs), was deposited on GaN μRods selectively grown by metal organic vapor phase epitaxy on patterned SiO2/n-GaN/sapphire template. High spatial resolution mappings reveal a very homogeneous emission intensity along the whole μRods including the tip despite a red-shift of 30 nm from the base to the tip along the 8.6 μm-long m-plane sidewalls. Looking at the Fabry-Perot interference fringes superimposed on the μPL spectra, we get structural information on the μRods. A high degree of linear polarization (DLP) of 0.6-0.66 is measured on the lower half of the m-plane side facets with a slight decrease toward the tip. We observe the typical drop of the DLP with an excitation density caused by degenerate filling of valence bands (Fermi regime). Local internal quantum efficiencies (IQEs) of 55 ±11 % up to 73 ±7 % are estimated on the m-plane facet from measurements at low temperature. Finally, simultaneously fitting the DLP and IQE as a function of the excitation density, we determine the carrier density inside the active region and the recombination rate coefficients of the m-plane MQWs. We show that phase-space filling and the background carrier density have to be included in the recombination rate model.

Development of 3D pseudo pin-by-pin calculation methodology in ANC

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zhang, B.; Mayhue, L.; Huria, H.

2012-07-01

Advanced cores and fuel assembly designs have been developed to improve operational flexibility, economic performance and further enhance safety features of nuclear power plants. The simulation of these new designs, along with strong heterogeneous fuel loading, have brought new challenges to the reactor physics methodologies currently employed in the industrial codes for core analyses. Control rod insertion during normal operation is one operational feature in the AP1000{sup R} plant of Westinghouse next generation Pressurized Water Reactor (PWR) design. This design improves its operational flexibility and efficiency but significantly challenges the conventional reactor physics methods, especially in pin power calculations. Themore » mixture loading of fuel assemblies with significant neutron spectrums causes a strong interaction between different fuel assembly types that is not fully captured with the current core design codes. To overcome the weaknesses of the conventional methods, Westinghouse has developed a state-of-the-art 3D Pin-by-Pin Calculation Methodology (P3C) and successfully implemented in the Westinghouse core design code ANC. The new methodology has been qualified and licensed for pin power prediction. The 3D P3C methodology along with its application and validation will be discussed in the paper. (authors)« less

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

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 Office of Scientific and Technical Information (OSTI.GOV)

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

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

Analysis of the mechanical response of biomimetic materials with highly oriented microstructures through 3D printing, mechanical testing and modeling.

PubMed

de Obaldia, Enrique Escobar; Jeong, Chanhue; Grunenfelder, Lessa Kay; Kisailus, David; Zavattieri, Pablo

2015-08-01

Many biomineralized organisms have evolved highly oriented nanostructures to perform specific functions. One key example is the abrasion-resistant rod-like microstructure found in the radular teeth of Chitons (Cryptochiton stelleri), a large mollusk. The teeth consist of a soft core and a hard shell that is abrasion resistant under extreme mechanical loads with which they are subjected during the scraping process. Such remarkable mechanical properties are achieved through a hierarchical arrangement of nanostructured magnetite rods surrounded with α-chitin. We present a combined biomimetic approach in which designs were analyzed with additive manufacturing, experiments, analytical and computational models to gain insights into the abrasion resistance and toughness of rod-like microstructures. Staggered configurations of hard hexagonal rods surrounded by thin weak interfacial material were printed, and mechanically characterized with a cube-corner indenter. Experimental results demonstrate a higher contact resistance and stiffness for the staggered alignments compared to randomly distributed fibrous materials. Moreover, we reveal an optimal rod aspect ratio that lead to an increase in the site-specific properties measured by indentation. Anisotropy has a significant effect (up to 50%) on the Young's modulus in directions parallel and perpendicular to the longitudinal axis of the rods, and 30% on hardness and fracture toughness. Optical microscopy suggests that energy is dissipated in the form of median cracks when the load is parallel to the rods and lateral cracks when the load is perpendicular to the rods. Computational models suggest that inelastic deformation of the rods at early stages of indentation can vary the resistance to penetration. As such, we found that the mechanical behavior of the system is influenced by interfacial shear strain which influences the lateral load transfer and therefore the spread of damage. This new methodology can help to elucidate the evolutionary designs of biomineralized microstructures and understand the tolerance to fracture and damage of chiton radular teeth. Copyright © 2015 Elsevier Ltd. All rights reserved.

Simulation of parameters of hydraulic drive with volumetric type controller

NASA Astrophysics Data System (ADS)

Mulyukin, V. L.; Boldyrev, A. V.; Karelin, D. L.; Belousov, A. M.

2017-09-01

The article presents a mathematical model of volumetric type hydraulic drive controller that allows to calculate the parameters of forward and reverse motion. According to the results of simulation static characteristics of rod’s speed and the force of the hydraulic cylinder rod were built and the influence of the angle of swash plate of the controller at the characteristics profile is shown. The results analysis showed that the proposed controller allows steplessly adjust the speed□ц of hydraulic cylinder’s rod motion and the force developed on the rod without the use of flow throttling.

Deterministic Modeling of the High Temperature Test Reactor with DRAGON-HEXPEDITE

DOE Office of Scientific and Technical Information (OSTI.GOV)

J. Ortensi; M.A. Pope; R.M. Ferrer

2010-10-01

The Idaho National Laboratory (INL) is tasked with the development of reactor physics analysis capability of the Next Generation Nuclear Power (NGNP) project. In order to examine the INL’s current prismatic reactor analysis tools, the project is conducting a benchmark exercise based on modeling the High Temperature Test Reactor (HTTR). This exercise entails the development of a model for the initial criticality, a 19 fuel column thin annular core, and the fully loaded core critical condition with 30 fuel columns. Special emphasis is devoted to physical phenomena and artifacts in HTTR that are similar to phenomena and artifacts in themore » NGNP base design. The DRAGON code is used in this study since it offers significant ease and versatility in modeling prismatic designs. DRAGON can generate transport solutions via Collision Probability (CP), Method of Characteristics (MOC) and Discrete Ordinates (Sn). A fine group cross-section library based on the SHEM 281 energy structure is used in the DRAGON calculations. The results from this study show reasonable agreement in the calculation of the core multiplication factor with the MC methods, but a consistent bias of 2–3% with the experimental values is obtained. This systematic error has also been observed in other HTTR benchmark efforts and is well documented in the literature. The ENDF/B VII graphite and U235 cross sections appear to be the main source of the error. The isothermal temperature coefficients calculated with the fully loaded core configuration agree well with other benchmark participants but are 40% higher than the experimental values. This discrepancy with the measurement partially stems from the fact that during the experiments the control rods were adjusted to maintain criticality, whereas in the model, the rod positions were fixed. In addition, this work includes a brief study of a cross section generation approach that seeks to decouple the domain in order to account for neighbor effects. This spectral interpenetration is a dominant effect in annular HTR physics. This analysis methodology should be further explored in order to reduce the error that is systematically propagated in the traditional generation of cross sections.« less

Qualification and characterization of electronics of the fast neutron Hodoscope detectors using neutrons from CABRI core

NASA Astrophysics Data System (ADS)

Mirotta, S.; Guillot, J.; Chevalier, V.; Biard, B.

2018-01-01

The study of Reactivity Initiated Accidents (RIA) is important to determine up to which limits nuclear fuels can withstand such accidents without clad failure. The CABRI International Program (CIP), conducted by IRSN under an OECD/NEA agreement, has been launched to perform representative RIA Integral Effect Tests (IET) on real irradiated fuel rods in prototypical Pressurized Water Reactors (PWR) conditions. For this purpose, the CABRI experimental pulse reactor, operated by CEA in Cadarache, France, has been strongly renovated, and equipped with a pressurized water loop. The behavior of the test rod, located in that loop in the center of the driver core, is followed in real time during the power transients thanks to the hodoscope, a unique online fuel motion monitoring system, and one of the major distinctive features of CABRI. The hodoscope measures the fast neutrons emitted by the tested rod during the power pulse with a complete set of 153 Fission Chambers and 153 Proton Recoil Counters. During the CABRI facility renovation, the electronic chain of these detectors has been upgraded. In this paper, the performance of the new system is presented describing gain calibration methodology in order to get maximal Signal/Noise ratio for amplification modules, threshold tuning methodology for the discrimination modules (old and new ones), and linear detectors response limit versus different reactor powers for the whole electronic chain.

Neural net controlled tag gas sampling system for nuclear reactors

DOEpatents

Gross, Kenneth C.; Laug, Matthew T.; Lambert, John D. B.; Herzog, James P.

1997-01-01

A method and system for providing a tag gas identifier to a nuclear fuel rod and analyze escaped tag gas to identify a particular failed nuclear fuel rod. The method and system include disposing a unique tag gas composition into a plenum of a nuclear fuel rod, monitoring gamma ray activity, analyzing gamma ray signals to assess whether a nuclear fuel rod has failed and is emitting tag gas, activating a tag gas sampling and analysis system upon sensing tag gas emission from a failed nuclear rod and evaluating the escaped tag gas to identify the particular failed nuclear fuel rod.

Pulsed ultrasonic stir welding method

NASA Technical Reports Server (NTRS)

Ding, R. Jeffrey (Inventor)

2013-01-01

A method of performing ultrasonic stir welding uses a welding head assembly to include a plate and a rod passing through the plate. The rod is rotatable about a longitudinal axis thereof. In the method, the rod is rotated about its longitudinal axis during a welding operation. During the welding operation, a series of on-off ultrasonic pulses are applied to the rod such that they propagate parallel to the rod's longitudinal axis. At least a pulse rate associated with the on-off ultrasonic pulses is controlled.

Dynamic Rod Worth Measurement

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chao, Y.A.; Chapman, D.M.; Hill, D.J.

2000-12-15

The dynamic rod worth measurement (DRWM) technique is a method of quickly validating the predicted bank worth of control rods and shutdown rods. The DRWM analytic method is based on three-dimensional, space-time kinetic simulations of the rapid rod movements. Its measurement data is processed with an advanced digital reactivity computer. DRWM has been used as the method of bank worth validation at numerous plant startups with excellent results. The process and methodology of DRWM are described, and the measurement results of using DRWM are presented.

Methods and codes for neutronic calculations of the MARIA research reactor.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Andrzejewski, K.; Kulikowska, T.; Bretscher, M. M.

2002-02-18

The core of the MARIA high flux multipurpose research reactor is highly heterogeneous. It consists of beryllium blocks arranged in 6 x 8 matrix, tubular fuel assemblies, control rods and irradiation channels. The reflector is also heterogeneous and consists of graphite blocks clad with aluminum. Its structure is perturbed by the experimental beam tubes. This paper presents methods and codes used to calculate the MARIA reactor neutronics characteristics and experience gained thus far at IAE and ANL. At ANL the methods of MARIA calculations were developed in connection with the RERTR program. At IAE the package of programs was developedmore » to help its operator in optimization of fuel utilization.« less

High power solid state laser modulator

DOEpatents

Birx, Daniel L.; Ball, Don G.; Cook, Edward G.

2004-04-27

A multi-stage magnetic modulator provides a pulse train of .+-.40 kV electrical pulses at a 5-7 kHz repetition rate to a metal vapor laser. A fractional turn transformer steps up the voltage by a factor of 80 to 1 and magnetic pulse compression is used to reduce the pulse width of the pulse train. The transformer is fabricated utilizing a rod and plate stack type of construction to achieve a high packing factor. The pulses are controlled by an SCR stack where a plurality of SCRs are electrically connected in parallel, each SCR electrically connected to a saturable inductor, all saturable inductors being wound on the same core of magnetic material for enhanced power handling characteristics.

Gamma-Radiation-Induced Degradation of Actively Pumped Single-Mode Ytterbium-Doped Optical Laser - Postprint

DTIC Science & Technology

2015-01-01

evaluated using the cobalt (Co)-60 gamma irradiation facility at The Ohio State University. A radiation dose rate of 43 krad(Si)/hr was used to expose the...Table 1. Description of the optical fibers used for in-situ analysis of the radiation damage Optical fiber Core Dopant Core/cladding diameters (μm...University is a pool-type gamma irradiation facility using a common cobalt cylindrical rod irradiator submerged 20 feet into a water tank. A

77 FR 36137 - Airworthiness Directives; AGUSTA S.p.A. Helicopters

Federal Register 2010, 2011, 2012, 2013, 2014

2012-06-18

... the upper end of collective control rod C2 to torque tube C3 is properly installed. This AD is... prevent separation of the collective control rod from the torque tube, loss of control of the collective... helicopters because the production quality control procedures did not require recording the applied torque on...

Urea controlled hydrothermal synthesis of ammonium aluminum carbonate hydroxide rods

NASA Astrophysics Data System (ADS)

Wang, Fang; Zhu, Jianfeng; Liu, Hui

2018-03-01

In this study, ammonium aluminum carbonate hydroxide (AACH) rods were controllably prepared using the hydrothermal method by manipulating the amount of urea in the reaction system. The experimental results showed that AACH in rod shape was able to be gradually transformed from γ-AlOOH in cluster shape during the molar ratios of urea to Al in the reactants were ranged from 8 to 10, and the yield of AACH has increased accordingly. When the molar ratio of urea to Al reaches 11, pure AACH rods with a diameter of 500 nm and a length of 10 μm approximately was able to be produced. Due to the slow decomposition of urea during the hydrothermal reaction, the nucleation and growth of AACH crystal proceed step by step. Therefore, the crystal can fully grow on each crystal plane and eventually produce a highly crystalline rod-shaped product. The role of urea in controlling the morphology and yield of AACH was also discussed in this paper systematically.

Study of the Optimum Zone of the Independent Variables of an ORGEL Reactor Connected to a 250-MWeb Power Plant. Self Supporting Fuel Elements Made of UC, with Sap Cladding with Four Fuel Rods and Individual Pressure Tubes; STUDIE DER OPTIMALEN ZONE DER UNABHANGIGEN PARAMETER EINES ORGEL- REAKTORS IN EINEM 250-MWe-KRAFTWERK. SELBST-TRAGENDES BRENNELEMENT AUS UC, SAP-UMHUL-LUNG MIT 4 BRENNSTOFFSTABEN UND INDIVIDUELLEN DRUCKROHREN

DOE Office of Scientific and Technical Information (OSTI.GOV)

LaFontaine, F.; Tauch, P.

The optimum range of the independent variables of and ORGEL reactor connected to a 250-Mw power plant (4 fuel rods of UC with individual pressure tubes), as well as the geometry of the reactor core and the operation of the plant, is described. (auth)

DYNAMIC AND STATIC PARAMETERS OF THE AQUEOUS HOMOGENEOUS ARMOUR RESEARCH REACTOR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Terrell, C.W.; McElroy, W.N.

1959-06-01

A brief description of the aqueous homogeneous Armour Research Reactor is given. The negative reactivity coefficient resulting from a temperature increase was determined over a fuel temperature range of 37 to 150 deg F. Possession of an accurately calibrated rod and temperature coefficient permitted a direct measurement of the void coefficient. The reactor was taken to different power levels, and from the calibrated rod the total reduction in excess reactivity was obtained. During the power increase program additional U/sup 235/ and water were added to the core to determine the worth of U/sup 235/ and water. (W.D.M.)

Ultrafast exciton dynamics and light-driven H2 evolution in colloidal semiconductor nanorods and Pt-tipped nanorods.

PubMed

Wu, Kaifeng; Zhu, Haiming; Lian, Tianquan

2015-03-17

Colloidal quantum confined one-dimensional (1D) semiconductor nanorods (NRs) and related semiconductor-metal heterostructures are promising new materials for efficient solar-to-fuel conversion because of their unique physical and chemical properties. NRs can simultaneously exhibit quantum confinement effects in the radial direction and bulk like carrier transport in the axial direction. The former implies that concepts well-established in zero-dimensional quantum dots, such as size-tunable energetics and wave function engineering through band alignment in heterostructures, can also be applied to NRs; while the latter endows NRs with fast carrier transport to achieve long distance charge separation. Selective growth of catalytic metallic nanoparticles, such as Pt, at the tips of NRs provides convenient routes to multicomponent heterostructures with photocatalytic capabilities and controllable charge separation distances. The design and optimization of such materials for efficient solar-to-fuel conversion require the understanding of exciton and charge carrier dynamics. In this Account, we summarize our recent studies of ultrafast charge separation and recombination kinetics and their effects on steady-state photocatalytic efficiencies of colloidal CdS and CdSe/CdS NRs and related NR-Pt heterostructures. After a brief introduction of their electronic structure, we discuss exciton dynamics of CdS NRs. By transient absorption and time-resolved photoluminescence decay, it is shown that although the conduction band electrons are long-lived, photogenerated holes in CdS NRs are trapped on an ultrafast time scale (∼0.7 ps), which forms localized excitons due to strong Coulomb interaction in 1D NRs. In quasi-type II CdSe/CdS dot-in-rod NRs, a large valence band offset drives the ultrafast localization of holes to the CdSe core, and the competition between this process and ultrafast hole trapping on a CdS rod leads to three types of exciton species with distinct spatial distributions. The effect of the exciton dynamics on photoreduction reactions is illustrated using methyl viologen (MV(2+)) as a model electron acceptor. The steady-state MV(2+) photoreduction quantum yield of CdSe/CdS dot-in-rod NRs approaches unity under rod excitation, much larger than CdSe QDs and CdSe/CdS core/shell QDs. Detailed time-resolved studies show that in quasi-type II CdSe/CdS NRs and type II ZnSe/CdS NRs strong quantum confinement in the radial direction facilitates fast electron transfer and hole removal, whereas the fast carrier mobility along the axial direction enables long distance charge separation and slow charge recombination, which is essential for efficient MV(2+) photoreduction. The NR/MV(2+) relay system can be coupled to Pt nanoparticles in solution for light-driven H2 generation. Alternatively, Pt-tipped CdS and CdSe/CdS NRs provide fully integrated all inorganic systems for light-driven H2 generation. In CdS-Pt and CdSe/CdS-Pt hetero-NRs, ultrafast hole trapping on the CdS rod surface or in CdSe core enables efficient electron transfer from NRs to Pt tips by suppressing hole and energy transfer. It is shown that the quantum yields of photodriven H2 generation using these heterostructures correlate well with measured hole transfer rates from NRs to sacrificial donors, revealing that hole removal is the key efficiency-limiting step. These findings provide important insights for designing more efficient quantum confined NR and NR-Pt based systems for solar-to-fuel conversion.

Benchmark Evaluation of Fuel Effect and Material Worth Measurements for a Beryllium-Reflected Space Reactor Mockup

DOE Office of Scientific and Technical Information (OSTI.GOV)

Marshall, Margaret A.; Bess, John D.

2015-02-01

The critical configuration of the small, compact critical assembly (SCCA) experiments performed at the Oak Ridge Critical Experiments Facility (ORCEF) in 1962-1965 have been evaluated as acceptable benchmark experiments for inclusion in the International Handbook of Evaluated Criticality Safety Benchmark Experiments. The initial intent of these experiments was to support the design of the Medium Power Reactor Experiment (MPRE) program, whose purpose was to study “power plants for the production of electrical power in space vehicles.” The third configuration in this series of experiments was a beryllium-reflected assembly of stainless-steel-clad, highly enriched uranium (HEU)-O 2 fuel mockup of a potassium-cooledmore » space power reactor. Reactivity measurements cadmium ratio spectral measurements and fission rate measurements were measured through the core and top reflector. Fuel effect worth measurements and neutron moderating and absorbing material worths were also measured in the assembly fuel region. The cadmium ratios, fission rate, and worth measurements were evaluated for inclusion in the International Handbook of Evaluated Criticality Safety Benchmark Experiments. The fuel tube effect and neutron moderating and absorbing material worth measurements are the focus of this paper. Additionally, a measurement of the worth of potassium filling the core region was performed but has not yet been evaluated Pellets of 93.15 wt.% enriched uranium dioxide (UO 2) were stacked in 30.48 cm tall stainless steel fuel tubes (0.3 cm tall end caps). Each fuel tube had 26 pellets with a total mass of 295.8 g UO 2 per tube. 253 tubes were arranged in 1.506-cm triangular lattice. An additional 7-tube cluster critical configuration was also measured but not used for any physics measurements. The core was surrounded on all side by a beryllium reflector. The fuel effect worths were measured by removing fuel tubes at various radius. An accident scenario was also simulated by moving outward twenty fuel rods from the periphery of the core so they were touching the core tank. The change in the system reactivity when the fuel tube(s) were removed/moved compared with the base configuration was the worth of the fuel tubes or accident scenario. The worth of neutron absorbing and moderating materials was measured by inserting material rods into the core at regular intervals or placing lids at the top of the core tank. Stainless steel 347, tungsten, niobium, polyethylene, graphite, boron carbide, aluminum and cadmium rods and/or lid worths were all measured. The change in the system reactivity when a material was inserted into the core is the worth of the material.« less

A prospective randomized controlled trial comparing early postoperative complications in patients undergoing loop colostomy with and without a stoma rod.

PubMed

Franklyn, J; Varghese, G; Mittal, R; Rebekah, G; Jesudason, M R; Perakath, B

2017-07-01

A stoma rod or bridge has been traditionally placed under the bowel loop while constructing a loop colostomy. This is believed to prevent stomal retraction and provide better faecal diversion. However, the rod can cause complications such as mucosal congestion, oedema and necrosis. This single-centre prospective randomized controlled trial compared outcomes after creation of loop colostomy with and without a supporting stoma rod. The primary outcome studied was stoma retraction rate; other stoma-related complications were studied as secondary outcomes. One hundred and fifty-one patients were randomly allotted to one of two arms, colostomy with or without a supporting rod. Postoperative complications such as retraction, mucocutaneous separation, congestion and re-exploration for stoma-related complications were recorded. There was no difference in the stoma retraction rate between the two arms (8.1% in the rod arm and 6.6% in the no-rod arm; P = 0.719). Stomal necrosis (10.7% vs 1.3%; P = 0.018), oedema (23% vs 3.9%; P = 0.001), congestion (20.3% vs 2.6%; P = 0.001) and re-admission rates (8.5% vs 0%; P = 0.027) were significantly increased in the arm randomized to the rod. The stoma rod does not prevent stomal retraction. However, complication rates are significantly higher when a stoma rod is used. Routine use of a stoma rod for construction of loop colostomy can be avoided. Colorectal Disease © 2017 The Association of Coloproctology of Great Britain and Ireland.

Highly sensitive and robust peroxidase-like activity of Au-Pt core/shell nanorod-antigen conjugates for measles virus diagnosis.

PubMed

Long, Lin; Liu, Jianbo; Lu, Kaishun; Zhang, Tao; Xie, Yunqing; Ji, Yinglu; Wu, Xiaochun

2018-05-02

As a promising candidate for artificial enzymes, catalytically active nanomaterials show several advantages over natural enzymes, such as controlled synthesis at low cost, tunability of catalytic activities, and high stability under stringent conditions. Rod-shaped Au-Pt core/shell nanoparticles (Au@Pt NRs), prepared by Au nanorod-mediated growth, exhibit peroxidase-like activities and could serve as an inexpensive replacement for horseradish peroxidase, with potential applications in various bio-detections. The determination of measles virus is accomplished by a capture-enzyme-linked immunosorbent assay (ELISA) using Au@Pt NR-antigen conjugates. Based on the enhanced catalytic properties of this nanozyme probe, a linear response was observed up to 10 ng/mL measles IgM antibodies in human serum, which is 1000 times more sensitive than commercial ELISA. Hence, these findings provide positive proof of concept for the potential of Au@Pt NR-antigen conjugates in the development of colorimetric biosensors that are simple, robust, and cost-effective.

Homeostatic Plasticity Mediated by Rod-Cone Gap Junction Coupling in Retinal Degenerative Dystrophic RCS Rats

PubMed Central

Hou, Baoke; Fu, Yan; Weng, Chuanhuang; Liu, Weiping; Zhao, Congjian; Yin, Zheng Qin

2017-01-01

Rod-cone gap junctions open at night to allow rod signals to pass to cones and activate the cone-bipolar pathway. This enhances the ability to detect large, dim objects at night. This electrical synaptic switch is governed by the circadian clock and represents a novel form of homeostatic plasticity that regulates retinal excitability according to network activity. We used tracer labeling and ERG recording in the retinae of control and retinal degenerative dystrophic RCS rats. We found that in the control animals, rod-cone gap junction coupling was regulated by the circadian clock via the modulation of the phosphorylation of the melatonin synthetic enzyme arylalkylamine N-acetyltransferase (AANAT). However, in dystrophic RCS rats, AANAT was constitutively phosphorylated, causing rod-cone gap junctions to remain open. A further b/a-wave ratio analysis revealed that dystrophic RCS rats had stronger synaptic strength between photoreceptors and bipolar cells, possibly because rod-cone gap junctions remained open. This was despite the fact that a decrease was observed in the amplitude of both a- and b-waves as a result of the progressive loss of rods during early degenerative stages. These results suggest that electric synaptic strength is increased during the day to allow cone signals to pass to the remaining rods and to be propagated to rod bipolar cells, thereby partially compensating for the weak visual input caused by the loss of rods. PMID:28473754

Loss-of-Flow and Loss-of-Pressure Simulations of the BR2 Research Reactor with HEU and LEU Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Licht, J.; Bergeron, A.; Dionne, B.

2016-01-01

Belgian Reactor 2 (BR2) is a research and test reactor located in Mol, Belgium and is primarily used for radioisotope production and materials testing. The Materials Management and Minimization (M3) Reactor Conversion Program of the National Nuclear Security Administration (NNSA) is supporting the conversion of the BR2 reactor from Highly Enriched Uranium (HEU) fuel to Low Enriched Uranium (LEU) fuel. The reactor core of BR2 is located inside a pressure vessel that contains 79 channels in a hyperboloid configuration. The core configuration is highly variable as each channel can contain a fuel assembly, a control or regulating rod, an experimentalmore » device, or a beryllium or aluminum plug. Because of this variability, a representative core configuration, based on current reactor use, has been defined for the fuel conversion analyses. The code RELAP5/Mod 3.3 was used to perform the transient thermal-hydraulic safety analyses of the BR2 reactor to support reactor conversion. The input model has been modernized relative to that historically used at BR2 taking into account the best modeling practices developed by Argonne National Laboratory (ANL) and BR2 engineers.« less

Analysis of the ORNL/TSF GCFR Grid-Plate Shield Design Confirmation Experiment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Slater, C.O.; Cramer, S.N.; Ingersoll, D.T.

1979-08-01

The results of the analysis of the GCFR Grid-Plate Shield Design Confirmation Experiment are presented. The experiment, performed at the ORNL Tower Shielding Facility, was designed to test the adequacy of methods and data used in the analysis of the GCFR design. In particular, the experiment tested the adequacy of methods to calculate: (1) axial neutron streaming in the GCFR core and axial blanket, (2) the amount and location of the maximum fast-neutron exposure to the grid plate, and (3) the neutron source leaving the top of the grid plate and entering the upper plenum. Other objectives of the experimentmore » were to verify the grid-plate shielding effectiveness and to assess the effects of fuel-pin and subassembly spacing on radiation levels in the GCFR. The experimental mockups contained regions representing the GCFR core/blanket region, the grid-plate shield section, and the grid plate. Most core design options were covered by allowing: (1) three different spacings between fuel subassemblies, (2) two different void fractions within a subassembly by variation of the number of fuel pins, and (3) a mockup of a control-rod channel.« less

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)

Neural net controlled tag gas sampling system for nuclear reactors

DOEpatents

Gross, K.C.; Laug, M.T.; Lambert, J.B.; Herzog, J.P.

1997-02-11

A method and system are disclosed for providing a tag gas identifier to a nuclear fuel rod and analyze escaped tag gas to identify a particular failed nuclear fuel rod. The method and system include disposing a unique tag gas composition into a plenum of a nuclear fuel rod, monitoring gamma ray activity, analyzing gamma ray signals to assess whether a nuclear fuel rod has failed and is emitting tag gas, activating a tag gas sampling and analysis system upon sensing tag gas emission from a failed nuclear rod and evaluating the escaped tag gas to identify the particular failed nuclear fuel rod. 12 figs.

Sample Federal Facility Land Use Control ROD Checklist and Suggested Language (LUC Checklist)

EPA Pesticide Factsheets

The LUC Checklist provides direction on describing and documenting land use controls (LUCs) in federal facility actrions under CERCLA in Records of Decision (RODs), remedial designs (RDs), and remedial action work plans (RAWPs).

ADVANCED DESIGNS OF MAGNETIC JACK-TYPE CONTROL ROD DRIVE

DOE Office of Scientific and Technical Information (OSTI.GOV)

Young, J.N.

1959-11-01

The magnetic jack is a device for positioning the control rods In a nuclear reactor, especially in a reactor containing water under pressure. Magnetic actuation precludes the need for shaft seals and eliminates the problems associated with mechanisms operating in water. It consists of a pressure shell, four sets of external stationary magnet coils (hold, grip, lift, pull down), and one Internal moving part (ammature) that impants linear motion to a cluster of rods. (W.L.H.)

Development of Improved Burnable Poisons for Commercial Nuclear Power Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Renier, J.A.

2002-04-17

Burnable poisons are used in all modern nuclear reactors to permit higher loading of fuel without the necessity of an overly large control rod system. This not only permits a longer core life but can also be used to level the power distribution. Commercial nuclear reactors commonly use B{sub 4}C in separate non-fueled rods and more recently, zirconium boride coatings on the fuel pellets or gadolinium oxide mixed with the fuel. Although the advantages are great, there are problems with using these materials. Boron, which is an effective neutron absorber, transmutes to lithium and helium upon absorption of a neutron.more » Helium is insoluble and is eventually released to the interior of the fuel rod, where it produces an internal pressure. When sufficiently high, this pressure stress could cause separation of the cladding from the fuel, causing overly high centerline temperatures. Gadolinium has several very strongly absorbing isotopes, but not all have large cross sections and result in residual burnable poison reactivity worth at the end of the fuel life. Even if the amount of this residual absorber is small and the penalty in operation small, the cost of this penalty, even if only several days, can be very high. The objective of this investigation was to study the performance of single isotopes in order to reduce the residual negative reactivity left over at the end of the fuel cycle. Since the behavior of burnable poisons can be strongly influenced by their configuration, four forms for the absorbers were studied: homogeneously mixed with the fuel, mixed with only the outer one-third of the fuel pellet, coated on the perimeter of the fuel pellets, and alloyed with the cladding. In addition, the numbers of fuel rods containing burnable poison were chosen as 8, 16, 64, and 104. Other configurations were chosen for a few special cases. An enrichment of 4.5 wt% {sup 235}U was chosen for most cases for study in order to achieve a 4-year fuel cycle. A standard pressurized water reactor fuel core was chosen for the study, and state-of-the-art neutronic reactor core computer codes were used for analysis. Power distribution, fuel burnup, reactivity due to burnable poisons and other fission products, spectrum shift, core reactivity, moderator void coefficients, as well as other parameters were calculated as a function of time and fuel burnup. The results not only showed advantages of separation of burnable poison isotopes but revealed benefits to be achieved by careful selection of the configuration of even naturally occurring elements used as burnable poisons. The savings in terms of additional days of operation is shown in Figure 1, where the savings is plotted for each of six favorable isotopes in the four configurations. The benefit of isotope separation is most dramatic for dysprosium, but even the time savings in the case of gadolinium is several days. For a modern nuclear plant, one day's worth of electricity is worth about one million dollars, so the resulting savings of only a few days is considerable. It is also apparent that the amount of savings depends upon the configuration of the burnable poison.« less

Development of hybrid braided composite rods for reinforcement and health monitoring of structures.

PubMed

Rana, Sohel; Zdraveva, Emilija; Pereira, Cristiana; Fangueiro, Raul; Correia, A Gomes

2014-01-01

In the present study, core-reinforced braided composite rods (BCRs) were developed and characterized for strain sensing capability. A mixture of carbon and glass fibre was used in the core, which was surrounded by a braided cover of polyester fibres. Three compositions of core with different carbon fibre/glass fibre weight ratios (23/77, 47/53, and 100/0) were studied to find out the optimum composition for both strain sensitivity and mechanical performance. The influence of carbon fibre positioning in BCR cross-section on the strain sensing behaviour was also investigated. Strain sensing property of BCRs was characterized by measuring the change in electrical resistance with flexural strain. It was observed that BCRs exhibited increase (positive response) or decrease (negative response) in electrical resistance depending on carbon fibre positioning. The BCR with lowest amount of carbon fibre was found to give the best strain sensitivity as well as the highest tensile strength and breaking extension. The developed BCRs showed reversible strain sensing behaviour under cyclic flexural loading with a maximum gauge factor of 23.4 at very low strain level (0.55%). Concrete beams reinforced with the optimum BCR (23/77) also exhibited strain sensing under cyclic flexural strain, although the piezoresistive behaviour in this case was irreversible.

Development of Hybrid Braided Composite Rods for Reinforcement and Health Monitoring of Structures

PubMed Central

Zdraveva, Emilija; Pereira, Cristiana; Fangueiro, Raul; Correia, A. Gomes

2014-01-01

In the present study, core-reinforced braided composite rods (BCRs) were developed and characterized for strain sensing capability. A mixture of carbon and glass fibre was used in the core, which was surrounded by a braided cover of polyester fibres. Three compositions of core with different carbon fibre/glass fibre weight ratios (23/77, 47/53, and 100/0) were studied to find out the optimum composition for both strain sensitivity and mechanical performance. The influence of carbon fibre positioning in BCR cross-section on the strain sensing behaviour was also investigated. Strain sensing property of BCRs was characterized by measuring the change in electrical resistance with flexural strain. It was observed that BCRs exhibited increase (positive response) or decrease (negative response) in electrical resistance depending on carbon fibre positioning. The BCR with lowest amount of carbon fibre was found to give the best strain sensitivity as well as the highest tensile strength and breaking extension. The developed BCRs showed reversible strain sensing behaviour under cyclic flexural loading with a maximum gauge factor of 23.4 at very low strain level (0.55%). Concrete beams reinforced with the optimum BCR (23/77) also exhibited strain sensing under cyclic flexural strain, although the piezoresistive behaviour in this case was irreversible. PMID:24574867

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.

Integrated head package for top mounted nuclear instrumentation

DOEpatents

Malandra, Louis J.; Hornak, Leonard P.; Meuschke, Robert E.

1993-01-01

A nuclear reactor such as a pressurized water reactor has an integrated head package providing structural support and increasing shielding leading toward the vessel head. A reactor vessel head engages the reactor vessel, and a control rod guide mechanism over the vessel head raises and lowers control rods in certain of the thimble tubes, traversing penetrations in the reactor vessel head, and being coupled to the control rods. An instrumentation tube structure includes instrumentation tubes with sensors movable into certain thimble tubes disposed in the fuel assemblies. Couplings for the sensors also traverse penetrations in the reactor vessel head. A shroud is attached over the reactor vessel head and encloses the control rod guide mechanism and at least a portion of the instrumentation tubes when retracted. The shroud forms a structural element of sufficient strength to support the vessel head, the control rod guide mechanism and the instrumentation tube structure, and includes radiation shielding material for limiting passage of radiation from retracted instrumentation tubes. The shroud is thicker at the bottom adjacent the vessel head, where the more irradiated lower ends of retracted sensors reside. The vessel head, shroud and contents thus can be removed from the reactor as a unit and rested safely and securely on a support.

Boiling water reactor radiation shielded Control Rod Drive Housing Supports

DOE Office of Scientific and Technical Information (OSTI.GOV)

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 nuclearmore » 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.« less

The influence of the membrane-polymer interface on colloidal membrane dynamics and phase behavior

NASA Astrophysics Data System (ADS)

Zakhary, Mark J.

A primary challenge in the field of self-assembly is to identify simple interactions that produce well-defined, complex, and controllable materials. A large part of this task is to creatively engineer appropriate assembly components with such suitable interactions built-in. Here, we demonstrate that rod-like subunits, experimentally modeled by fd bacteriophage viruses, with simple and predictable hard-core repulsive interactions, exhibit a great wealth of fascinating self-assembly behavior. These rods form two-dimensional liquid crystalline colloidal membranes consisting of monolayers of aligned particles owing purely to entropic considerations. Due to surface tension, rods near the edge of the monolayers twist, resulting in an elastic nematic ring surrounding the fluid-like membrane interior, and it is the rich phenomena rooted in the interplay between the edge and the interior that is the subject of this thesis. The chiral nature of the fd subunits causes a symmetry breaking at the membrane edge, which leads to chiral control of interfacial tension and resultantly a controllable, reversible morphological transition between membranes and one-dimensional twisted ribbons. Using optical microscopic and optical tweezer techniques, we show that a nucleation barrier exists in association with the membrane-ribbon transition, and investigate this barrier using fluctuation analysis as well as highly controlled force-extension experiments. The finite bending rigidity of the membrane edge is studied, and we show that long filamentous polymers spontaneously adhere to the edge, introducing the concept of geometrical edge-active agents. By analyzing the suppressed edge fluctuations of filament-bound membranes, it is found that the edge bending rigidity varies by up to an order of magnitude in a predictable and controllable way. Finally, we study the effect of the monolayer edge on the membrane coalescence, and observe two types of stable liquid crystalline defects that form at the coalescence site due to chiral incompatibility and frustration. By observing the fluctuations of these structures under various sample conditions, we quantify physical parameters associated with the defects, as well as their respective regions of stability. Optical tweezers are used to easily effect controllable membrane self-coalescence, which allows for imprinting defect networks, transforming between defect types, and imparting irreversible topological alterations to defects.

Pleiotropic function of DLX3 in amelogenesis: from regulating pH and keratin expression to controlling enamel rod decussation.

PubMed

Duverger, Olivier; Morasso, Maria I

2018-12-01

DLX3 is essential for tooth enamel development and is so far the only transcription factor known to be mutated in a syndromic form of amelogenesis imperfecta. Through conditional deletion of Dlx3 in the dental epithelium in mouse, we have previously established the involvement of DLX3 in enamel pH regulation, as well as in controlling the expression of sets of keratins that contribute to enamel rod sheath formation. Here, we show that the decussation pattern of enamel rods was lost in conditional knockout animals, suggesting that DLX3 controls the coordinated migration of ameloblasts during enamel secretion. We further demonstrate that DLX3 regulates the expression of some components of myosin II complexes potentially involved in driving the movement of ameloblasts that leads to enamel rod decussation.

Facile synthesis of Fe3O4/C composites for broadband microwave absorption properties

NASA Astrophysics Data System (ADS)

Liu, Xiang; Ma, Yating; Zhang, Qinfu; Zheng, Zhiming; Wang, Lai-Sen; Peng, Dong-Liang

2018-07-01

Rod-like and flower-like Fe3O4/C composites were successfully synthesized via a facile approach in aqueous phase. The morphologies, structures and static magnetic properties of as-prepared rod-like and flower-like Fe3O4/C composites were characterized thoroughly. The relative complex permittivity and permeability of Fe3O4/C/paraffin composites were recorded by a vector network analyzer (VNA) in the range of 1-18 GHz. The resonant-antiresonant electromagnetic behavior was observed simultaneously in both rod-like and flower-like Fe3O4/C composites. Moreover, the resonant-antiresonant behavior was explained using displacement current lag at the "core/shell" interface. The flower-like Fe3O4/C/paraffin composites show superior microwave absorption performance with minimum reflection loss (RL) of up to -18.73 dB at 15.37 GHz. Comparatively, the rod-like Fe3O4/C/paraffin composites have uncommon continuous trinal absorption peaks at a thickness of 2.5 mm that effectively broadens the absorption bandwidth which is from 8.0 to 13.4 GHz. Furthermore, the microwave absorption mechanism has been discussed to provide a novel design for microwave absorption materials.

The Treatment Effect of Porous Titanium Alloy Rod on the Early Stage Talar Osteonecrosis of Sheep

PubMed Central

Zhang, Yong-Quan; Zhang, Zhi-Yong; Guo, Zheng

2013-01-01

Osteonecrosis of the talus (ONT) may severely affect the function of the ankle joint. Most orthopedists believe that ONT should be treated at an early stage, but a concise and effective surgical treatment is lacking. In this study, porous titanium alloy rods were prepared and implanted into the tali of sheep with early-stage ONT (IM group). The curative effect of the rods was compared to treatment by core decompression (DC group). No significant differences in bone reconstruction were observed between the two groups at 1 month after intervention. After 3 months, the macroscopic view of gross specimens of the IM group showed ordinary contours, but the specimens of the DC group showed obvious partial bone defects and cartilage degeneration. Quantitative analysis of the reconstructed trabeculae by micro-CT and histological study suggested that the curative effect of the IM group was superior to that of the DC group at 3 months after intervention. These favorable short-term results of the implantation of porous titanium alloy rods into the tali of sheep with early-stage ONT may provide insight into an innovative surgical treatment for ONT. PMID:23516485

How Well Does Dual-Energy Computed Tomography With Metal Artifact Reduction Software Improve Image Quality and Quantify Computed Tomography Number and Iodine Concentration?

PubMed

Ohira, Shingo; Kanayama, Naoyuki; Wada, Kentaro; Karino, Tsukasa; Nitta, Yuya; Ueda, Yoshihiro; Miyazaki, Masayoshi; Koizumi, Masahiko; Teshima, Teruki

2018-04-02

The objective of this study was to assess the accuracy of the quantitative measurements obtained using dual-energy computed tomography with metal artifact reduction software (MARS). Dual-energy computed tomography scans (fast kV-switching) are performed on a phantom, by varying the number of metal rods (Ti and Pb) and reference iodine materials. Objective and subjective image analyses are performed on retroreconstructed virtual monochromatic images (VMIs) (VMI at 70 keV). The maximum artifact indices for VMI-Ti and VMI-Pb (5 metal rods) with MARS (without MARS) were 17.4 (166.7) and 34.6 (810.6), respectively; MARS significantly improved the mean subjective 5-point score (P < 0.05). The maximum differences between the measured Hounsfield unit and theoretical values for 5 mg/mL iodine and 2-mm core rods were -42.2% and -68.5%, for VMI-Ti and VMI-Pb (5 metal rods), respectively, and the corresponding differences in the iodine concentration were -64.7% and -73.0%, respectively. Metal artifact reduction software improved the objective and subjective image quality; however, the quantitative values were underestimated.

Coupling procedure for TRANSURANUS and KTF codes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jimenez, J.; Alglave, S.; Avramova, M.

2012-07-01

The nuclear industry aims to ensure safe and economic operation of each single fuel rod introduced in the reactor core. This goal is even more challenging nowadays due to the current strategy of going for higher burn-up (fuel cycles of 18 or 24 months) and longer residence time. In order to achieve that goal, fuel modeling is the key to predict the fuel rod behavior and lifetime under thermal and pressure loads, corrosion and irradiation. In this context, fuel performance codes, such as TRANSURANUS, are used to improve the fuel rod design. The modeling capabilities of the above mentioned toolsmore » can be significantly improved if they are coupled with a thermal-hydraulic code in order to have a better description of the flow conditions within the rod bundle. For LWR applications, a good representation of the two phase flow within the fuel assembly is necessary in order to have a best estimate calculation of the heat transfer inside the bundle. In this paper we present the coupling methodology of TRANSURANUS with KTF (Karlsruhe Two phase Flow subchannel code) as well as selected results of the coupling proof of principle. (authors)« less

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)

Genetic algorithm based active vibration control for a moving flexible smart beam driven by a pneumatic rod cylinder

NASA Astrophysics Data System (ADS)

Qiu, Zhi-cheng; Shi, Ming-li; Wang, Bin; Xie, Zhuo-wei

2012-05-01

A rod cylinder based pneumatic driving scheme is proposed to suppress the vibration of a flexible smart beam. Pulse code modulation (PCM) method is employed to control the motion of the cylinder's piston rod for simultaneous positioning and vibration suppression. Firstly, the system dynamics model is derived using Hamilton principle. Its standard state-space representation is obtained for characteristic analysis, controller design, and simulation. Secondly, a genetic algorithm (GA) is applied to optimize and tune the control gain parameters adaptively based on the specific performance index. Numerical simulations are performed on the pneumatic driving elastic beam system, using the established model and controller with tuned gains by GA optimization process. Finally, an experimental setup for the flexible beam driven by a pneumatic rod cylinder is constructed. Experiments for suppressing vibrations of the flexible beam are conducted. Theoretical analysis, numerical simulation and experimental results demonstrate that the proposed pneumatic drive scheme and the adopted control algorithms are feasible. The large amplitude vibration of the first bending mode can be suppressed effectively.

Thermoreversible Gels Composed of Colloidal Silica Rods with Short-Range Attractions

DOE PAGES

Murphy, Ryan P.; Hong, Kunlun; Wagner, Norman J.

2016-07-28

Dynamic arrest transitions of colloidal suspensions containing non-spherical particles are of interest for the design and processing of various particle technologies. To better understand the effects of particle shape anisotropy and attraction strength on gel and glass formation, we present a colloidal model system of octadecyl-coated silica rods, termed as adhesive hard rods (AHR), which enables control of rod aspect ratio and temperature-dependent interactions. The aspect ratios of silica rods were controlled by varying the initial TEOS concentration following the work of Kuijk et al. (J. Am. Chem. Soc., 2011, 133, 2346–2349) and temperature-dependent attractions were introduced by coating themore » calcined silica rods with an octadecyl-brush and suspending in tetradecane. The rod length and aspect ratio were found to increase with TEOS concentration as expected, while other properties such as the rod diameter, coating coverage, density, and surface roughness were nearly independent of the aspect ratio. Ultra-small angle X-ray scattering measurements revealed temperature-dependent attractions between octadecyl-coated silica rods in tetradecane, as characterized by a low-q upturn in the scattered intensity upon thermal quenching. Lastly, the rheology of a concentrated AHR suspension in tetradecane demonstrated thermoreversible gelation behavior, displaying a nearly 5 orders of magnitude change in the dynamic moduli as the temperature was cycled between 15 and 40 °C. We find the adhesive hard rod model system serves as a tunable platform to explore the combined influence of particle shape anisotropy and attraction strength on the dynamic arrest transitions in colloidal suspensions with thermoreversible, short-range attractions.« less

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.

Controlled self-assembly of conjugated rod-coil block copolymers for applications in organic optoelectronics

NASA Astrophysics Data System (ADS)

Tao, Yuefei

Organic electronics are of great interest in manufacturing light weight, mechanical flexible, and inexpensive large area devices. While significant improvements have been made over the last several years and it is now clear that morphology on the lengthscale of exciton diffusion (10nm) is of crucial importance, a clear relationship between structure and device properties has not emerged. This lack of understanding largely emerges from an inability to control morphology on this lengthscale. This thesis will center around an approach, based on block copolymer self-assembly, to generate equilibrium nanostructures on the 10 nm lengthscale of exciton diffusion and study their effects on device performance. Self-assembly of semiconducting block copolymers is complicated by the non-classical chain shape of conjugated polymers. Unlike classical polymers, the chains do not assume a Gaussian coil shape which is stretched near block copolymer interfaces, instead the chains are elongated and liquid crystalline. Previous work has demonstrated how these new molecular interactions and shapes control the phase diagram of so-called rod-coil block copolymers. Here, we will focus on controlling domain size, orientation, and chemical structure. While domain size can be controlled directly through molecular weight, this requires significant additional synthesis of domain size is to be varied. Here, the domain size is controlled by blending homopolymers into a self-assembling rod-coil block copolymer. When coil-like blocks are incorporated, the domains swell, as expected. When rod-like blocks are incorporated, they interdigitate with the rods of the block copolymers. This results in an increase in interfacial area which forces the coils to rearrange and an overall decrease in domain size with increasing rod content. Control over lamellar orientation is crucial in order to design and control charge transport pathways and exciton recombination or separation interfaces. While numerous techniques have been demonstrated for classical block copolymers, the pi conjugation in the rod blocks allow for additional control mechanisms. Liquid crystals are traditionally aligned in magnetic fields. Here, it is demonstrated that if the rod-like blocks are aligned unidirectionally, the block copolymer interfaces follow to create macroscopic alignment of the nanostructures. Organic Light Emitting Diodes (OLEDs) are generally composed of electron transporting and hole transporting moieties to balance charge recombination. Here, a new multifunctional bipolar rod-coil block copolymer containing the hole transporting and electron transporting materials is synthesized. Self-assembly of this new block copolymer results in 15nm lamellae oriented in grains both parallel and perpendicula to the anode. The self-assembled block copolymer shows superior device performance to controls consisting of a luminescent, analogous homopolymer, and a blend of the two component homopolymers. The effects of the morphologies and chemical structure on photovoltaics is explored with a rod-coil block copolymer, (poly(3-hexylthiophene-b-acrylic perylene)). By varying the kinetics of self-assembly through processing, the block copolymer can be disordered, ordered with only short range registry between the nanodomains, or with long-range order. The short range ordered samples showed the best device performance suggesting that the connectivity that is a biproduct of poor order is beneficial for device performance.

Correlated cone noise decreases rod signal contributions to the post-receptoral pathways.

PubMed

Hathibelagal, Amithavikram R; Feigl, Beatrix; Zele, Andrew J

2018-04-01

This study investigated how invisible extrinsic temporal white noise that correlates with the activity of one of the three [magnocellular (MC), parvocellular (PC), or koniocellular (KC)] post-receptoral pathways alters mesopic rod signaling. A four-primary photostimulator provided independent control of the rod and three cone photoreceptor excitations. The rod contributions to the three post-receptoral pathways were estimated by perceptually matching a 20% contrast rod pulse by independently varying the LMS (MC pathway), +L-M (PC pathway), and S-cone (KC pathway) excitations. We show that extrinsic cone noise caused a predominant decrease in the overall magnitude and ratio of the rod contributions to each pathway. Thus, the relative cone activity in the post-receptoral pathways determines the relative mesopic rod inputs to each pathway.

RECOVERY OF ROD PHOTORESPONSES IN ABCR-DEFICIENT MICE

PubMed Central

Pawar, Ambarish S.; Qtaishat, Nasser M.; Little, Deborah M.; Pepperberg, David R.

2010-01-01

Purpose ABCR protein in the rod outer segment is thought to facilitate movement of the all-trans retinal photoproduct of rhodopsin bleaching out of the disk lumen. We investigated the extent to which ABCR deficiency affects post-bleach recovery of the rod photoresponse in ABCR-deficient (abcr−/−) mice. Methods Electroretinographic (ERG) a-wave responses were recorded from abcr−/− mice and two control strains. Using a bright probe flash, we examined the course of rod recovery following fractional rhodopsin bleaches of ~10−6, ~3×10−5, ~0.03 and ~0.30–0.40. Results Dark-adapted abcr−/− mice and controls exhibited similar normalized near-peak amplitudes of the paired-flash-ERG-derived, weak-flash response. Response recovery following ~10−6 bleaching exhibited an average exponential time constant of 319, 171 and 213 ms, respectively, in the abcr−/− and the two control strains. Recovery time constants determined for ~3×10−5 bleaching did not differ significantly among strains. However, those determined for the ~0.03 bleach indicated significantly faster recovery in abcr−/− (2.34 ± 0.74 min) than in the controls (5.36 ± 2.20 min, and 5.92 ± 2.44 min). Following ~0.30–0.40 bleaching, the initial recovery in the abcr−/− was on average faster than in controls. Conclusions By comparison with controls, abcr−/− mice exhibit faster rod recovery following a bleach of ~0.03. The data suggest that ABCR in normal rods may directly or indirectly prolong all-trans retinal clearance from the disk lumen over a significant bleaching range, and that the essential function of ABCR may be to promote the clearance of residual amounts of all-trans retinal that remain in the disks long after bleaching. PMID:18263807

Surface properties of sprayed and electrodeposited ZnO rod layers

NASA Astrophysics Data System (ADS)

Gromyko, I.; Krunks, M.; Dedova, T.; Katerski, A.; Klauson, D.; Oja Acik, I.

2017-05-01

Herein we present a comparative study on as-deposited, two-month-stored, and heat-treated ZnO rods obtained by spray pyrolysis (SP) at 550 °C, and electrodeposition (ED) at 80 °C. The aim of the study is to establish the reason for different behaviour of wettability and photocatalytic activity (PA) of SP and ED rods. Samples were studied using XPS, SEM, XRD, Raman, contact angle (CA) measurements and photocatalytic oxidation of doxycycline. Wettability and PA are mainly controlled by surface composition rather than by morphology. The relative amount of hydroxyl groups on the surface of as-deposited ED rods is four times higher compared to as-deposited SP rods. Opposite to SP rods, ED rods contain oxygen vacancy defects (Vo). Therefore, as-deposited ED rods are superhydrophilic (CA ∼ 3°) and show highest PA among studied samples, being three times higher compared to SP rods (removing of 75% of doxycycline after 30 min). It was revealed that as-deposited ED rods are inclined to faster contamination. The amount of Cdbnd C groups on the surface of aged ED rods is six times higher compared to aged SP rods. Stored ED samples become hydrophobic (CA ∼ 120°) and PA decreases sharply while SP rods remain hydrophilic (CA ∼ 50°), being more resistive to the contamination.

Ho3+-doped AlF3-TeO2-based glass fibers for 2.1 µm laser applications

NASA Astrophysics Data System (ADS)

Wang, S. B.; Jia, Z. X.; Yao, C. F.; Ohishi, Y.; Qin, G. S.; Qin, W. P.

2017-05-01

Ho3+-doped AlF3-TeO2-based glass fibers based on AlF3-BaF2-CaF2-YF3-SrF2-MgF2-TeO2 glasses are fabricated by using a rod-in-tube method. The glass rod including a core and a thick cladding layer is prepared by using a suction method, where the thick cladding layer is used to protect the core from the effect of surface crystallization during the fiber drawing. By inserting the glass rod into a glass tube, the glass fibers with relatively low loss (~2.3 dB m-1 @ 1560 nm) are prepared. By using a 38 cm long Ho3+-doped AlF3-TeO2-based glass fiber as the gain medium and a 1965 nm fiber laser as the pump source, 2065 nm lasing is obtained for a threshold pump power of ~220 mW. With further increasing the pump power to ~325 mW, the unsaturated output power of the 2065 nm laser is about 82 mW and the corresponding slope efficiency is up to 68.8%. The effects of the gain fiber length on the lasing threshold, the slope efficiency, and the operating wavelength are also investigated. Our experimental results show that Ho3+-doped AlF3-TeO2-based glass fibers are promising gain media for 2.1 µm laser applications.

Single-particle studies of band alignment effects on electron transfer dynamics from semiconductor hetero-nanostructures to single-walled carbon nanotubes.

PubMed

Yuan, Chi-Tsu; Wang, Yong-Gang; Huang, Kuo-Yen; Chen, Ting-Yu; Yu, Pyng; Tang, Jau; Sitt, Amit; Banin, Uri; Millo, Oded

2012-01-24

We utilize single-molecule spectroscopy combined with time-correlated single-photon counting to probe the electron transfer (ET) rates from various types of semiconductor hetero-nanocrystals, having either type-I or type-II band alignment, to single-walled carbon nanotubes. A significantly larger ET rate was observed for type-II ZnSe/CdS dot-in-rod nanostructures as compared to type-I spherical CdSe/ZnS core/shell quantum dots and to CdSe/CdS dot-in-rod structures. Furthermore, such rapid ET dynamics can compete with both Auger and radiative recombination processes, with significance for effective photovoltaic operation. © 2011 American Chemical Society

Application of ATHLET/DYN3D coupled codes system for fast liquid metal cooled reactor steady state simulation

NASA Astrophysics Data System (ADS)

Ivanov, V.; Samokhin, A.; Danicheva, I.; Khrennikov, N.; Bouscuet, J.; Velkov, K.; Pasichnyk, I.

2017-01-01

In this paper the approaches used for developing of the BN-800 reactor test model and for validation of coupled neutron-physic and thermohydraulic calculations are described. Coupled codes ATHLET 3.0 (code for thermohydraulic calculations of reactor transients) and DYN3D (3-dimensional code of neutron kinetics) are used for calculations. The main calculation results of reactor steady state condition are provided. 3-D model used for neutron calculations was developed for start reactor BN-800 load. The homogeneous approach is used for description of reactor assemblies. Along with main simplifications, the main reactor BN-800 core zones are described (LEZ, MEZ, HEZ, MOX, blankets). The 3D neutron physics calculations were provided with 28-group library, which is based on estimated nuclear data ENDF/B-7.0. Neutron SCALE code was used for preparation of group constants. Nodalization hydraulic model has boundary conditions by coolant mass-flow rate for core inlet part, by pressure and enthalpy for core outlet part, which can be chosen depending on reactor state. Core inlet and outlet temperatures were chosen according to reactor nominal state. The coolant mass flow rate profiling through the core is based on reactor power distribution. The test thermohydraulic calculations made with using of developed model showed acceptable results in coolant mass flow rate distribution through the reactor core and in axial temperature and pressure distribution. The developed model will be upgraded in future for different transient analysis in metal-cooled fast reactors of BN type including reactivity transients (control rods withdrawal, stop of the main circulation pump, etc.).

MEANS FOR CONTROLLING REACTIONS

DOEpatents

Nordheim, L.W.; Wigner, E.P.

1961-06-27

The patented means is described for controlling a nuclear reactor which comprises a tank containing a dispersion of a thermally fissionable material in a liquid moderator and a material convertible to a thermally fissionable material in a container disposed about the tank. The control means comprises a control rod chamber, containing only a liquid moderator, disposed within the container and adjacent to the tank and a control rod designed to be inserted into the chamber.

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.

Analysis of failed nuclear plant components

NASA Astrophysics Data System (ADS)

Diercks, D. R.

1993-12-01

Argonne National Laboratory has conducted analyses of failed components from nuclear power- gener-ating stations since 1974. The considerations involved in working with and analyzing radioactive compo-nents are reviewed here, and the decontamination of these components is discussed. Analyses of four failed components from nuclear plants are then described to illustrate the kinds of failures seen in serv-ice. The failures discussed are (1) intergranular stress- corrosion cracking of core spray injection piping in a boiling water reactor, (2) failure of canopy seal welds in adapter tube assemblies in the control rod drive head of a pressurized water reactor, (3) thermal fatigue of a recirculation pump shaft in a boiling water reactor, and (4) failure of pump seal wear rings by nickel leaching in a boiling water reactor.

Pretest analysis document for Semiscale Test S-LH-1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shaw, R.A.

Results from various pretest calculations which were performed for Test S-LH-1 are included in this report. Test S-LH-1 has been designed to produce primary liquid holdup in the steam generator U-tubes similar to Tests S-UT-8. The analyses included in this report indicate liquid will be held in the tubes, the core liquid level will be appropriately depressed, and a core heater rod temperature excursion should occur. Several sensitivity studies are also included which identify parameters which could affect the response.

VIEW OF CABLES AND TAPES ASSOCIATED WITH ADRIVE CONTROL ROD ...

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

VIEW OF CABLES AND TAPES ASSOCIATED WITH A-DRIVE CONTROL ROD SYSTEM, AT LEVEL +15’, DIRECTLY ABOVE PDP CONTROL ROOM, LOOKING NORTHWEST. THE CABLES FROM THE PDP ROOM GO THROUGH THE CONCRETE WALL, MAKE A RIGHT ANGLE TURN DOWNWARD, AND DESCEND INTO THE PDP CONTROL ROOM AS VERTICAL TAPES - Physics Assembly Laboratory, Area A/M, Savannah River Site, Aiken, Aiken County, SC

Fuel inspection and reconstitution experience at Surry Power Station

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brookmire, T.A.

Surry Power Station, located on the James River near Williamsburg, Virginia, has two Westinghouse pressurized water reactors. Unit 2 consistently sets a high standard of fuel performance (no indication of fuel failures in recent cycles), while unit 1, since cycle 6, has been plagued with numerous fuel failures. Both Surry units operate with Westinghouse standard 15 x 15 fuel. Virginia Power management set goals to reduce the coolant activity, thus reducing person-rem exposure and the associated costs of high coolant activity. To achieve this goal, extensive fuel examination campaigns were undertaken that included high-magnification video inspectionsa, debris cleaning, wet andmore » vacuum fuel sipping, fuel rod ultrasonic testing, and eddy current examination. In the summer of 1985, during cycle 8 operation, Kraftwerk Union reconstituted (repaired) the damage, once-burned assemblies from cycles 6 and 7 by replacing failed fuel rods with solid Zircaloy-4 rods. Currently, cycle 9 has operated for 5 months without any indication of fuel failure (the cycle 9 core has two reconstituted assemblies).« less

Hybrid microfabrication of nanofiber-based sheets and rods for tissue engineering applications.

PubMed

Park, Suk-Hee; Kim, Min Sung; Lee, Dasom; Choi, Yong Whan; Kim, Deok-Ho; Suh, Kahp-Yang

2013-12-01

Electrospun nanofibers have been developed into a variety of forms for tissue engineering scaffolds to regulate the cellular functions guided by nanotopographical cues. Here, we have successfully fabricated nanofiber-based scaffold complexes of rod and sheet type by combining the three microfabrication techniques of electrospinning, spin coating, and polymer melt deposition. It was demonstrated that this hybrid fabrication could produce uniaxially aligned nanofiber scaffolds supported by a thin film, allowing for a mechanically enforced substrate for cell culture as well as facile scaffold manipulation. The results of cell analysis indicated that nanofibers on spin-coated films could provide contact guidance effects on cells and retain them even after manipulation. As an application of the cell-laden nanofiber film, we built a rod-type structure by rolling up the film around a mechanically supporting core microfiber, which was incorporated by polymer melt deposition. A biocompatible and biodegradable polymer, polycaprolactone, was used throughout the processes and thus could be used as a directly implantable substitute in tissue regeneration.

Use of Computer-Aided Tomography (CT) Imaging for Quantifying Coarse Roots, Rhizomes, Peat, and Particle Densities in Marsh Soils

EPA Science Inventory

Computer-aided Tomography (CT) imaging was utilized to quantify wet mass of coarse roots, rhizomes, and peat in cores collected from organic-rich (Jamaica Bay, NY) and mineral (North Inlet, SC) Spartina alterniflora soils. Calibration rods composed of materials with standard dens...

Fabrication of light water reactor tritium targets

DOE Office of Scientific and Technical Information (OSTI.GOV)

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 beenmore » 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.« less

G T-Mohr Start-up Reactivity Insertion Transient Analysis Using Simulink

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fard, Mehdi Reisi; Blue, Thomas E.; Miller, Don W.

2006-07-01

As a part of a Department of Energy-Nuclear Engineering Research Initiative (NERI) Project, we at OSU are investigating SiC semiconductor detectors as neutron power monitors for Generation IV power reactors. As a part of this project, we are investigating the power monitoring requirements for a specific type of Generation IV reactor, namely the GT-MHR. To evaluate the power monitoring requirements for the GT-MHR that are most demanding for a SiC diode power monitor, we have developed a Simulink model to study the transient behavior of the GT-MHR. In this paper, we describe the application of the Simulink code to themore » analysis of a series of Start-up Reactivity Insertion Transients (SURITs). The SURIT is considered to be a limiting protectable accident in terms of establishing the dynamic range of a SiC power monitor because of the low count rate of the detector during the start-up and absence of the reactivity feedback mechanism at the beginning of transient. The SURIT is studied with the ultimate goal of identifying combinations of 1) reactor power scram setpoints and 2) cram initiation times (the time in which a scram must be initiated once the setpoint is exceeded) for which the GT-MHR core is protected in the event of a continuous withdrawal of a control rod bank from the core from low powers. The SURIT is initiated by withdrawing a rod bank when the reactor is cold (300 K) and sub-critical at the BOEC (Beginning of Equilibrium Cycle) condition. Various initial power levels have been considered corresponding to various degrees of sub-criticality and various source strengths. An envelope of response is determined to establish which initial powers correspond to the worst case SURIT. (authors)« less

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.

Controlling Tensegrity Robots through Evolution using Friction based Actuation

NASA Technical Reports Server (NTRS)

Kothapalli, Tejasvi; Agogino, Adrian K.

2017-01-01

Traditional robotic structures have limitations in planetary exploration as their rigid structural joints are prone to damage in new and rough terrains. In contrast, robots based on tensegrity structures, composed of rods and tensile cables, offer a highly robust, lightweight, and energy efficient solution over traditional robots. In addition tensegrity robots can be highly configurable by rearranging their topology of rods, cables and motors. However, these highly configurable tensegrity robots pose a significant challenge for locomotion due to their complexity. This study investigates a control pattern for successful locomotion in tensegrity robots through an evolutionary algorithm. A twelve-rod hardware model is rapidly prototyped to utilize a new actuation method based on friction. A web-based physics simulation is created to model the twelve-rod tensegrity ball structure. Square-waves are used as control policies for the actuators of the tensegrity structure. Monte Carlo trials are run to find the most successful number of amplitudes for the square-wave control policy. From the results, an evolutionary algorithm is implemented to find the most optimized solution for locomotion of the twelve-rod tensegrity structure. The software pattern coupled with the new friction based actuation method can serve as the basis for highly efficient tensegrity robots in space exploration.

Solid State Welding Development at Marshall Space Flight Center

NASA Technical Reports Server (NTRS)

Ding, Robert J.; Walker, Bryant

2012-01-01

What is TSW and USW? TSW is a solid state weld process consisting of an induction coil heating source, a stir rod, and non-rotating containment plates Independent heating, stirring and forging controls Decouples the heating, stirring and forging process elements of FSW. USW is a solid state weld process consisting of an induction coil heating source, a stir rod, and a non-rotating containment plate; Ultrasonic energy integrated into non-rotating containment plate and stir rod; Independent heating, stirring and forging controls; Decouples the heating, stirring and forging process elements of FSW.

Geochemical Fingerprinting of the World Trade Center Attack in New York Harbor Sediments

NASA Astrophysics Data System (ADS)

Brabander, D. J.; Oktay, S.; Smith, J.; Kada, J.; Bullen, T.; Olsen, C.

2002-12-01

By comparing the textural, chemical, and isotopic composition of World Trade Center (WTC) ash samples (collected near Ground Zero one week after the terrorist attack) with sediment samples from cores taken on October 12, 2001 in known deposition areas in New York Harbor (NYH), we characterized a unique suite of geochemical-textural tracers that allow us to both identify and quantify the input of WTC derived material to adjacent areas in the Hudson River estuary. Scanning electron microscopy coupled with energy dispersive spectroscopy revealed two chemically distinct (Si-rich and Ca-rich) rod-like features (40-200 æm in length) in both ash and sediment samples. The Si-rich rods are consistent with a fiberglass parent material while the Ca-rich rods originate from gypsum. An 87Sr/86Sr ratio for the ash material of 0.7088 (n=2) coupled with Ca/Sr (wt. ratio) ranging from 260-300 suggest that the ash material analyzed is approximately 70% gypsum. As a function of depth within the sediment core, correlations exist between the measured activities of 7Be (a naturally occurring short-lived radionuclide), elemental weight-percent ratios of Ca/Sr, and the isotopic ratios of 87Sr/86Sr ratios. . These combined isotopic approaches allow us to constrain the timing (via 7Be), and the composition and amount (via 87Sr/86Sr and Ca/Sr) of WTC material input into the NYH sediments. These down-core isotope-ratio profiles can be described by a mixing line between background NYH 87Sr/86Sr ratios (>0.724) and the WTC derived ash material. The geochemical-textural tracers associated with the WTC terrorist attack may provide a potential tool for assessing the fate and transport of WTC material in the Lower Hudson River and aid in assessing the environmental and human health impacts of the WTC catastrophe.

Bounded parametric control of plane motions of space tethered system

NASA Astrophysics Data System (ADS)

Bezglasnyi, S. P.; Mukhametzyanova, A. A.

2018-05-01

This paper is focused on the problem of control of plane motions of a space tethered system (STS). The STS is modeled as a heavy rod with two point masses. Point masses are fixed on the rod. A third point mass can move along the rod. The control is realized as a continuous change of the distance from the centre of mass of the tethered system to the movable mass. New limited control laws processes of excitation and damping are built. Diametric reorientation and gravitational stabilization to the local vertical of an STS were obtained. The problem is solved by the method of Lyapunov's functions of the classical theory of stability. The theoretical results are confirmed by numerical calculations.

Is An Ostomy Rod Useful for Bridging the Retraction During the Creation of a Loop Ileostomy? A Randomized Control Trial.

PubMed

Uchino, Motoi; Ikeuchi, Hiroki; Bando, Toshihiro; Chohno, Teruhiro; Sasaki, Hirofumi; Horio, Yuki

2017-08-01

A loop ileostomy is generally created during restorative proctocolectomy (RPC) for treating ulcerative colitis (UC), and an ostomy rod is often used to prevent stoma retraction. However, its usefulness or harmfulness has not been proven. We performed a prospective randomized control study to investigate the non-inferiority of ostomy creation without a rod to prevent stoma retraction. Patients with UC who underwent RPC were enrolled and randomly divided into groups either with or without ostomy rod use. Incidences of stoma retraction and dermatitis were compared. Of the 320 patients in the study groups, 308 qualified for the intention-to-treat (ITT) analysis, and 257 were included in the per-protocol (PP) analysis. Ostomy retraction was recognized in 6 patients, 3 with a rod and 3 without. The difference with rod use (95% confidence interval) was 0.1 (-2.9 to 3.1)% in the PP analysis and 0.0 (-2.2 to 2.2)% in the ITT analysis. There were no significant differences in stoma retraction regardless of whether an ostomy rod was used in either analysis. Dermatitis was more common in patients with rod use (84/154) than in those without (40/154) (p < 0.01). Although median body mass indices were extremely low (20 kg/m 2 ), an ostomy rod is not routinely needed as it may increase the risk of dermatitis. However, results in obese patients may differ from those shown here, which should be clarified via further studies.

Rod electrical coupling is controlled by a circadian clock and dopamine in mouse retina

PubMed Central

Jin, Nan Ge; Chuang, Alice Z; Masson, Philippe J; Ribelayga, Christophe P

2015-01-01

Key points Rod photoreceptors play a key role in vision in dim light; in the mammalian retina, although rods are anatomically connected or coupled by gap junctions, a type of electrical synapse, the functional importance and regulation of rod coupling has remained elusive. We have developed a new technique in the mouse: perforated patch-clamp recording of rod inner segments in isolated intact retinae maintained by superfusion. We find that rod electrical coupling is controlled by a circadian clock and dopamine, and is weak during the day and stronger at night. The results also indicate that the signal-to-noise ratio for a dim light response is increased at night because of coupling. Our observations will provide a framework for understanding the daily variations in human vision as well as the basis of specific retinal malfunctions. Abstract Rod single-photon responses are critical for vision in dim light. Electrical coupling via gap junction channels shapes the light response properties of vertebrate photoreceptors, but the regulation of rod coupling and its impact on the single-photon response have remained unclear. To directly address these questions, we developed a perforated patch-clamp recording technique and recorded from single rod inner segments in isolated intact neural mouse retinae, maintained by superfusion. Experiments were conducted at different times of the day or under constant environmental conditions, at different times across the circadian cycle. We show that rod electrical coupling is regulated by a circadian clock and dopamine, so that coupling is weak during the day and strong at night. Altogether, patch-clamp recordings of single-photon responses in mouse rods, tracer coupling, receptive field measurements and pharmacological manipulations of gap junction and dopamine receptor activity provide compelling evidence that rod coupling is modulated in a circadian manner. These data are consistent with computer modelling. At night, single-photon responses are smaller due to coupling, but the signal-to-noise ratio for a dim (multiphoton) light response is increased at night because of signal averaging between coupled rods. PMID:25616058

DESIGN AND DEVELOPMENT REPORT ON TREAT CONTROL ROD DRIVE II

DOE Office of Scientific and Technical Information (OSTI.GOV)

Batch, R.V.

1961-05-01

A discussion is given of the development of TREAT control rod drive II, which describes the basic design, the problems involved with the design, the various design methods pursued, the testing procedures, and the evaluation of the performance characteristics of the final drive. (B.O.G.)

75 FR 68548 - Airworthiness Directives; Airbus Model A318, A319, A320, and A321 Series Airplanes

Federal Register 2010, 2011, 2012, 2013, 2014

2010-11-08

...: One case of elevator servo-control disconnection has been experienced on an aeroplane of the A320 family. Investigation has revealed that the failure occurred at the servo-control rod eye-end. Further to... servo-control rod eye-ends. In several cases, both actuators of the same elevator surface were affected...

Experimental Study of Two Phase Flow Behavior Past BWR Spacer Grids

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ratnayake, Ruwan K.; Hochreiter, L.E.; Ivanov, K.N.

2002-07-01

Performance of best estimate codes used in the nuclear industry can be significantly improved by reducing the empiricism embedded in their constitutive models. Spacer grids have been found to have an important impact on the maximum allowable Critical Heat Flux within the fuel assembly of a nuclear reactor core. Therefore, incorporation of suitable spacer grids models can improve the critical heat flux prediction capability of best estimate codes. Realistic modeling of entrainment behavior of spacer grids requires understanding the different mechanisms that are involved. Since visual information pertaining to the entrainment behavior of spacer grids cannot possibly be obtained frommore » operating nuclear reactors, experiments have to be designed and conducted for this specific purpose. Most of the spacer grid experiments available in literature have been designed in view of obtaining quantitative data for the purpose of developing or modifying empirical formulations for heat transfer, critical heat flux or pressure drop. Very few experiments have been designed to provide fundamental information which can be used to understand spacer grid effects and phenomena involved in two phase flow. Air-water experiments were conducted to obtain visual information on the two-phase flow behavior both upstream and downstream of Boiling Water Reactor (BWR) spacer grids. The test section was designed and constructed using prototypic dimensions such as the channel cross-section, rod diameter and other spacer grid configurations of a typical BWR fuel assembly. The test section models the flow behavior in two adjacent sub channels in the BWR core. A portion of a prototypic BWR spacer grid accounting for two adjacent channels was used with industrial mild steel rods for the purpose of representing the channel internals. Symmetry was preserved in this practice, so that the channel walls could effectively be considered as the channel boundaries. Thin films were established on the rod surfaces by injecting water through a set of perforations at the bottom ends of the rods, ensuring that the flow upstream of the bottom-most spacer grid is predominantly annular. The flow conditions were regulated such that they represent typical BWR operating conditions. Photographs taken during experiments show that the film entrainment increases significantly at the spacer grids, since the points of contact between the rods and the grids result in a peeling off of large portions of the liquid film from the rod surfaces. Decreasing the water flow resulted in eventual drying out, beginning at positions immediately upstream of the spacer grids. (authors)« less

Thermodynamic modelling of the C-U and B-U binary systems

NASA Astrophysics Data System (ADS)

Chevalier, P. Y.; Fischer, E.

2001-02-01

The thermodynamic modelling of the carbon-uranium (C-U) and boron-uranium (B-U) binary systems is being performed in the framework of the development of a thermodynamic database for nuclear materials, for increasing the basic knowledge of key phenomena which may occur in the event of a severe accident in a nuclear power plant. Applications are foreseen in the nuclear safety field to the physico-chemical interaction modelling, on the one hand the in-vessel core degradation producing the corium (fuel, zircaloy, steel, control rods) and on the other hand the ex-vessel molten corium-concrete interaction (MCCI). The key O-U-Zr ternary system, previously modelled, allows us to describe the first interaction of the fuel with zircaloy cladding. Then, the three binary systems Fe-U, Cr-U and Ni-U were modelled as a preliminary work for modelling the O-U-Zr-Fe-Cr-Ni multicomponent system, allowing us to introduce the steel components in the corium. In the existing database (TDBCR, thermodynamic data base for corium), Ag and In were introduced for modelling AIC (silver-indium-cadmium) control rods which are used in French pressurized water reactors (PWR). Elsewhere, B 4C is also used for control rods. That is why it was agreed to extend in the next years the database with two new components, B and C. Such a work needs the thermodynamic modelling of all the binary and pseudo-binary sub-systems resulting from the combination of B, B 2O 3 and C with the major components of TDBCR, O-U-Zr-Fe-Cr-Ni-Ag-In-Ba-La-Ru-Sr-Al-Ca-Mg-Si + Ar-H. The critical assessment of the very numerous experimental information available for the C-U and B-U binary systems was performed by using a classical optimization procedure and the Scientific Group Thermodata Europe (SGTE). New optimized Gibbs energy parameters are given, and comparisons between calculated and experimental equilibrium phase diagrams or thermodynamic properties are presented. The self-consistency obtained is quite satisfactory.

Combustion of solid carbon rods in zero and normal gravity. Ph.D. Thesis - Toledo Univ., Ohio

NASA Technical Reports Server (NTRS)

Spuckler, C. M.

1981-01-01

In order to investigate the mechanism of carbon combustion and to assess the importance of gravitational induced convection on the process, zero and normal gravity experiments were conducted in which spectroscopic carbon rods were resistance ignitied and burned in dry oxygen environments. In the zero-gravity drop tower tests, a blue flame surrounded the rod, showing that a gas phase reaction in which carbon monoxide was oxidized to carbon dioxide was taking place. The ratio of flame diameter to rod diameter was obtained as a function of time. It was found that this ratio was inversely proportional to both the oxygen pressure and the rod diameter. In the normal gravity tests, direct mass spectrometric sampling was used to measure gas phase concentrations. The gas sampling probe was positioned near the circumference of a horizontally mounted 0.615 cm diameter carbon rod, either at the top or at angles of 45 deg to 90 deg from the top, and yielded concentration profiles of CO2, CO, and O2 as a function of distance from the surface. The mechanism controlling the combustion process was found to change from chemical process control at the 90 deg and 45 deg probe positions to mass transfer control at the 0 deg probe position at the top of the rod. Under the experimental conditions used, carbon combustion was characterized by two surface reactions, 2C + O2 yields 2CO and CO2 + C yields 2CO, and a gas phase reaction, 2CO + O2 yields 2CO2.

Structure-dependent performance of TiO 2/C as anode material for Na-ion batteries

DOE Office of Scientific and Technical Information (OSTI.GOV)

He, Hanna; Gan, Qingmeng; Wang, Haiyan

The performance of energy storage materials is highly dependent on their nanostructures. Herein, hierarchical rod-in-tube TiO 2 with a uniform carbon coating is synthesized as the anode material for sodium-ion batteries by a facile solvothermal method. This unique structure consists of a tunable nanorod core, interstitial hollow spaces, and a functional nanotube shell assembled from two-dimensional nanosheets. By adjusting the types of solvents used and reaction time, the morphologies of TiO 2/C composites can be tuned to nanoparticles, microrods, rod-in-tube structures, or microtubes. Among these materials, rod-in-tube TiO 2 with a uniform carbon coating shows the highest electronic conductivity, specificmore » surface area (336.4 m(2) g(-1)), and porosity, and these factors lead to the best sodium storage capability. Benefiting from the unique structural features and improved electronic/ionic conductivity, the as-obtained rod-in-tube TiO2/C in coin cell tests exhibits a high discharge capacity of 277.5 and 153.9 mAh g(-1) at 50 and 5000 mA g(-1), respectively, and almost 100% capacity retention over 14,000 cycles at 5000 mA g(-1). In operando high-energy X-ray diffraction further confirms the stable crystal structure of the rod-in-tube TiO 2/C during Na+ insertion/extraction. This work highlights that nanostructure design is an effective strategy to achieve advanced energy storage materials.« less

Development of an ATCA IPMI controller mezzanine board to be used in the ATCA developments for the ATLAS Liquid Argon upgrade

NASA Astrophysics Data System (ADS)

Dumont Dayot, Nicolas

2012-01-01

In the context of the LHC upgrade, we develop a new Read Out Driver (ROD) for the ATLAS Liquid Argon (LAr) community. ATCA and μTCA (Advanced/Micro Telecom Computing Architecture) is becoming a standard in high energy physics and a strong candidate to be used for boards and crates. We work to master ATCA and to integrate a large number of high speed links (96 links at 8.5 Gbps) on a ROD evaluation ATCA board. A versatile ATCA IPMI controller for ATCA boards which is FPGA Mezzanine Card (FMC) compliant has been developed to control the ROD evaluation board.

The biomechanical consequences of rod reduction on pedicle screws: should it be avoided?

PubMed

Paik, Haines; Kang, Daniel G; Lehman, Ronald A; Gaume, Rachel E; Ambati, Divya V; Dmitriev, Anton E

2013-11-01

Rod contouring is frequently required to allow for appropriate alignment of pedicle screw-rod constructs. When residual mismatch is still present, a rod persuasion device is often used to achieve further rod reduction. Despite its popularity and widespread use, the biomechanical consequences of this technique have not been evaluated. To evaluate the biomechanical fixation strength of pedicle screws after attempted reduction of a rod-pedicle screw mismatch using a rod persuasion device. Fifteen 3-level, human cadaveric thoracic specimens were prepared and scanned for bone mineral density. Osteoporotic (n=6) and normal (n=9) specimens were instrumented with 5.0-mm-diameter pedicle screws; for each pair of comparison level tested, the bilateral screws were equal in length, and the screw length was determined by the thoracic level and size of the vertebra (35 to 45 mm). Titanium 5.5-mm rods were contoured and secured to the pedicle screws at the proximal and distal levels. For the middle segment, the rod on the right side was intentionally contoured to create a 5-mm residual gap between the inner bushing of the pedicle screw and the rod. A rod persuasion device was then used to engage the setscrew. The left side served as a control with perfect screw/rod alignment. After 30 minutes, constructs were disassembled and vertebrae individually potted. The implants were pulled in-line with the screw axis with peak pullout strength (POS) measured in Newton (N). For the proximal and distal segments, pedicle screws on the right side were taken out and reinserted through the same trajectory to simulate screw depth adjustment as an alternative to rod reduction. Pedicle screws reduced to the rod generated a 48% lower mean POS (495±379 N) relative to the controls (954±237 N) (p<.05) and significantly decreased work energy to failure (p<.05). Nearly half (n=7) of the pedicle screws had failed during the reduction attempt with visible pullout of the screw. After reduction, decreased POS was observed in both normal (p<.05) and osteoporotic (p<.05) bone. Back out and reinsertion of the screw resulted in no significant difference in mean POS, stiffness, and work energy to failure (p>.05). In circumstances where a rod is not fully seated within the pedicle screw, the use of a rod persuasion device decreases the overall POS and work energy to failure of the screw or results in outright failure. Further rod contouring or correction of pedicle screw depth of insertion may be warranted to allow for appropriate alignment of the longitudinal rods. Published by Elsevier Inc.

Probing Bioluminescence Resonance Energy Transfer in Quantum Rod-Luciferase Nanoconjugates.

PubMed

Alam, Rabeka; Karam, Liliana M; Doane, Tennyson L; Coopersmith, Kaitlin; Fontaine, Danielle M; Branchini, Bruce R; Maye, Mathew M

2016-02-23

We describe the necessary design criteria to create highly efficient energy transfer conjugates containing luciferase enzymes derived from Photinus pyralis (Ppy) and semiconductor quantum rods (QRs) with rod-in-rod (r/r) microstructure. By fine-tuning the synthetic conditions, CdSe/CdS r/r-QRs were prepared with two different emission colors and three different aspect ratios (l/w) each. These were hybridized with blue, green, and red emitting Ppy, leading to a number of new BRET nanoconjugates. Measurements of the emission BRET ratio (BR) indicate that the resulting energy transfer is highly dependent on QR energy accepting properties, which include absorption, quantum yield, and optical anisotropy, as well as its morphological and topological properties, such as aspect ratio and defect concentration. The highest BR was found using r/r-QRs with lower l/w that were conjugated with red Ppy, which may be activating one of the anisotropic CdSe core energy levels. The role QR surface defects play on Ppy binding, and energy transfer was studied by growth of gold nanoparticles at the defects, which indicated that each QR set has different sites. The Ppy binding at those sites is suggested by the observed BRET red-shift as a function of Ppy-to-QR loading (L), where the lowest L results in highest efficiency and furthest shift.

Mutations in cell elongation genes mreB, mrdA and mrdB suppress the shape defect of RodZ-deficient cells.

PubMed

Shiomi, Daisuke; Toyoda, Atsushi; Aizu, Tomoyuki; Ejima, Fumio; Fujiyama, Asao; Shini, Tadasu; Kohara, Yuji; Niki, Hironori

2013-03-01

RodZ interacts with MreB and both factors are required to maintain the rod shape of Escherichia coli. The assembly of MreB into filaments regulates the subcellular arrangement of a group of enzymes that synthesizes the peptidoglycan (PG) layer. However, it is still unknown how polymerization of MreB determines the rod shape of bacterial cells. Regulatory factor(s) are likely to be involved in controlling the function and dynamics of MreB. We isolated suppressor mutations to partially recover the rod shape in rodZ deletion mutants and found that some of the suppressor mutations occurred in mreB. All of the mreB mutations were in or in the vicinity of domain IA of MreB. Those mreB mutations changed the property of MreB filaments in vivo. In addition, suppressor mutations were found in the periplasmic regions in PBP2 and RodA, encoded by mrdA and mrdB genes. Similar to MreB and RodZ, PBP2 and RodA are pivotal to the cell wall elongation process. Thus, we found that mutations in domain IA of MreB and in the periplasmic domain of PBP2 and RodA can restore growth and rod shape to ΔrodZ cells, possibly by changing the requirements of MreB in the process. © 2013 Blackwell Publishing Ltd.

Mutations in cell elongation genes mreB, mrdA and mrdB suppress the shape defect of RodZ-deficient cells

PubMed Central

Shiomi, Daisuke; Toyoda, Atsushi; Aizu, Tomoyuki; Ejima, Fumio; Fujiyama, Asao; Shini, Tadasu; Kohara, Yuji; Niki, Hironori

2013-01-01

RodZ interacts with MreB and both factors are required to maintain the rod shape of Escherichia coli. The assembly of MreB into filaments regulates the subcellular arrangement of a group of enzymes that synthesizes the peptidoglycan (PG) layer. However, it is still unknown how polymerization of MreB determines the rod shape of bacterial cells. Regulatory factor(s) are likely to be involved in controlling the function and dynamics of MreB. We isolated suppressor mutations to partially recover the rod shape in rodZ deletion mutants and found that some of the suppressor mutations occurred in mreB. All of the mreB mutations were in or in the vicinity of domain IA of MreB. Those mreB mutations changed the property of MreB filaments in vivo. In addition, suppressor mutations were found in the periplasmic regions in PBP2 and RodA, encoded by mrdA and mrdB genes. Similar to MreB and RodZ, PBP2 and RodA are pivotal to the cell wall elongation process. Thus, we found that mutations in domain IA of MreB and in the periplasmic domain of PBP2 and RodA can restore growth and rod shape to ΔrodZ cells, possibly by changing the requirements of MreB in the process. PMID:23301723

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.

Overview and Current Status of Analyses of Potential LEU Design Concepts for TREAT

DOE Office of Scientific and Technical Information (OSTI.GOV)

Connaway, H. M.; Kontogeorgakos, D. C.; Papadias, D. D.

2015-10-01

Neutronic and thermal-hydraulic analyses have been performed to evaluate the performance of different low-enriched uranium (LEU) fuel design concepts for the conversion of the Transient Reactor Test Facility (TREAT) from its current high-enriched uranium (HEU) fuel. TREAT is an experimental reactor developed to generate high neutron flux transients for the testing of nuclear fuels. The goal of this work was to identify an LEU design which can maintain the performance of the existing HEU core while continuing to operate safely. A wide variety of design options were considered, with a focus on minimizing peak fuel temperatures and optimizing the powermore » coupling between the TREAT core and test samples. Designs were also evaluated to ensure that they provide sufficient reactivity and shutdown margin for each control rod bank. Analyses were performed using the core loading and experiment configuration of historic M8 Power Calibration experiments (M8CAL). The Monte Carlo code MCNP was utilized for steady-state analyses, and transient calculations were performed with the point kinetics code TREKIN. Thermal analyses were performed with the COMSOL multi-physics code. Using the results of this study, a new LEU Baseline design concept is being established, which will be evaluated in detail in a future report.« less

76 FR 52356 - Indiana Michigan Power Company, Donald C. Cook Nuclear Plant, Unit 1; Environmental Assessment...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-08-22

... adversely affect plant safety, and would have no adverse effect on the probability of any accident. For the accidents that involve damage or melting of the fuel in the reactor core, fuel rod integrity has been shown to be unaffected by extended burnup under consideration; therefore, the probability of an accident...

77 FR 51071 - Indiana Michigan Power Company, Donald C. Cook Nuclear Plant, Unit 2, Environmental Assessment...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-08-23

... adversely affect plant safety, and would have no adverse effect on the probability of any accident. For the accidents that involve damage or melting of the fuel in the reactor core, fuel rod integrity has been shown to be unaffected by extended burnup under consideration; therefore, the consequences of an accident...

COUPLING

DOEpatents

Hawke, B.C.

1963-02-26

This patent relates to a releasable coupling connecting a control rod to a control rod drive. This remotely operable coupling mechanism can connect two elements which are laterally and angviarly misaligned, and provides a means for sensing the locked condition of the elements. The coupling utilizes a spherical bayonet joint which is locked against rotation by a ball detent lock. (AEC)

Decommissioning of the Dragon High Temperature Reactor (HTR) Located at the Former United Kingdom Atomic Energy Authority (UKAEA) Research Site at Winfrith - 13180

DOE Office of Scientific and Technical Information (OSTI.GOV)

Smith, Anthony A.

2013-07-01

The Dragon Reactor was constructed at the United Kingdom Atomic Energy Research Establishment at Winfrith in Dorset through the late 1950's and into the early 1960's. It was a High Temperature Gas Cooled Reactor (HTR) with helium gas coolant and graphite moderation. It operated as a fuel testing and demonstration reactor at up to 20 MW (Thermal) from 1964 until 1975, when international funding for this project was terminated. The fuel was removed from the core in 1976 and the reactor was put into Safestore. To meet the UK's Nuclear Decommissioning Authority (NDA) objective to 'drive hazard reduction' [1] itmore » is necessary to decommission and remediate all the Research Sites Restoration Ltd (RSRL) facilities. This includes the Dragon Reactor where the activated core, pressure vessel and control rods and the contaminated primary circuit (including a {sup 90}Sr source) still remain. It is essential to remove these hazards at the appropriate time and return the area occupied by the reactor to a safe condition. (author)« less

Time-delay control of a magnetic levitated linear positioning system

NASA Technical Reports Server (NTRS)

Tarn, J. H.; Juang, K. Y.; Lin, C. E.

1994-01-01

In this paper, a high accuracy linear positioning system with a linear force actuator and magnetic levitation is proposed. By locating a permanently magnetized rod inside a current-carrying solenoid, the axial force is achieved by the boundary effect of magnet poles and utilized to power the linear motion, while the force for levitation is governed by Ampere's Law supplied with the same solenoid. With the levitation in a radial direction, there is hardly any friction between the rod and the solenoid. The high speed motion can hence be achieved. Besides, the axial force acting on the rod is a smooth function of rod position, so the system can provide nanometer resolution linear positioning to the molecule size. Since the force-position relation is highly nonlinear, and the mathematical model is derived according to some assumptions, such as the equivalent solenoid of the permanently magnetized rod, so there exists unknown dynamics in practical application. Thus 'robustness' is an important issue in controller design. Meanwhile the load effect reacts directly on the servo system without transmission elements, so the capability of 'disturbance rejection; is also required. With the above consideration, a time-delay control scheme is chosen and applied. By comparing the input-output relation and the mathematical model, the time-delay controller calculates an estimation of unmodeled dynamics and disturbances and then composes the desired compensation into the system. Effectiveness of the linear positioning system and control scheme are illustrated with simulation results.

Computer program for optimal BWR congtrol rod programming

DOE Office of Scientific and Technical Information (OSTI.GOV)

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 peakmore » 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.« less

Pulse repetition rate multiplication by Talbot effect in a coaxial fiber

NASA Astrophysics Data System (ADS)

Dhingra, Nikhil; Saxena, Geetika Jain; Anand, Jyoti; Sharma, Enakshi K.

2018-03-01

We use a coaxial fiber, which is a cylindrical coupled waveguide structure consisting of two concentric cores, the inner rod and an outer ring core as a first order dispersive media to achieve temporal Talbot effect for pulse repetition rate multiplication (PRRM) in high bit rate optical fiber communication. It is observed that for an input Gaussian pulse train with pulse width, 2τ0=1ps at a repetition rate of 40 Gbps (repetition period, T=25ps), an output repetition rate of 640 Gbps can be achieved without significant distortion at a length of 40.92 m.

The Effects of Diabetic Retinopathy and Pan-Retinal Photocoagulation on Photoreceptor Cell Function as Assessed by Dark Adaptometry

PubMed Central

Bavinger, J. Clay; Dunbar, Grace E.; Stem, Maxwell S.; Blachley, Taylor S.; Kwark, Leon; Farsiu, Sina; Jackson, Gregory R.; Gardner, Thomas W.

2016-01-01

Purpose The pathophysiology of vision loss in persons with diabetic retinopathy (DR) is complex and incompletely defined. We hypothesized that retinal pigment epithelium (RPE) and rod and cone photoreceptor dysfunction, as measured by dark adaptometry, would increase with severity of DR, and that pan-retinal photocoagulation (PRP) would exacerbate this dysfunction. Methods Dark adaptation (DA) was measured in subjects with diabetes mellitus and healthy controls. Dark adaptation was measured at 5° superior to the fovea following a flash bleach, and the data were analyzed to yield cone and rod sensitivity curves. Retinal layer thicknesses were quantified using spectral-domain optical coherence tomography (OCT). Results The sample consisted of 23 controls and 73 diabetic subjects. Subjects with moderate nonproliferative diabetic retinopathy (NPDR) exhibited significant impairment of rod recovery rate compared with control subjects (P = 0.04). Cone sensitivity was impaired in subjects with proliferative diabetic retinopathy (PDR) (type 1 diabetes mellitus [T1DM]: P = 0.0047; type 2 diabetes mellitus [T2DM]: P < 0.001). Subjects with untreated PDR compared with subjects treated with PRP exhibited similar rod recovery rates and cone sensitivities. Thinner RPE as assessed by OCT was associated with slower rod recovery and lower cone sensitivity, and thinner photoreceptor inner segment/outer segment layer was associated with lower cone sensitivity. Conclusions The results suggest that RPE and photoreceptor cell dysfunction, as assessed by cone sensitivity level and rod- and RPE-mediated dark adaptation, progresses with worsening DR, and rod recovery dysfunction occurs earlier than cone dysfunction. Function was preserved following PRP. The findings suggest multiple defects in retinoid function and provide potential points to improve visual function in persons with PDR. PMID:26803796

Performance of U3Si2 Fuel in a Reactivity Insertion Accident

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cheng, Lap Y.; Cuadra, Arantxa; Todosow, Michael

In this study we examined the performance of the U3Si2 fuel cladded with Zircaloy (Zr) in a reactivity insertion accident (RIA) in a PWR core. The power excursion as a result of a $1 reactivity insertion was calculated by a TRACE PWR plant model using point-kinetics, for alternative cores with UO2 and U3Si2 fuel assemblies. The point-kinetics parameters (feedback coefficients, prompt-neutron lifetime and group constants for six delayed-neutron groups) were obtained from beginning-of-cycle equilibrium full core calculations with PARCS. In the PARCS core calculations, the few-group parameters were developed utilizing the TRITON/NEWT tools in the SCALE package. In order tomore » assess the fuel response in finer detail (e.g. the maximum fuel temperature) the power shape and thermal boundary conditions from the TRACE/PARCS calculations were used to drive a BISON model of a fuel pin with U3Si2 and UO2 respectively. For a $1 reactivity transient both TRACE and BISON predicted a higher maximum fuel temperature for the UO2 fuel than the U3Si2 fuel. Furthermore, BISON is noted to calculate a narrower gap and a higher gap heat transfer coefficient than TRACE. This resulted in BISON predicting consistently lower fuel temperatures than TRACE. This study also provides a systematic comparison between TRACE and BISON using consistent transient boundary conditions. The TRACE analysis of the RIA only reflects the core-wide response in power. A refinement to the analysis would be to predict the local peaking in a three-dimensional core as a result of control rod ejection.« less

Checkerboard seed-blanket thorium fuel core concepts for heavy water moderated reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bromley, B.P.; Hyland, B.

2013-07-01

New reactor concepts to implement thorium-based fuel cycles have been explored to achieve maximum resource utilization. Pressure tube heavy water reactors (PT-HWR) are highly advantageous for implementing the use of thorium-based fuels because of their high neutron economy and on-line re-fuelling capability. The use of heterogeneous seed-blanket core concepts in a PT-HWR where higher-fissile-content seed fuel bundles are physically separate from lower-fissile-content blanket bundles allows more flexibility and control in fuel management to maximize the fissile utilization and conversion of fertile fuel. The lattice concept chosen was a 35-element bundle made with a homogeneous mixture of reactor grade Pu (aboutmore » 67 wt% fissile) and Th, and with a central zirconia rod to help reduce coolant void reactivity. Several checkerboard heterogeneous seed-blanket core concepts with plutonium-thorium-based fuels in a 700-MWe-class PT-HWR were analyzed, using a once-through thorium (OTT) cycle. Different combinations of seed and blanket fuel were tested to determine the impact on core-average burnup, fissile utilization, power distributions, and other performance parameters. It was found that various checkerboard core concepts can achieve a fissile utilization that is up to 26% higher than that achieved in a PT-HWR using more conventional natural uranium fuel bundles. Up to 60% of the Pu is consumed; up to 43% of the energy is produced from thorium, and up to 303 kg/year of Pa-233/U-233/U-235 are produced. Checkerboard cores with about 50% of low-power blanket bundles may require power de-rating (65% to 74%) to avoid exceeding maximum limits for channel and bundle powers and linear element ratings. (authors)« less

Annular seed-blanket thorium fuel core concepts for heavy water moderated reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bromley, B.P.; Hyland, B.

2013-07-01

New reactor concepts to implement thorium-based fuel cycles have been explored to achieve maximum resource utilization. Pressure tube heavy water reactors (PT-HWR) are highly advantageous for implementing the use of thorium-based fuels because of their high neutron economy and on-line re-fuelling capability. The use of heterogeneous seed-blanket core concepts in a PT-HWR where higher-fissile-content seed fuel bundles are physically separate from lower-fissile-content blanket bundles allows more flexibility and control in fuel management to maximize the fissile utilization and conversion of fertile fuel. The lattice concept chosen is a 35-element bundle made with a homogeneous mixture of reactor grade Pu andmore » Th, and with a central zirconia rod to help reduce coolant void reactivity. Several annular heterogeneous seed-blanket core concepts with plutonium-thorium-based fuels in a 700-MWe-class PT-HWR were analyzed, using a once-through thorium (OTT) cycle. Different combinations of seed and blanket fuel were tested to determine the impact on core-average burnup, fissile utilization, power distributions, and other performance parameters. It was found that the various core concepts can achieve a fissile utilization that is up to 30% higher than is currently achieved in a PT-HWR using conventional natural uranium fuel bundles. Up to 67% of the Pu is consumed; up to 43% of the energy is produced from thorium, and up to 363 kg/year of U-233 is produced. Seed-blanket cores with ∼50% content of low-power blanket bundles may require power de-rating (∼58% to 65%) to avoid exceeding maximum limits for peak channel power, bundle power and linear element ratings. (authors)« less

Domain and network aggregation of CdTe quantum rods within Langmuir Blodgett monolayers

NASA Astrophysics Data System (ADS)

Zimnitsky, Dmitry; Xu, Jun; Lin, Zhiqun; Tsukruk, Vladimir V.

2008-05-01

Control over the organization of quantum rods was demonstrated by changing the surface area at the air-liquid interface by means of the Langmuir-Blodgett (LB) technique. The LB isotherm of CdTe quantum rods capped with a mixture of alkylphosphines shows a transition point in the liquid-solid state, which is caused by the inter-rod reorganization. As we observed, at low surface pressure the quantum rods are assembled into round-shaped aggregates composed of a monolayer of nanorods packed in limited-size clusters with random orientation. The increase of the surface pressure leads to the rearrangement of these aggregates into elongated bundles composed of uniformly oriented nanorod clusters. Further compression results in denser packing of nanorods aggregates and in the transformation of monolayered domains into a continuous network of locally ordered quantum rods.

Light Water Breeder Reactor fuel rod design and performance characteristics (LWBR Development Program)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Campbell, W.R.; Giovengo, J.F.

1987-10-01

Light Water Breeder Reactor (LWBR) fuel rods were designed to provide a reliable fuel system utilizing thorium/uranium-233 mixed-oxide fuel while simultaneously minimizing structural material to enhance fuel breeding. The fuel system was designed to be capable of operating successfully under both load follow and base load conditions. The breeding objective required thin-walled, low hafnium content Zircaloy cladding, tightly spaced fuel rods with a minimum number of support grid levels, and movable fuel rod bundles to supplant control rods. Specific fuel rod design considerations and their effects on performance capability are described. Successful completion of power operations to over 160 percentmore » of design lifetime including over 200 daily load follow cycles has proven the performance capability of the fuel system. 68 refs., 19 figs., 44 tabs.« less

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.

A method for development of efficient 3D models for neutronic calculations of ASTRA critical facility using experimental information

DOE Office of Scientific and Technical Information (OSTI.GOV)

Balanin, A. L.; Boyarinov, V. F.; Glushkov, E. S.

The application of experimental information on measured axial distributions of fission reaction rates for development of 3D numerical models of the ASTRA critical facility taking into account azimuthal asymmetry of the assembly simulating a HTGR with annular core is substantiated. Owing to the presence of the bottom reflector and the absence of the top reflector, the application of 2D models based on experimentally determined buckling is impossible for calculation of critical assemblies of the ASTRA facility; therefore, an alternative approach based on the application of the extrapolated assembly height is proposed. This approach is exemplified by the numerical analysis ofmore » experiments on measurement of efficiency of control rods mockups and protection system (CPS).« less

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.

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)

Light responses in rods of vitamin A-deprived Xenopus.

PubMed

Solessio, Eduardo; Umino, Yumiko; Cameron, David A; Loew, Ellis; Engbretson, Gustav A; Knox, Barry E; Barlow, Robert B

2009-09-01

Accumulation of free opsin by mutations in rhodopsin or insufficiencies in the visual cycle can lead to retinal degeneration. Free opsin activates phototransduction; however, the link between constitutive activation and retinal degeneration is unclear. In this study, the photoresponses of Xenopus rods rendered constitutively active by vitamin A deprivation were examined. Unlike their mammalian counterparts, Xenopus rods do not degenerate. Contrasting phototransduction in vitamin A-deprived Xenopus rods with phototransduction in constitutively active mammalian rods may provide new understanding of the mechanisms that lead to retinal degeneration. The photocurrents of Xenopus tadpole rods were measured with suction electrode recordings, and guanylate cyclase activity was measured with the IBMX (3-isobutyl-1-methylxanthine) jump technique. The amount of rhodopsin in rods was determined by microspectrophotometry. The vitamin A-deprived rod outer segments were 60% to 70% the length and diameter of the rods in age-matched animals. Approximately 90% of its opsin content was in the free or unbound form. Analogous to bleaching adaptation, the photoresponses were desensitized (10- to 20-fold) and faster. Unlike bleaching adaptation, the vitamin A-deprived rods maintained near normal saturating (dark) current densities by developing abnormally high rates of cGMP synthesis. Their rate of cGMP synthesis in the dark (15 seconds(-1)) was twofold greater than the maximum levels attainable by control rods ( approximately 7 seconds(-1)). Preserving circulating current density and response range appears to be an important goal for rod homeostasis. However, the compensatory changes associated with vitamin A deprivation in Xenopus rods come at the high metabolic cost of a 15-fold increase in basal ATP consumption.

The effect of silicon on the interaction between metallic uranium and aluminum: A 50 year long diffusion experiment

NASA Astrophysics Data System (ADS)

Leenaers, A.; Detavernier, C.; Van den Berghe, S.

2008-11-01

The core of the BR1 research reactor at SCK•CEN, Mol (Belgium) has a graphite matrix loaded with fuel rods consisting of a natural uranium slug in aluminum cladding. The BR1 reactor has been in operation since 1956 and still contains its original fuel rods. After more than 50 years irradiation at low temperature, some of the fuel rods have been examined. Fabrication reports indicate that a so-called AlSi bonding layer and an U(Al,Si) 3 anti-diffusion layer on the natural uranium fuel slug were applied to limit the interaction between the uranium fuel and aluminum cladding. The microstructure of the fuel, bonding and anti-diffusion layer and cladding were analysed using optical microscopy, scanning electron microscopy and electron microprobe analysis. It was found that the AlSi bonding layer does provide a tight bond between fuel and cladding but that it is a thin USi layer that acts as effective anti-diffusion layer and not the intended U(Al,Si) 3 layer.

Landing gear energy absorption system

NASA Technical Reports Server (NTRS)

Hansen, Christopher P. (Inventor)

1994-01-01

A landing pad system is described for absorbing horizontal and vertical impact forces upon engagement with a landing surface where circumferentially arranged landing struts respectively have a clevis which receives a slidable rod member and where the upper portion of a slidable rod member is coupled to the clevis by friction washers which are force fit onto the rod member to provide for controlled constant force energy absorption when the rod member moves relative to the clevis. The lower end of the friction rod is pivotally attached by a ball and socket to a support plate where the support plate is arranged to slide in a transverse direction relative to a housing which contains an energy absorption material for absorbing energy in a transverse direction.

Pulse compression of a high-power thin disk laser using rod-type fiber amplifiers.

PubMed

Saraceno, C J; Heckl, O H; Baer, C R E; Südmeyer, T; Keller, U

2011-01-17

We report on two pulse compressors for a high-power thin disk laser oscillator using rod-type fiber amplifiers. Both systems are seeded by a standard SESAM modelocked thin disk laser that delivers 16 W of average power at a repetition rate of 10.6 MHz with a pulse energy of 1.5 μJ and a pulse duration of 1 ps. We discuss two results with different fiber parameters with different trade-offs in pulse duration, average power, damage and complexity. The first amplifier setup consists of a Yb-doped fiber amplifier with a 2200 μm2 core area and a length of 55 cm, resulting in a compressed average power of 55 W with 98-fs pulses at a repetition rate of 10.6 MHz. The second system uses a shorter 36-cm fiber with a larger core area of 4500 μm2. In a stretcher-free configuration we obtained 34 W of compressed average power and 65-fs pulses. In both cases peak powers of > 30 MW were demonstrated at several μJ pulse energies. The power scaling limitations due to damage and self-focusing are discussed.

Optically Active CdSe-Dot/CdS-Rod Nanocrystals with Induced Chirality and Circularly Polarized Luminescence.

PubMed

Cheng, Jiaji; Hao, Junjie; Liu, Haochen; Li, Jiagen; Li, Junzi; Zhu, Xi; Lin, Xiaodong; Wang, Kai; He, Tingchao

2018-05-30

Ligand-induced chirality in semiconductor nanocrystals (NCs) has attracted attention because of the tunable optical properties of the NCs. Induced circular dichroism (CD) has been observed in CdX (X = S, Se, Te) NCs and their hybrids, but circularly polarized luminescence (CPL) in these fluorescent nanomaterials has been seldom reported. Herein, we describe the successful preparation of l- and d-cysteine-capped CdSe-dot/CdS-rods (DRs) with tunable CD and CPL behaviors and a maximum anisotropic factor ( g lum ) of 4.66 × 10 -4 . The observed CD and CPL activities are sensitive to the relative absorption ratio of the CdS shell to the CdSe core, suggesting that the anisotropic g-factors in both CD and CPL increase to some extent for a smaller shell-to-core absorption ratio. In addition, the molar ratio of chiral cysteine to the DRs is investigated. Instead of enhancing the chiral interactions between the chiral molecules and DRs, an excess of cysteine molecules in aqueous solution inhibits both the CD and CPL activities. Such chiral and emissive NCs provide an ideal platform for the rational design of semiconductor nanomaterials with chiroptical properties.

OVERALL CONTROL SYSTEM FOR HIGH FLUX PILE

DOEpatents

Newson, H.W.; Durham, N.C.; Wigner, E.P.; Princeton, N.J.; Epler, E.P.

1961-05-23

A control system is given for a high fiux reactor incorporating an anti- scram control feature whereby a neutron absorbing control rod acts as a fine adjustment while a neutron absorbing shim rod, actuated upon a command received from reactor period and level signals, has substantially greater effect on the neutron level and is moved prior to scram conditions to alter the reactor activity before a scram condition is created. Thus the probability that a scram will have to be initiated is substantially decreased.

Pathology in a tube step 2: simple rapid fabrication of curved circular cross section millifluidic channels for biopsy preparation/3D imaging towards pancreatic cancer detection and diagnosis

NASA Astrophysics Data System (ADS)

Das, Ronnie; Burfeind, Chris W.; Lim, Saniel D.; Patle, Shubham; Seibel, Eric J.

2018-02-01

3D pathology is intrinsically dependent on 3D microscopy, or the whole tissue imaging of patient tissue biopsies (TBs). Consequently, unsectioned needle specimens must be processed whole: a procedure which cannot necessarily be accomplished through manual methods, or by retasking automated pathology machines. Thus "millifluidic" devices (for millimeter-scale biopsies) are an ideal solution for tissue handling/preparation. TBs are large, messy and a solid-liquid mixture; they vary in material, geometry and structure based on the organ biopsied, the clinician skill and the needle type used. As a result, traditional microfluidic devices are insufficient to handle such mm-sized samples and their associated fabrication techniques are impractical and costly with respect to time/efficiency. Our research group has devised a simple, rapid fabrication process for millifluidic devices using jointed skeletal molds composed of machined, reusable metal rods, segmented rods and stranded wire as structural cores; these cores are surrounded by Teflon outer housing. We can therefore produce curving, circular-cross-section (CCCS) millifluidic channels in rapid fashion that cannot normally be achieved by microfabrication, micro-/CNC-machining, or 3D printing. The approach has several advantages. CLINICAL: round channels interface coring needles. PROCESSING: CCCS channels permit multi-layer device designs for additional (processing, monitoring, testing) stages. REUSABILITY: for a biopsy/needle diameter, molding (interchangeable) components may be produced one-time then reused for other designs. RAPID: structural cores can be quickly removed due to Teflon®'s ultra-low friction; housing may be released with ethanol; PDMS volumes cure faster since metal skeleton molds conduct additional heat from within the curing elastomer.

Defect scriber

DOEpatents

Russell, Harold C.

1979-01-01

This disclosure describes a device for repeatably scribing a V-shaped scratch having sharply defined dimensions on the interior surface of a nuclear reactor fuel rod tube. A cutting tool having a V-shaped cutting tip is supported within the fuel rod tube so that the V-shaped cutting tip can be pivoted about an axis and scribe a scratch on the interior surface of the fuel rod tube. Lengthwise the scratch runs parallel to a line drawn through the axis of the fuel rod tube and is in the shape of an arc, and widthwise the scratch is V-shaped. This shape is used because the dimensions of the scratch can be plugged into appropriate formulas to calculate stress intensity of cracks in fuel rod tubes. Since the fuel rod tubes which are to be scribed may be radioactive, the scratching assembly is designed for use in a fixture which allows it to be operated in a cave by remote control handling devices.

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)

VERA and VERA-EDU 3.5 Release Notes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sieger, Matt; Salko, Robert K.; Kochunas, Brendan M.

The Virtual Environment for Reactor Applications components included in this distribution include selected computational tools and supporting infrastructure that solve neutronics, thermal-hydraulics, fuel performance, and coupled neutronics-thermal hydraulics problems. The infrastructure components provide a simplified common user input capability and provide for the physics integration with data transfer and coupled-physics iterative solution algorithms. Neutronics analysis can be performed for 2D lattices, 2D core and 3D core problems for pressurized water reactor geometries that can be used to calculate criticality and fission rate distributions by pin for input fuel compositions. MPACT uses the Method of Characteristics transport approach for 2D problems.more » For 3D problems, MPACT uses the 2D/1D method which uses 2D MOC in a radial plane and diffusion or SPn in the axial direction. MPACT includes integrated cross section capabilities that provide problem-specific cross sections generated using the subgroup methodology. The code can be executed both 2D and 3D problems in parallel to reduce overall run time. A thermal-hydraulics capability is provided with CTF (an updated version of COBRA-TF) that allows thermal-hydraulics analyses for single and multiple assemblies using the simplified VERA common input. This distribution also includes coupled neutronics/thermal-hydraulics capabilities to allow calculations using MPACT coupled with CTF. The VERA fuel rod performance component BISON calculates, on a 2D or 3D basis, fuel rod temperature, fuel rod internal pressure, free gas volume, clad integrity and fuel rod waterside diameter. These capabilities allow simulation of power cycling, fuel conditioning and deconditioning, high burnup performance, power uprate scoping studies, and accident performance. Input/Output capabilities include the VERA Common Input (VERAIn) script which converts the ASCII common input file to the intermediate XML used to drive all of the physics codes in the VERA Core Simulator (VERA-CS). VERA component codes either input the VERA XML format directly, or provide a preprocessor which can convert the XML into native input. VERAView is an interactive graphical interface for the visualization and engineering analyses of output data from VERA. The python-based software is easy to install and intuitive to use, and provides instantaneous 2D and 3D images, 1D plots, and alpha-numeric data from VERA multi-physics simulations. Testing within CASL has focused primarily on Westinghouse four-loop reactor geometries and conditions with example problems included in the distribution.« less

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.

Accident Analysis for the NIST Research Reactor Before and After Fuel Conversion

DOE Office of Scientific and Technical Information (OSTI.GOV)

Baek J.; Diamond D.; Cuadra, A.

Postulated accidents have been analyzed for the 20 MW D2O-moderated research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The analysis has been carried out for the present core, which contains high enriched uranium (HEU) fuel and for a proposed equilibrium core with low enriched uranium (LEU) fuel. The analyses employ state-of-the-art calculational methods. Three-dimensional Monte Carlo neutron transport calculations were performed with the MCNPX code to determine homogenized fuel compositions in the lower and upper halves of each fuel element and to determine the resulting neutronic properties of the core. The accident analysis employed a modelmore » of the primary loop with the RELAP5 code. The model includes the primary pumps, shutdown pumps outlet valves, heat exchanger, fuel elements, and flow channels for both the six inner and twenty-four outer fuel elements. Evaluations were performed for the following accidents: (1) control rod withdrawal startup accident, (2) maximum reactivity insertion accident, (3) loss-of-flow accident resulting from loss of electrical power with an assumption of failure of shutdown cooling pumps, (4) loss-of-flow accident resulting from a primary pump seizure, and (5) loss-of-flow accident resulting from inadvertent throttling of a flow control valve. In addition, natural circulation cooling at low power operation was analyzed. The analysis shows that the conversion will not lead to significant changes in the safety analysis and the calculated minimum critical heat flux ratio and maximum clad temperature assure that there is adequate margin to fuel failure.« less

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

Transient Simulation of the Multi-SERTTA Experiment with MAMMOTH

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ortensi, Javier; Baker, Benjamin; Wang, Yaqi

This work details the MAMMOTH reactor physics simulations of the Static Environment Rodlet Transient Test Apparatus (SERTTA) conducted at Idaho National Laboratory in FY-2017. TREAT static-environment experiment vehicles are being developed to enable transient testing of Pressurized Water Reactor (PWR) type fuel specimens, including fuel concepts with enhanced accident tolerance (Accident Tolerant Fuels, ATF). The MAMMOTH simulations include point reactor kinetics as well as spatial dynamics for a temperature-limited transient. The strongly coupled multi-physics solutions of the neutron flux and temperature fields are second order accurate both in the spatial and temporal domains. MAMMOTH produces pellet stack powers that are within 1.5% of the Monte Carlo reference solutions. Some discrepancies between the MCNP model used in the design of the flux collars and the Serpent/MAMMOTH models lead to higher power and energy deposition values in Multi-SERTTA unit 1. The TREAT core results compare well with the safety case computed with point reactor kinetics in RELAP5-3D. The reactor period is 44 msec, which corresponds to a reactivity insertion of 2.685% delta k/kmore » $. The peak core power in the spatial dynamics simulation is 431 MW, which the point kinetics model over-predicts by 12%. The pulse width at half the maximum power is 0.177 sec. Subtle transient effects are apparent at the beginning insertion in the experimental samples due to the control rod removal. Additional difference due to transient effects are observed in the sample powers and enthalpy. The time dependence of the power coupling factor (PCF) is calculated for the various fuel stacks of the Multi-SERTTA vehicle. Sample temperatures in excess of 3100 K, the melting point UO$$_2$$, are computed with the adiabatic heat transfer model. The planned shaped-transient might introduce additional effects that cannot be predicted with PRK models. Future modeling will be focused on the shaped-transient by improving the control rod models in MAMMOTH and adding the BISON thermo-elastic models and thermal-fluids heat transfer.« less

Experimental Study on Surrogate Nuclear Fuel Rods under Reversed Cyclic Bending

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wang, Hong; Wang, Jy-An John

The mechanical behavior of spent nuclear fuel (SNF) rods under reversed cyclic bending or bending fatigue must be understood to evaluate their vibration integrity in a transportation environment. This is especially important for high-burnup fuels (>45 GWd/MTU), which have the potential for increased structural damage. It has been demonstrated that the bending fatigue of SNF rods can be effectively studied using surrogate rods. In this investigation, surrogate rods made of stainless steel (SS) 304 cladding and aluminum oxide pellets were tested under load or moment control at a variety of amplitude levels at 5 Hz using the Cyclic Integrated Reversible-Bendingmore » Fatigue Tester developed at Oak Ridge National Laboratory. The behavior of the rods was further characterized using flexural rigidity and hysteresis data, and fractography was performed on the failed rods. The proposed surrogate rods captured many of the characteristics of deformation and failure mode observed in SNF, including the linear-to-nonlinear deformation transition and large residual curvature in static tests, PPI and PCMI failure mechanisms, and large variation in the initial structural condition. Rod degradation was measured and characterized by measuring the flexural rigidity; the degradation of the rigidity depended on both the moment amplitude applied and the initial structural condition of the rods. It was also shown that a cracking initiation site can be located on the internal surface or the external surface of cladding. Finally, fatigue damage to the bending rods can be described in terms of flexural rigidity, and the fatigue life of rods can be predicted once damage model parameters are properly evaluated. The developed experimental approach, test protocol, and analysis method can be used to study the vibration integrity of SNF rods in the future.« less

Stochastic dynamics of penetrable rods in one dimension: occupied volume and spatial order.

PubMed

Craven, Galen T; Popov, Alexander V; Hernandez, Rigoberto

2013-06-28

The occupied volume of a penetrable hard rod (HR) system in one dimension is probed through the use of molecular dynamics simulations. In these dynamical simulations, collisions between penetrable rods are governed by a stochastic penetration algorithm (SPA), which allows for rods to either interpenetrate with a probability δ, or collide elastically otherwise. The limiting values of this parameter, δ = 0 and δ = 1, correspond to the HR and the ideal limits, respectively. At intermediate values, 0 < δ < 1, mixing of mutually exclusive and independent events is observed, making prediction of the occupied volume nontrivial. At high hard core volume fractions φ0, the occupied volume expression derived by Rikvold and Stell [J. Chem. Phys. 82, 1014 (1985)] for permeable systems does not accurately predict the occupied volume measured from the SPA simulations. Multi-body effects contribute significantly to the pair correlation function g2(r) and the simplification by Rikvold and Stell that g2(r) = δ in the penetrative region is observed to be inaccurate for the SPA model. We find that an integral over the penetrative region of g2(r) is the principal quantity that describes the particle overlap ratios corresponding to the observed penetration probabilities. Analytic formulas are developed to predict the occupied volume of mixed systems and agreement is observed between these theoretical predictions and the results measured from simulation.

Characterization of the genuine type 2 chromatic acclimation in the two Geminocystis cyanobacteria.

PubMed

Hirose, Yuu; Misawa, Naomi; Yonekawa, Chinatsu; Nagao, Nobuyoshi; Watanabe, Mai; Ikeuchi, Masahiko; Eki, Toshihiko

2017-08-01

Certain cyanobacteria can adjust the wavelengths of light they absorb by remodeling their photosynthetic antenna complex phycobilisome via a process called chromatic acclimation (CA). Although several types of CA have been reported, the diversity of the molecular mechanisms of CA among the cyanobacteria phylum is not fully understood. Here, we characterized the molecular process of CA of Geminocystis sp. strains National Institute of Environmental Studies (NIES)-3708 and NIES-3709. Absorption and fluorescence spectroscopy revealed that both strains dramatically alter their phycoerythrin content in response to green and red light. Whole-genome comparison revealed that the two strains share the typical phycobilisome structure consisting of a central core and peripheral rods, but they differ in the number of rod linkers of phycoerythrin and thus have differing capacity for phycoerythrin accumulation. RNA sequencing analysis suggested that the length of phycoerythrin rods in each phycobilisome is strictly regulated by the green light and red light-sensing CcaS/R system, whereas the total number of phycobilisomes is governed by the excitation-balancing system between phycobilisomes and photosystems. We reclassify the conventional CA types based on the genome information and designate CA of the two strains as genuine type 2, where components of phycoerythrin, but not rod-membrane linker of phycocyanin, are regulated by the CcaS/R system. © The Author 2017. Published by Oxford University Press on behalf of Kazusa DNA Research Institute.

Sensitivity and Uncertainty Analysis of Plutonium and Cesium Isotopes in Modeling of BR3 Reactor Spent Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Conant, Andrew; Erickson, Anna; Robel, Martin

Nuclear forensics has a broad task to characterize recovered nuclear or radiological material and interpret the results of investigation. One approach to isotopic characterization of nuclear material obtained from a reactor is to chemically separate and perform isotopic measurements on the sample and verify the results with modeling of the sample history, for example, operation of a nuclear reactor. The major actinide plutonium and fission product cesium are commonly measured signatures of the fuel history in a reactor core. This study investigates the uncertainty of the plutonium and cesium isotope ratios of a fuel rod discharged from a research pressurizedmore » water reactor when the location of the sample is not known a priori. A sensitivity analysis showed overpredicted values for the 240Pu/ 239Pu ratio toward the axial center of the rod and revealed a lower probability of the rod of interest (ROI) being on the periphery of the assembly. The uncertainty analysis found the relative errors due to only the rod position and boron concentration to be 17% to 36% and 7% to 15% for the 240Pu/ 239Pu and 137Cs/ 135Cs ratios, respectively. Lastly, this study provides a method for uncertainty quantification of isotope concentrations due to the location of the ROI. Similar analyses can be performed to verify future chemical and isotopic analyses.« less

Association of anatase (TiO2) and microbes: unusual fossilization effect or a potential biosignature?

USGS Publications Warehouse

Glamoclija, Mihaela; Andrew Steele,; Marc Fries,; Juergen Schieber,; Voytek, Mary A.; Charles S. Cockell,

2015-01-01

We combined microbial paleontology and molecular biology methods to study the Eyreville B drill core from the 35.3-Ma-old Chesapeake Bay impact structure,Virginia, USA. The investigated sample is a pyrite vein collected from the 1353.81-1353.89 m depth interval, located within a section of biotite granite. The granite is a pre-impact rock that was disrupted by the impact event. A search for inorganic (mineral) biosignatures revealed the presence of micron-size rod morphologies of anatase (TiO2) embedded in chlorite coatings on pyrite grains. Neither the Acridine Orange microbial probe nor deoxyribonucleic acid (DNA) extraction followed by polymerase chain reaction (PCR) amplifi cation showed the presence of DNA or ribonucleic acid (RNA) at the location of anatase rods, implying the absence of viable cells in the investigated area. A Nile Red microbial probe revealed the presence of lipids in the rods. Because most of the lipids are resistant over geologic time spans, they are good biomarkers, and they are an indicator of biogenicity for these possibly 35-Ma-old microbial fossils. The mineral assemblage suggests that rod morphologies are associated with low-temperature (<100 °C) hydrothermal alteration that involved aqueous fl uids. The temporal constraints on the anatase fossils are still uncertain because pre-impact alteration of the granite and postimpact heating may have provided identical conditions for anatase precipitation and microbial preservation.

Self-assembly of proteins into a three-dimensional multilayer system: investigation of the surface of the human fungal pathogen Aspergillus fumigatus.

PubMed

Zykwinska, Agata; Pihet, Marc; Radji, Sadia; Bouchara, Jean-Philippe; Cuenot, Stéphane

2014-06-01

Hydrophobins are small surface active proteins that fulfil a wide spectrum of functions in fungal growth and development. The human fungal pathogen Aspergillus fumigatus expresses RodA hydrophobins that self-assemble on the outer conidial surface into tightly organized nanorods known as rodlets. AFM investigation of the conidial surface allows us to evidence that RodA hydrophobins self-assemble into rodlets through bilayers. Within bilayers, hydrophilic domains of hydrophobins point inward, thus making a hydrophilic core, while hydrophobic domains point outward. AFM measurements reveal that several rodlet bilayers are present on the conidial surface thus showing that proteins self-assemble into a complex three-dimensional multilayer system. The self-assembly of RodA hydrophobins into rodlets results from attractive interactions between stacked β-sheets, which conduct to a final linear cross-β spine structure. A Monte Carlo simulation shows that anisotropic interactions are the main driving forces leading the hydrophobins to self-assemble into parallel rodlets, which are further structured in nanodomains. Taken together, these findings allow us to propose a mechanism, which conducts RodA hydrophobins to a highly ordered rodlet structure. The mechanism of hydrophobin assembly into rodlets offers new prospects for the development of more efficient strategies leading to disruption of rodlet formation allowing a rapid detection of the fungus by the immune system. Copyright © 2014 Elsevier B.V. All rights reserved.

Cutaneous and inflammatory response to long-term percutaneous implants of sphere-templated porous/solid poly(HEMA) and silicone in Mice

PubMed Central

Fleckman, Philip; Usui, Marcia; Zhao, Ge; Underwood, Robert; Maginness, Max; Marshall, Andrew; Glaister, Christine; Ratner, Buddy; Olerud, John

2012-01-01

This study investigates mouse cutaneous responses to long-term percutaneously implanted rods surrounded by sphere-templated porous biomaterials engineered to mimic medical devices surrounded by a porous cuff. We hypothesized that keratinocytes would migrate through the pores and stop, permigrate, or marsupialize along the porous/solid interface. Porous/solid-core poly(2-hydroxyethyl methacrylate) [poly(HEMA)] and silicone rods were implanted in mice for 14 days, 1 month, 3 months, and 6 months. Implants with surrounding tissue were analyzed (immuno)histochemically by light microscopy. Poly(HEMA)/skin implants yielded better morphologic data than silicone implants. Keratinocytes at the poly(HEMA) interface migrated in two different directions. “Ventral” keratinocytes contiguous with the dermal-epidermal junction migrated into the outermost pores, forming an integrated collar surrounding the rods. ”Dorsal” keratinocytes appearing to emanate from the differentiated epithelial layer, extended upward along and into the exterior portion of the rod, forming an integrated sheath. Leukocytes persisted in poly(HEMA) and silicone pores for the duration of the study. Vascular and collagen networks within the poly(HEMA) pores matured as a function of time up to 3 months implantation. Nerves were not observed within the pores. Poly(HEMA) underwent morphological changes by 6 months of implantation. Marsupialization, foreign body encapsulation and infection were not observed in any implants. PMID:22359383

Sensitivity and Uncertainty Analysis of Plutonium and Cesium Isotopes in Modeling of BR3 Reactor Spent Fuel

DOE PAGES

Conant, Andrew; Erickson, Anna; Robel, Martin; ...

2017-02-03

Nuclear forensics has a broad task to characterize recovered nuclear or radiological material and interpret the results of investigation. One approach to isotopic characterization of nuclear material obtained from a reactor is to chemically separate and perform isotopic measurements on the sample and verify the results with modeling of the sample history, for example, operation of a nuclear reactor. The major actinide plutonium and fission product cesium are commonly measured signatures of the fuel history in a reactor core. This study investigates the uncertainty of the plutonium and cesium isotope ratios of a fuel rod discharged from a research pressurizedmore » water reactor when the location of the sample is not known a priori. A sensitivity analysis showed overpredicted values for the 240Pu/ 239Pu ratio toward the axial center of the rod and revealed a lower probability of the rod of interest (ROI) being on the periphery of the assembly. The uncertainty analysis found the relative errors due to only the rod position and boron concentration to be 17% to 36% and 7% to 15% for the 240Pu/ 239Pu and 137Cs/ 135Cs ratios, respectively. Lastly, this study provides a method for uncertainty quantification of isotope concentrations due to the location of the ROI. Similar analyses can be performed to verify future chemical and isotopic analyses.« less

Conformational changes accompany activation of reovirus RNA-dependent RNA transcription

PubMed Central

Mendez, Israel I.; Weiner, Scott G.; She, Yi-Min; Yeager, Mark; Coombs, Kevin M.

2009-01-01

Many critical biologic processes involve dynamic interactions between proteins and nucleic acids. Such dynamic processes are often difficult to delineate by conventional static methods. For example, while a variety of nucleic acid polymerase structures have been determined at atomic resolution, the details of how some multi-protein transcriptase complexes actively produce mRNA, as well as conformational changes associated with activation of such complexes, remain poorly understood. The mammalian reovirus innermost capsid (core) manifests all enzymatic activities necessary to produce mRNA from each of the 10 encased double-stranded RNA genes. We used rapid freezing and electron cryo-microscopy to trap and visualize transcriptionally active reovirus core particles and compared them to inactive core images. Rod-like density centered within actively transcribing core spike channels was attributed to exiting nascent mRNA. Comparative radial density plots of active and inactive core particles identified several structural changes in both internal and external regions of the icosahedral core capsid. Inactive and transcriptionally active cores were partially digested with trypsin and identities of initial tryptic peptides determined by mass spectrometry. Differentially-digested peptides, which also suggest transcription-associated conformational changes, were placed within the known 3-dimensional structures of major core proteins. PMID:18321727

Ultrastructural organization of connective tissue microfibrils in the posterior chamber of the eye in vivo and in vitro.

PubMed

Inoue, S

1995-02-01

The ultrastructural organization of connective tissue microfibrils was studied in the mouse eye and also by means of in vitro experiments for reconstituting microfibrils. In the posterior chamber of the eye of the C57BL/6J mouse, 3 nm-wide ribbon-like double-tracked structures were present and were periodically associated on either side with 3.5 nm-wide particulate structures identified as pentosomes, the subunits of amyloid P component (AP). At certain sites, such composite structures were observed in various stages of helical winding, and in these helices, pentosomes were preferentially localized internally. In helices in the final stages of winding, the resulting rods appeared increasingly similar to those of microfibrils. In experiments in vitro, incubation of chondroitin sulfate proteoglycan (CSPG) in TRIS buffer, pH 7.4, at 35 degrees C for 1 h produced random aggregates of 3 nm-wide double-tracked structures similar to those observed in the eye. Co-incubation of CSPG and AP resulted in the formation of rod-like structures arranged parallel to one another in approximately 50 nm-thick sheet-like layers. These rods were ultrastructurally similar to microfibrils and were made up of helically wound, 3 nm-wide double-tracked structures containing pentosomes within their core. The results of in vivo as well as in vitro experiments suggest the possibility that the connective tissue microfibril is composed of helically wound, CSPG-containing, 3 nm-wide double-tracked structures periodically associated with pentosomes which, as the helix becomes progressively tighter, fit with one another at the core of the helix to form successive 8.5 nm-wide disks of AP segments.

LMFBR fuel assembly design for HCDA fuel dispersal

DOEpatents

Lacko, Robert E.; Tilbrook, Roger W.

1984-01-01

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

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.

Nb 3Sn RRP® strand and Rutherford cable development for a 15 T dipole demonstrator

DOE PAGES

Barzi, E.; Andreev, N.; Li, P.; ...

2016-03-16

Keystoned Rutherford cables made of 28 strands and with a stainless steel core were developed and manufactured using 1 mm Nb3Sn composite wires produced by Oxford Superconducting Technology with 127 and 169 restacks using the Restacked-Rod-Process ®. Furthermore, the performance and properties of these cables were studied to evaluate possible candidates for 15 T accelerator magnets.

Development of control system of coating of rod hydraulic cylinders

NASA Astrophysics Data System (ADS)

Aizhambaeva, S. Zh; Maximova, A. V.

2018-01-01

In this article, requirements to materials of hydraulic cylinders and methods of eliminating the main factors affecting the quality of the applied coatings rod hydraulic cylinders. The chromium plating process - one of ways of increase of anti-friction properties of coatings rods, stability to the wear and corrosion. The article gives description of differences of the stand-speed chromium plating process from other types of chromium plating that determines a conclusion about cutting time of chromium plating process. Conducting the analysis of technological equipment suggested addressing the modernization of high-speed chromium plating processes by automation and mechanization. Control system developed by design of schematic block diagram of a modernized and stand-speed chromium plating process.

A new mouse model for stationary night blindness with mutant Slc24a1 explains the pathophysiology of the associated human disease

PubMed Central

Vinberg, Frans; Wang, Tian; Molday, Robert S.; Chen, Jeannie; Kefalov, Vladimir J.

2015-01-01

Mutations that affect calcium homeostasis (Ca2+) in rod photoreceptors are linked to retinal degeneration and visual disorders such as retinitis pigmentosa and congenital stationary night blindness (CSNB). It is thought that the concentration of Ca2+ in rod outer segments is controlled by a dynamic balance between influx via cGMP-gated (CNG) channels and extrusion via Na+/Ca2+, K+ exchangers (NCKX1). The extrusion-driven lowering of rod [Ca2+]i following light exposure controls their light adaptation and response termination. Mutant NCKX1 has been linked to autosomal-recessive stationary night blindness. However, whether NCKX1 contributes to light adaptation has not been directly tested and the mechanisms by which human NCKX1 mutations cause night blindness are not understood. Here, we report that the deletion of NCKX1 in mice results in malformed outer segment disks, suppressed expression and function of rod CNG channels and a subsequent 100-fold reduction in rod responses, while preserving normal cone responses. The compensating loss of CNG channel function in the absence of NCKX1-mediated Ca2+ extrusion may prevent toxic Ca2+ buildup and provides an explanation for the stationary nature of the associated disorder in humans. Surprisingly, the lack of NCKX1 did not compromise rod background light adaptation, suggesting additional Ca2+-extruding mechanisms exist in these cells. PMID:26246500

Design principles of a simple and safe 200-MW(thermal) nuclear district heating plant

DOE Office of Scientific and Technical Information (OSTI.GOV)

Goetzmann, C.; Bittermann, D.; Gobel, A.

Kraftwerk Union AG has almost completed the development of a dedicated 200-MW(thermal) nuclear district heating plant to provide environmentally clean energy at a predictably low cost. The concept can easily be adapted to meet power requirements within the 100- to 500-MW(thermal) range. This technology is the product of the experience gained with large pressurized water reactor and boiling water reactor power plants, with respect to both plant and fuel performance. The major development task is that of achieving sufficiently low capital cost by tailoring components and systems designed for large plants to the specific requirements of district heating. These requirementsmore » are small absolute power, low temperatures and pressures, and modest load following, all of which result in the characteristics that are summarized. A fully integrated primary system with natural circulation permits a very compact reactor building containing all safety-related systems and components. Plant safety is essentially guaranteed by inherent features. The reactor containment is tightly fitted around the reactor pressure vessel in such a way that, in the event of any postulated coolant leak, the core cannot become uncovered, even temporarily. Shutdown is assured by gravity drop of the control rods mounted above the core. Decay heat is removed from the core by means of natural circulation via dedicated intermediate circuits of external aircoolers.« less

Steady-State Thermal-Hydraulics Analyses for the Conversion of the BR2 Reactor to LEU

DOE Office of Scientific and Technical Information (OSTI.GOV)

Licht, J. R.; Bergeron, A.; Dionne, B.

2015-12-01

BR2 is a research reactor used for radioisotope production and materials testing. It’s a tank-in-pool type reactor cooled by light water and moderated by beryllium and light water (Figure 1). The reactor core consists of a beryllium moderator forming a matrix of 79 hexagonal prisms in a hyperboloid configuration; each having a central bore that can contain a variety of different components such as a fuel assembly, a control or regulating rod, an experimental device, or a beryllium or aluminum plug. Based on a series of tests, the BR2 operation is currently limited to a maximum allowable heat flux ofmore » 470 W/cm2 to ensure fuel plate integrity during steady-state operation and after a loss-of-flow/loss-of-pressure accident.« less

Controlling the Mechanical Properties of Bulk Metallic Glasses by Superficial Dealloyed Layer

PubMed Central

Wang, Chaoyang; Li, Man; Zhu, Mo; Wang, Han; Qin, Chunling; Zhao, Weimin

2017-01-01

Cu50Zr45Al5 bulk metallic glass (BMG) presents high fracture strength. For improving its plasticity and controlling its mechanical properties, superficial dealloying of the BMG was performed. A composite structure containing an inner rod-shaped Cu-Zr-Al amorphous core with high strength and an outer dealloyed nanoporous layer with high energy absorption capacity was obtained. The microstructures and mechanical properties of the composites were studied in detail. It was found, for the first time, that the mechanical properties of Cu50Zr45Al5 BMG can be controlled by adjusting the width of the buffer deformation zone in the dealloyed layer, which can be easily manipulated with different dealloying times. As a result, the compressive strength, compressive strain, and energy absorption capacity of the BMGs can be effectively modulated from 0.9 to 1.5 GPa, from 2.9% to 4.7%, and from 29.1 to 40.2 MJ/m3, respectively. The paper may open a door for developing important engineering materials with regulable and comprehensive performances. PMID:29077072

Interfacial charge separation and recombination in InP and quasi-type II InP/CdS core/shell quantum dot-molecular acceptor complexes.

PubMed

Wu, Kaifeng; Song, Nianhui; Liu, Zheng; Zhu, Haiming; Rodríguez-Córdoba, William; Lian, Tianquan

2013-08-15

Recent studies of group II-VI colloidal semiconductor heterostuctures, such as CdSe/CdS core/shell quantum dots (QDs) or dot-in-rod nanorods, show that type II and quasi-type II band alignment can facilitate electron transfer and slow down charge recombination in QD-molecular electron acceptor complexes. To explore the general applicability of this wave function engineering approach for controlling charge transfer properties, we investigate exciton relaxation and dissociation dynamics in InP (a group III-V semiconductor) and InP/CdS core/shell (a heterostructure beween group III-V and II-VI semiconductors) QDs by transient absorption spectroscopy. We show that InP/CdS QDs exhibit a quasi-type II band alignment with the 1S electron delocalized throughout the core and shell and the 1S hole confined in the InP core. In InP-methylviologen (MV(2+)) complexes, excitons in the QD can be dissociated by ultrafast electron transfer to MV(2+) from the 1S electron level (with an average time constant of 11.4 ps) as well as 1P and higher electron levels (with a time constant of 0.39 ps), which is followed by charge recombination to regenerate the complex in its ground state (with an average time constant of 47.1 ns). In comparison, InP/CdS-MV(2+) complexes show similar ultrafast charge separation and 5-fold slower charge recombination rates, consistent with the quasi-type II band alignment in these heterostructures. This result demonstrates that wave function engineering in nanoheterostructures of group III-V and II-VI semiconductors provides a promising approach for optimizing their light harvesting and charge separation for solar energy conversion applications.

A study to compare the efficacy of polyether ether ketone rod device with titanium devices in posterior spinal fusion in a canine model.

PubMed

Wang, Nanxiang; Xie, Huanxin; Xi, Chunyang; Zhang, Han; Yan, Jinglong

2017-03-09

The benefits of posterior lumbar fusion surgery with orthotopic paraspinal muscle-pediculated bone flaps are well established. However, the problem of non-union due to mechanical support is not completely resolved. The aim of the study was to compare the efficacy of polyether ether ketone (PEEK) rod device with conventional titanium devices in the posterior lumbar fusion surgery with orthotopic paraspinal muscle-pediculated bone flaps. This was a randomized controlled study with an experimental animal model. Thirty-two mongrel dogs were randomly divided into two groups-control group (n = 16), which received the titanium device and the treatment group (n = 16), which received PEEK rods. The animals were sacrificed 8 or 16 weeks after surgery. Lumbar spines of dogs in both groups were removed, harvested, and assessed for radiographic, biomechanical, and histological changes. Results in the current study indicated that there was no significant difference in the lumbar spine of the control and treatment groups in terms of radiographic, manual palpation, and gross examination. However, certain parameters of biomechanical testing showed significant differences (p < 0.05) in stiffness and displacement, revealing a better fusion (treatment group showed decreased stiffness with decreased displacement) of the bone graft. Similarly, the histological analysis also revealed a significant fusion mass in both treatment and control groups (p < 0.05). These findings revealed that fixation using PEEK connecting rod could improve the union of the bone graft in the posterior lumbar spine fusion surgery compared with that of the titanium rod fixation.

Process-based tolerance assessment of connecting rod machining process

NASA Astrophysics Data System (ADS)

Sharma, G. V. S. S.; Rao, P. Srinivasa; Surendra Babu, B.

2016-06-01

Process tolerancing based on the process capability studies is the optimistic and pragmatic approach of determining the manufacturing process tolerances. On adopting the define-measure-analyze-improve-control approach, the process potential capability index ( C p) and the process performance capability index ( C pk) values of identified process characteristics of connecting rod machining process are achieved to be greater than the industry benchmark of 1.33, i.e., four sigma level. The tolerance chain diagram methodology is applied to the connecting rod in order to verify the manufacturing process tolerances at various operations of the connecting rod manufacturing process. This paper bridges the gap between the existing dimensional tolerances obtained via tolerance charting and process capability studies of the connecting rod component. Finally, the process tolerancing comparison has been done by adopting a tolerance capability expert software.

A preliminary investigation of shape memory alloys in the surgical correction of scoliosis.

PubMed

Sanders, J O; Sanders, A E; More, R; Ashman, R B

1993-09-15

Nitinol, a shape memory alloy, is flexible at low temperatures but retains its original shape when heated. This offers interesting possibilities for scoliosis correction. Of the shape memory alloys, nitinol is the most promising medically because of biocompatibility and the ability to control transition temperature. In vivo: Six goats with experimental scoliosis were instrumented with 6-mm nitinol rods. The rods were transformed, and the scoliosis corrected, in the awakened goats by 450-kHz radio frequency induction heating. The curves averaged 41 degrees before instrumentation, 33 degrees after instrumentation, and 11 degrees after rod transformation. The animals tolerated the heating without discomfort, neurologic injury, or evidence of thermal injury to the tissues or the spinal cord. In vitro: Nitinol rods were tested under both constant deflection and constant loading conditions and plotted temperature versus either force or displacement. The 6-mm rod generated forces of 200 N. The 9-mm rod generated up to 500 N. We safely coupled shape memory alloy transformation to the spine and corrected an experimental spinal deformity in awake animals. The forces generated can be estimated by the rod's curvature and temperature. The use of shape memory alloys allows continuous neurologic monitoring during awake correction, true rotational correction by rod torsion, and the potential option of periodic correction to take advantage of spinal viscoelasticity and the potential of true rotational correction by rod torsion.

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. Copyright © 2015 Elsevier Ltd. All rights reserved.

Mechatronic Aeropendulum: Demonstration of Linear and Nonlinear Feedback Control Principles with MATLAB/Simulink Real-Time Windows Target

ERIC Educational Resources Information Center

Enikov, E. T.; Campa, G.

2012-01-01

This paper presents a low-cost hands-on experiment for a classical undergraduate controls course for non-electrical engineering majors. The setup consists of a small dc electrical motor attached to one of the ends of a light rod. The motor drives a 2-in propeller and allows the rod to swing. Angular position is measured by a potentiometer attached…

Probabilistic analysis on the failure of reactivity control for the PWR

NASA Astrophysics Data System (ADS)

Sony Tjahyani, D. T.; Deswandri; Sunaryo, G. R.

2018-02-01

The fundamental safety function of the power reactor is to control reactivity, to remove heat from the reactor, and to confine radioactive material. The safety analysis is used to ensure that each parameter is fulfilled during the design and is done by deterministic and probabilistic method. The analysis of reactivity control is important to be done because it will affect the other of fundamental safety functions. The purpose of this research is to determine the failure probability of the reactivity control and its failure contribution on a PWR design. The analysis is carried out by determining intermediate events, which cause the failure of reactivity control. Furthermore, the basic event is determined by deductive method using the fault tree analysis. The AP1000 is used as the object of research. The probability data of component failure or human error, which is used in the analysis, is collected from IAEA, Westinghouse, NRC and other published documents. The results show that there are six intermediate events, which can cause the failure of the reactivity control. These intermediate events are uncontrolled rod bank withdrawal at low power or full power, malfunction of boron dilution, misalignment of control rod withdrawal, malfunction of improper position of fuel assembly and ejection of control rod. The failure probability of reactivity control is 1.49E-03 per year. The causes of failures which are affected by human factor are boron dilution, misalignment of control rod withdrawal and malfunction of improper position for fuel assembly. Based on the assessment, it is concluded that the failure probability of reactivity control on the PWR is still within the IAEA criteria.

Weld Nugget Temperature Control in Thermal Stir Welding

NASA Technical Reports Server (NTRS)

Ding, R. Jeffrey (Inventor)

2014-01-01

A control system for a thermal stir welding system is provided. The control system includes a sensor and a controller. The sensor is coupled to the welding system's containment plate assembly and generates signals indicative of temperature of a region adjacent and parallel to the welding system's stir rod. The controller is coupled to the sensor and generates at least one control signal using the sensor signals indicative of temperature. The controller is also coupled to the welding system such that at least one of rotational speed of the stir rod, heat supplied by the welding system's induction heater, and feed speed of the welding system's weld material feeder are controlled based on the control signal(s).

Cell fiber-based three-dimensional culture system for highly efficient expansion of human induced pluripotent stem cells.

PubMed

Ikeda, Kazuhiro; Nagata, Shogo; Okitsu, Teru; Takeuchi, Shoji

2017-06-06

Human pluripotent stem cells are a potentially powerful cellular resource for application in regenerative medicine. Because such applications require large numbers of human pluripotent stem cell-derived cells, a scalable culture system of human pluripotent stem cell needs to be developed. Several suspension culture systems for human pluripotent stem cell expansion exist; however, it is difficult to control the thickness of cell aggregations in these systems, leading to increased cell death likely caused by limited diffusion of gases and nutrients into the aggregations. Here, we describe a scalable culture system using the cell fiber technology for the expansion of human induced pluripotent stem (iPS) cells. The cells were encapsulated and cultured within the core region of core-shell hydrogel microfibers, resulting in the formation of rod-shaped or fiber-shaped cell aggregations with sustained thickness and high viability. By encapsulating the cells with type I collagen, we demonstrated a long-term culture of the cells by serial passaging at a high expansion rate (14-fold in four days) while retaining its pluripotency. Therefore, our culture system could be used for large-scale expansion of human pluripotent stem cells for use in regenerative medicine.

Ordering of Glass Rods in Nematic and Cholesteric Liquid Crystals

DTIC Science & Technology

2011-12-01

3), 483–508 (2007). 2. M. D. Lynch and D. L. Patrick, “Controlling the orientation of micron-sized rod-shaped SiC particles with nematic liquid...Elastic torque and the levitation of metal wires by a nematic liquid crystal,” Science 303(5658), 652–655 (2004). 17. R. Eelkema, M. M. Pollard, J...Building Blocks for Iterative Methods, 2nd ed. (SIAM, 1994). 1. Introduction Incorporating rod-like particles into liquid crystal (LC) media can lead

Formation of gold nanorods by a stochastic "popcorn" mechanism.

PubMed

Edgar, Jonathan A; McDonagh, Andrew M; Cortie, Michael B

2012-02-28

Gold nanorods have significant technological potential and are of broad interest to the nanotechnology community. The discovery of the seeded, wet-chemical synthetic process to produce them may be regarded as a landmark in the control of metal nanoparticle shape. However, the mechanism by which the initial spherical gold seeds acquire anisotropy is a critical, yet poorly understood, factor. Here we examine the very early stages of rod growth using a combination of techniques including cryogenic transmission electron microscopy, optical spectroscopy, and computational modeling. Reconciliation of the available experimental observations can only be achieved by invoking a stochastic, "popcorn"-like mechanism of growth, in which individual seeds lie quiescent for some time before suddenly and rapidly growing into rods. This is quite different from the steady, concurrent growth of nanorods that has been previously generally assumed. Furthermore we propose that the shape is controlled by the ratio of surface energy of rod sides to rod ends, with values of this quantity in the range of 0.3-0.8 indicated for typical growth solutions.

An analytical approach to thermal modeling of Bridgman-type crystal growth. I - One-dimensional analysis

NASA Technical Reports Server (NTRS)

Naumann, R. J.

1982-01-01

A relatively simple one-dimensional thermal model of the Bridgman growth process has been developed which is applicable to the growth of small diameter samples with conductivities similar to those of metallic alloys. The heat flow in a translating rod is analyzed in a way that is applicable to Biot numbers less than unity. The model accommodates an adiabatic zone, different heat transfer coefficients in the hot and cold zones, and changes in sample material properties associated with phase change. The analysis is applied to several simplified cases. The effect of the rod's motion is studied in a three-zone furnace for a rod sufficiently long that end effects can be neglected; end effects are then investigated for a motionless rod. Finally, the addition of a fourth zone, an independently controlled booster heater between the main heater and the adiabatic zone, is evaluated for its ability to increase the gradient in the sample at the melt interface and to control the position of the interface.

Samd7 is a cell type-specific PRC1 component essential for establishing retinal rod photoreceptor identity

PubMed Central

Omori, Yoshihiro; Kubo, Shun; Kon, Tetsuo; Furuhashi, Mayu; Narita, Hirotaka; Kominami, Taro; Ueno, Akiko; Tsutsumi, Ryotaro; Chaya, Taro; Yamamoto, Haruka; Suetake, Isao; Ueno, Shinji; Koseki, Haruhiko; Furukawa, Takahisa

2017-01-01

Precise transcriptional regulation controlled by a transcription factor network is known to be crucial for establishing correct neuronal cell identities and functions in the CNS. In the retina, the expression of various cone and rod photoreceptor cell genes is regulated by multiple transcription factors; however, the role of epigenetic regulation in photoreceptor cell gene expression has been poorly understood. Here, we found that Samd7, a rod-enriched sterile alpha domain (SAM) domain protein, is essential for silencing nonrod gene expression through H3K27me3 regulation in rod photoreceptor cells. Samd7-null mutant mice showed ectopic expression of nonrod genes including S-opsin in rod photoreceptor cells and rod photoreceptor cell dysfunction. Samd7 physically interacts with Polyhomeotic homologs (Phc proteins), components of the Polycomb repressive complex 1 (PRC1), and colocalizes with Phc2 and Ring1B in Polycomb bodies. ChIP assays showed a significant decrease of H3K27me3 in the genes up-regulated in the Samd7-deficient retina, showing that Samd7 deficiency causes the derepression of nonrod gene expression in rod photoreceptor cells. The current study suggests that Samd7 is a cell type-specific PRC1 component epigenetically defining rod photoreceptor cell identity. PMID:28900001

Samd7 is a cell type-specific PRC1 component essential for establishing retinal rod photoreceptor identity.

PubMed

Omori, Yoshihiro; Kubo, Shun; Kon, Tetsuo; Furuhashi, Mayu; Narita, Hirotaka; Kominami, Taro; Ueno, Akiko; Tsutsumi, Ryotaro; Chaya, Taro; Yamamoto, Haruka; Suetake, Isao; Ueno, Shinji; Koseki, Haruhiko; Nakagawa, Atsushi; Furukawa, Takahisa

2017-09-26

Precise transcriptional regulation controlled by a transcription factor network is known to be crucial for establishing correct neuronal cell identities and functions in the CNS. In the retina, the expression of various cone and rod photoreceptor cell genes is regulated by multiple transcription factors; however, the role of epigenetic regulation in photoreceptor cell gene expression has been poorly understood. Here, we found that Samd7, a rod-enriched sterile alpha domain (SAM) domain protein, is essential for silencing nonrod gene expression through H3K27me3 regulation in rod photoreceptor cells. Samd7- null mutant mice showed ectopic expression of nonrod genes including S-opsin in rod photoreceptor cells and rod photoreceptor cell dysfunction. Samd7 physically interacts with Polyhomeotic homologs (Phc proteins), components of the Polycomb repressive complex 1 (PRC1), and colocalizes with Phc2 and Ring1B in Polycomb bodies. ChIP assays showed a significant decrease of H3K27me3 in the genes up-regulated in the Samd7 -deficient retina, showing that Samd7 deficiency causes the derepression of nonrod gene expression in rod photoreceptor cells. The current study suggests that Samd7 is a cell type-specific PRC1 component epigenetically defining rod photoreceptor cell identity.

Etude des performances de solveurs deterministes sur un coeur rapide a caloporteur sodium

NASA Astrophysics Data System (ADS)

Bay, Charlotte

The reactors of next generation, in particular SFR model, represent a true challenge for current codes and solvers, used mainly for thermic cores. There is no guarantee that their competences could be straight adapted to fast neutron spectrum, or to major design differences. Thus it is necessary to assess the validity of solvers and their potential shortfall in the case of fast neutron reactors. As part of an internship with CEA (France), and at the instigation of EPM Nuclear Institute, this study concerns the following codes : DRAGON/DONJON, ERANOS, PARIS and APOLLO3. The precision assessment has been performed using Monte Carlo code TRIPOLI4. Only core calculation was of interest, namely numerical methods competences in precision and rapidity. Lattice code was not part of the study, that is to say nuclear data, self-shielding, or isotopic compositions. Nor was tackled burnup or time evolution effects. The study consists in two main steps : first evaluating the sensitivity of each solver to calculation parameters, and obtain its optimal calculation set ; then compare their competences in terms of precision and rapidity, by collecting usual quantities (effective multiplication factor, reaction rates map), but also more specific quantities which are crucial to the SFR design, namely control rod worth and sodium void effect. The calculation time is also a key factor. Whatever conclusion or recommendation that could be drawn from this study, they must first of all be applied within similar frameworks, that is to say small fast neutron cores with hexagonal geometry. Eventual adjustments for big cores will have to be demonstrated in developments of this study.

The energy of naturally curved elastic rods with an application to the stretching and contraction of a free helical spring as a model for DNA

NASA Astrophysics Data System (ADS)

Manning, Gerald S.

2015-09-01

We give a contemporary and direct derivation of a classical, but insufficiently familiar, result in the theory of linear elasticity—a representation for the energy of a stressed elastic rod with central axis that intrinsically takes the shape of a general space curve. We show that the geometric torsion of the space curve, while playing a crucial role in the bending energy, is physically unrelated to the elastic twist. We prove that the twist energy vanishes in the lowest-energy states of a rod subject to constraints that do not restrict the twist. The stretching and contraction energies of a free helical spring are computed. There are local high-energy minima. We show the possibility of using the spring to model the chirality of DNA. We then compare our results with an available atomic level energy simulation that was performed on DNA unconstrained in the same sense as the free spring. We find some possible reflections of springlike behavior in the mechanics of DNA, but, unsurprisingly, the base pairs lend a material substance to the core of DNA that a spring does not capture.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

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

1997-08-01

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

The energy of naturally curved elastic rods with an application to the stretching and contraction of a free helical spring as a model for DNA.

PubMed

Manning, Gerald S

2015-09-14

We give a contemporary and direct derivation of a classical, but insufficiently familiar, result in the theory of linear elasticity-a representation for the energy of a stressed elastic rod with central axis that intrinsically takes the shape of a general space curve. We show that the geometric torsion of the space curve, while playing a crucial role in the bending energy, is physically unrelated to the elastic twist. We prove that the twist energy vanishes in the lowest-energy states of a rod subject to constraints that do not restrict the twist. The stretching and contraction energies of a free helical spring are computed. There are local high-energy minima. We show the possibility of using the spring to model the chirality of DNA. We then compare our results with an available atomic level energy simulation that was performed on DNA unconstrained in the same sense as the free spring. We find some possible reflections of springlike behavior in the mechanics of DNA, but, unsurprisingly, the base pairs lend a material substance to the core of DNA that a spring does not capture.

Calculation of Heat-Bearing Agent’s Steady Flow in Fuel Bundle

NASA Astrophysics Data System (ADS)

Amosova, E. V.; Guba, G. G.

2017-11-01

This paper introduces the result of studying the heat exchange in the fuel bundle of the nuclear reactor’s fuel magazine. The article considers the fuel bundle of the infinite number of fuel elements, fuel elements are considered in the checkerboard fashion (at the tops of a regular triangle a fuel element is a plain round rod. The inhomogeneity of volume energy release in the rod forms the inhomogeneity of temperature and velocity fields, and pressure. Computational methods for studying hydrodynamics in magazines and cores with rod-shape fuel elements are based on a significant simplification of the problem: using basic (averaged) equations, isobaric section hypothesis, porous body model, etc. This could be explained by the complexity of math description of the three-dimensional fluid flow in the multi-connected area with the transfer coefficient anisotropy, curved boundaries and technical computation difficulties. Thus, calculative studying suggests itself as promising and important. There was developed a method for calculating the heat-mass exchange processes of inter-channel fuel element motions, which allows considering the contribution of natural convection to the heat-mass exchange based on the Navier-Stokes equations and Boussinesq approximation.

The energy of naturally curved elastic rods with an application to the stretching and contraction of a free helical spring as a model for DNA

DOE Office of Scientific and Technical Information (OSTI.GOV)

Manning, Gerald S., E-mail: jerrymanning@rcn.com

We give a contemporary and direct derivation of a classical, but insufficiently familiar, result in the theory of linear elasticity—a representation for the energy of a stressed elastic rod with central axis that intrinsically takes the shape of a general space curve. We show that the geometric torsion of the space curve, while playing a crucial role in the bending energy, is physically unrelated to the elastic twist. We prove that the twist energy vanishes in the lowest-energy states of a rod subject to constraints that do not restrict the twist. The stretching and contraction energies of a free helicalmore » spring are computed. There are local high-energy minima. We show the possibility of using the spring to model the chirality of DNA. We then compare our results with an available atomic level energy simulation that was performed on DNA unconstrained in the same sense as the free spring. We find some possible reflections of springlike behavior in the mechanics of DNA, but, unsurprisingly, the base pairs lend a material substance to the core of DNA that a spring does not capture.« less

METHOD FOR SENSING DEGREE OF FLUIDIZATION IN FLUIDIZED BED

DOEpatents

Levey, R.P. Jr.; Fowler, A.H.

1961-12-12

A method is given for detecting, indicating, and controlling the degree of fluidization in a fluid-bed reactor into which powdered material is fed. The method comprises admitting of gas into the reactor, inserting a springsupported rod into the powder bed of the reactor, exciting the rod to vibrate at its resonant frequency, deriving a signal responsive to the amplitude of vibi-ation of the rod and spring, the signal being directiy proportional to the rate of flow of the gas through the reactor, displaying the signal to provide an indication of the degree of fluidization within the reactor, and controlling the rate of gas flow into the reactor until said signal stabilizes at a constant value to provide substantially complete fluidization within the reactor. (AEC)

NEUTRONIC REACTOR MANIPULATING DEVICE

DOEpatents

Ohlinger, L.A.

1962-08-01

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

Measuring the efficiency of control rods in the RBMK critical assembly using a model of RKI-1 reactimeter

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zhitarev, V. E., E-mail: vejitarev@yandex.ru; Lebedev, G. V.; Sergevnin, A. Yu.

2016-12-15

The efficiency of control rods of the RBMK critical assembly is measured in a series of experiments. The aim of measurements is to determine the characteristics of the model of an RKI-1 reactimeter. The RKI-1 reactimeter is intended for measuring the efficiency of control rods when, according to conditions of operation, the metrological certification of results of an experiment is required. Complications with the metrological certification of reactimeters arise owing to the fact that usually calculated corrections to the results of measurements are required. When the RKI-1 reactimeter is used, there is no need to introduce calculated corrections; the resultmore » of measurements is given with the indication of substantiated errors. In connection with this, the metrological certification of the results of measurements using the RKI-1 reactimeter is simplified.« less

Choice of mathematical models for technological process of glass rod drawing

NASA Astrophysics Data System (ADS)

Alekseeva, L. B.

2017-10-01

The technological process of drawing glass rods (light guides) is considered. Automated control of the drawing process is reduced to the process of making decisions to ensure a given quality. The drawing process is considered as a control object, including the drawing device (control device) and the optical fiber forming zone (control object). To study the processes occurring in the formation zone, mathematical models are proposed, based on the continuum mechanics basics. To assess the influence of disturbances, a transfer function is obtained from the basis of the wave equation. Obtaining the regression equation also adequately describes the drawing process.

Association of anatase (TiO2) and microbes: Unusual fossilization effect or a potential biosignature?

USGS Publications Warehouse

Glamoclija, M.; Steele, A.; Fries, M.; Schieber, J.; Voytek, M.A.; Cockell, C.S.

2009-01-01

We combined microbial paleontology and molecular biology methods to study the Eyreville B drill core from the 35.3-Ma-old Chesapeake Bay impact structure, Virginia, USA. The investigated sample is a pyrite vein collected from the 1353.81- 1353.89 m depth interval, located within a section of biotite granite. The granite is a pre-impact rock that was disrupted by the impact event. A search for inorganic (mineral) biosignatures revealed the presence of micron-size rod morphologies of anatase (TiO2) embedded in chlorite coatings on pyrite grains. Neither the Acridine Orange microbial probe nor deoxyribonucleic acid (DNA) extraction followed by polymerase chain reaction (PCR) amplifi cation showed the presence of DNA or ribonucleic acid (RNA) at the location of anatase rods, implying the absence of viable cells in the investigated area. A Nile Red microbial probe revealed the presence of lipids in the rods. Because most of the lipids are resistant over geologic time spans, they are good biomarkers, and they are an indicator of biogenicity for these possibly 35-Ma-old microbial fossils. The mineral assemblage suggests that rod morphologies are associated with low-temperature (<100 ??C) hydrothermal alteration that involved aqueous fluids. The temporal constraints on the anatase fossils are still uncertain because pre-impact alteration of the granite and postimpact heating may have provided identical conditions for anatase precipitation and microbial preservation. ?? 2009 The Geological Society of America.

CNGA3 deficiency affects cone synaptic terminal structure and function and leads to secondary rod dysfunction and degeneration.

PubMed

Xu, Jianhua; Morris, Lynsie M; Michalakis, Stylianos; Biel, Martin; Fliesler, Steven J; Sherry, David M; Ding, Xi-Qin

2012-03-01

To investigate rod function and survival after cone dysfunction and degeneration in a mouse model of cone cyclic nucleotide-gated (CNG) channel deficiency. Rod function and survival in mice with cone CNG channel subunit CNGA3 deficiency (CNGA3-/- mice) were evaluated by electroretinographic (ERG), morphometric, and Western blot analyses. The arrangement, integrity, and ultrastructure of photoreceptor terminals were investigated by immunohistochemistry and electron microscopy. The authors found loss of cone function and cone death accompanied by impairment of rods and rod-driven signaling in CNGA3-/- mice. Scotopic ERG b-wave amplitudes were reduced by 15% at 1 month, 30% at 6 months, and 40% at 9 months and older, while scotopic a-wave amplitudes were decreased by 20% at 9 months, compared with ERGs of age-matched wild-type mice. Outer nuclear layer thickness in CNGA3-/- retina was reduced by 15% at 12 months compared with age-matched wild-type controls. This was accompanied by a 30%-40% reduction in expression of rod-specific proteins, including rhodopsin, rod transducin α-subunit, and glutamic acid-rich protein (GARP). Cone terminals in the CNGA3-/- retina showed a progressive loss of neurochemical and ultrastructural integrity. Abnormalities were observed as early as 1 month. Disorganized rod terminal ultrastructure was noted by 12 months. These findings demonstrate secondary rod impairment and degeneration after cone degeneration in mice with cone CNG channel deficiency. Loss of cone phototransduction accompanies the compromised integrity of cone terminals. With time, rod synaptic structure, function, and viability also become compromised.

Nutrient resuscitation and growth of starved cells in sandstone cores: a novel approach to enhanced oil recovery. [Klebsiella pneumoniae

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lappin-Scott, H.M.; Cusack, F.; Costerton, J.W.

1988-06-01

Klebsiella pneumoniae, which was reduced in size (0.25 by 0.5 ..mu..m) by carbon deprivation, was injected into a series of sandstone cores and subjected to separate treatments. Scanning electron microscopy of 400-mD cores showed these small starved cells in nearly every core section. The cells were a mixture of small rods and cocci with little or no biofilm production. Continuous or dose stimulation with sodium citrate allowed the cells to grow throughout the sandstone and completely plug the length of the core. The resuscitated cells were larger than the starved cells (up to 1.7 ..mu..m) and were encased in glycocalyx.more » Scanning electron microscopic results of resuscitation in situ with half-strength brain heart infusion broth showed that a shallow skin plug of cells formed at the core inlet and that fewer cells were located in the lower sections. Starved cells also penetrated 200-mD cores and were successfully resuscitated in situ with sodium citrate, so that the entire core was plugged. Nutrient resuscitation of injected starved cells to produce full-size cells which grow and block the rock pores may be successfully applied to selective plugging and may effectively increase oil recovery.« less

Photonic crystal fiber technology for compact fiber-delivered high-power ultrafast fiber lasers

NASA Astrophysics Data System (ADS)

Triches, Marco; Michieletto, Mattia; Johansen, Mette M.; Jakobsen, Christian; Olesen, Anders S.; Papior, Sidsel R.; Kristensen, Torben; Bondue, Magalie; Weirich, Johannes; Alkeskjold, Thomas T.

2018-02-01

Photonic crystal fiber (PCF) technology has radically impacted the scientific and industrial ultrafast laser market. Reducing platform dimensions are important to decrease cost and footprint while maintaining high optical efficiency. We present our recent work on short 85 μm core ROD-type fiber amplifiers that maintain single-mode performance and excellent beam quality. Robust long-term performance at 100 W average power and 250 kW peak power in 20 ps pulses at 1030 nm wavelength is presented, exceeding 500 h with stable performance in terms of both polarization and power. In addition, we present our recent results on hollow-core ultrafast fiber delivery maintaining high beam quality and polarization purity.

Fabrication of a magnetic helical mesostructured silica rod

NASA Astrophysics Data System (ADS)

Zhang, Lei; Zhang Qiao, Shi; Cheng, Lina; Yan, Zifeng; Qing Lu, Gao Max

2008-10-01

We report a one-step synthesis of magnetic helical mesostructured silica (MHMS) by self-assembly of an achiral surfactant, magnetic nanocrystals with stearic acid ligands and silicate. This core-shell structured material consists of an Fe3O4 superparamagnetic nanocrystal core and a highly ordered periodic helical mesoporous silica shell. We propose that the formation of the helical structure is induced by the interaction between the surfactant and dissociated stearic acid ligands. The MHMS obtained possesses superparamagnetism, uniform mesostructure, narrow pore size distribution, high surface area, and large pore volume. Furthermore, the drug release process is demonstrated using aspirin as a drug model and MHMS as a drug carrier in a sodium phosphate buffer solution.

Analysis of the Gas Core Actinide Transmutation Reactor (GCATR)

NASA Technical Reports Server (NTRS)

Clement, J. D.; Rust, J. H.

1977-01-01

Design power plant studies were carried out for two applications of the plasma core reactor: (1) As a breeder reactor, (2) As a reactor able to transmute actinides effectively. In addition to the above applications the reactor produced electrical power with a high efficiency. A reactor subsystem was designed for each of the two applications. For the breeder reactor, neutronics calculations were carried out for a U-233 plasma core with a molten salt breeding blanket. A reactor was designed with a low critical mass (less than a few hundred kilograms U-233) and a breeding ratio of 1.01. The plasma core actinide transmutation reactor was designed to transmute the nuclear waste from conventional LWR's. The spent fuel is reprocessed during which 100% of Np, Am, Cm, and higher actinides are separated from the other components. These actinides are then manufactured as oxides into zirconium clad fuel rods and charged as fuel assemblies in the reflector region of the plasma core actinide transmutation reactor. In the equilibrium cycle, about 7% of the actinides are directly fissioned away, while about 31% are removed by reprocessing.

46 CFR 32.65-20 - Pumprooms-TB/ALL.

Code of Federal Regulations, 2011 CFR

2011-10-01

... bulkhead between the pumproom and the pump-engine compartment may be pierced by fixed lights, drive shaft and pump-engine control rods, provided that the shafts and rods are fitted with stuffing boxes where... their cargo pumps isolated from all sources of vapor ignition by gastight bulkheads. Totally enclosed...

46 CFR 32.65-20 - Pumprooms-TB/ALL.

Code of Federal Regulations, 2013 CFR

2013-10-01

... bulkhead between the pumproom and the pump-engine compartment may be pierced by fixed lights, drive shaft and pump-engine control rods, provided that the shafts and rods are fitted with stuffing boxes where... their cargo pumps isolated from all sources of vapor ignition by gastight bulkheads. Totally enclosed...

46 CFR 32.65-20 - Pumprooms-TB/ALL.

Code of Federal Regulations, 2010 CFR

2010-10-01

... bulkhead between the pumproom and the pump-engine compartment may be pierced by fixed lights, drive shaft and pump-engine control rods, provided that the shafts and rods are fitted with stuffing boxes where... their cargo pumps isolated from all sources of vapor ignition by gastight bulkheads. Totally enclosed...

46 CFR 32.65-20 - Pumprooms-TB/ALL.

Code of Federal Regulations, 2012 CFR

2012-10-01

... bulkhead between the pumproom and the pump-engine compartment may be pierced by fixed lights, drive shaft and pump-engine control rods, provided that the shafts and rods are fitted with stuffing boxes where... their cargo pumps isolated from all sources of vapor ignition by gastight bulkheads. Totally enclosed...

46 CFR 32.65-20 - Pumprooms-TB/ALL.

Code of Federal Regulations, 2014 CFR

2014-10-01

... bulkhead between the pumproom and the pump-engine compartment may be pierced by fixed lights, drive shaft and pump-engine control rods, provided that the shafts and rods are fitted with stuffing boxes where... their cargo pumps isolated from all sources of vapor ignition by gastight bulkheads. Totally enclosed...

The actin-activated ATPase of co-polymer filaments of myosin and myosin-rod.

PubMed Central

Stepkowski, D; Orlova, A A; Moos, C

1994-01-01

The actin activated ATPase of myosin at low ionic strength shows a complex dependence on actin concentration, in contrast with the simple hyperbolic actin activation kinetics of heavy meromyosin and subfragment-1. To investigate how the aggregation of myosin influences the actomyosin ATPase kinetics, we have studied the actin-activated ATPase of mixed filaments in which the myosin molecules are separated from each other by copolymerization with myosin rod. Electron microscopy of copolymer filaments, alone and bound to actin, indicates that the myosin heads are distributed randomly along the co-polymer filaments. The actin-activated ATPase of myosin decreases with increasing rod, approaching a plateau of about 30% of the control at a rod/myosin molar ratio of 4:1. The decrease in ATPase persists even at Vmax, the extrapolated limit at infinite actin, indicating that it is not due merely to the loss of cooperative actin binding. Furthermore, the actin dependence of the ATPase still shows a biphasic character like that of control myosin, even at rod/myosin ratio of 12:1, so this complexity is not probably due solely to the structural proximity of myosin molecules, but may involve a non-equivalence of myosin heads or myosin molecules in the filament environment. Images Figure 1 Figure 2 PMID:8198528

Role of recoverin in rod photoreceptor light adaptation.

PubMed

Morshedian, Ala; Woodruff, Michael L; Fain, Gordon L

2018-04-15

Recoverin is a small molecular-weight, calcium-binding protein in rod outer segments that can modulate the rate of rhodopsin phosphorylation. We describe two additional and perhaps more important functions during photoreceptor light adaptation. Recoverin influences the rate of change of adaptation. In wild-type rods, sensitivity and response integration time adapt with similar time constants of 150-200 ms. In Rv-/- rods lacking recoverin, sensitivity declines faster and integration time is already shorter and not significantly altered. During steady light exposure, rod circulating current slowly increases during a time course of tens of seconds, gradually extending the operating range of the rod. In Rv-/- rods, this mechanism is deleted, steady-state currents are already larger and rods saturate at brighter intensities. We propose that recoverin modulates spontaneous and light-activated phophodiesterase-6, the phototransduction effector enzyme, to increase sensitivity in dim light but improve responsiveness to change in brighter illumination. Recoverin is a small molecular-weight, calcium-binding protein in rod outer segments that binds to G-protein receptor kinase 1 and can alter the rate of rhodopsin phosphorylation. A change in phosphorylation should change the lifetime of light-activated rhodopsin and the gain of phototransduction, but deletion of recoverin has little effect on the sensitivity of rods either in the dark or in dim-to-moderate background light. We describe two additional functions perhaps of greater physiological significance. (i) When the ambient intensity increases, sensitivity and integration time decrease in wild-type (WT) rods with similar time constants of 150-200 ms. Recoverin is part of the mechanism controlling this process because, in Rv-/- rods lacking recoverin, sensitivity declines more rapidly and integration time is already shorter and not further altered. (ii) During steady light exposure, WT rod circulating current slowly increases during a time course of tens of seconds, gradually extending the operating range of the rod. In Rv-/- rods, this mechanism is also deleted, steady-state currents are already larger and rods saturate at brighter intensities. We argue that neither (i) nor (ii) can be caused by modulation of rhodopsin phosphorylation but may instead be produced by direct modulation of phophodiesterase-6 (PDE6), the phototransduction effector enzyme. We propose that recoverin in dark-adapted rods keeps the integration time long and the spontaneous PDE6 rate relatively high to improve sensitivity. In background light, the integration time is decreased to facilitate detection of change and motion and the spontaneous PDE6 rate decreases to augment the rod working range. © 2018 The Authors. The Journal of Physiology © 2018 The Physiological Society.

Robust technology and system for management of sucker rod pumping units in oil wells

NASA Astrophysics Data System (ADS)

Aliev, T. A.; Rzayev, A. H.; Guluyev, G. A.; Alizada, T. A.; Rzayeva, N. E.

2018-01-01

We propose a technology for calculating the robust, normalized correlation functions of the signal from the force sensor on the rod string attached to the hanger of the sucker rod pumping unit. The robust normalized correlation functions are used to form sets of informative attribute combinations, each of which corresponds to a technical condition of the sucker rod pumping unit. We demonstrate how these sets can be used to solve identification and management problems in the oil production process in real time using inexpensive controllers. The results obtained from using the system on real objects are also presented in this paper. It was determined that the energy saved and prolonged overhaul period substantially increased the cost-effectiveness.

Clip gage attachment for frictionless measurement of displacement during high-temperature mechanical testing

DOEpatents

Alexander, David J.

1994-01-01

An attachment for placement between a test specimen and a remote clip gage extensometer providing improved fracture toughness tests of materials at elevated temperature. Using a cylindrical tube and axial rod in new relationship, the device transfers the displacement signal of the fracture toughness test specimen directly to a clip gage extensometer located outside the high temperature furnace. Virtually frictionless operation is assured by having the test specimen center one end of the rod in one end of the tube, while the clip gage extensometer arms center the other end of the rod in the other end of the tube. By providing positive control over both ends of both rod and tube, the attachment may be operated in orientations other than vertical.

Autofusion in the immature spine treated with growing rods.

PubMed

Cahill, Patrick J; Marvil, Sean; Cuddihy, Laury; Schutt, Corey; Idema, Jocelyn; Clements, David H; Antonacci, M Darryl; Asghar, Jahangir; Samdani, Amer F; Betz, Randal R

2010-10-15

Retrospective case review of skeletally immature patients treated with growing rods. Patients received an average of 9.6 years follow-up care. (1) to identify the rate of autofusion in the growing spine with the use of growing rods; (2) to quantify how much correction can be attained with definitive instrumented fusion after long-term treatment with growing rods; and (3) to describe the extent of Smith-Petersen osteotomies required to gain correction of an autofused spine following growing rod treatment. The safety and use of growing rods for curve correction and maintenance in the growing spine population has been established in published reports. While autofusion has been reported, the prevalence and sequelae are not known. Nine skeletally immature children with scoliosis were identified who had been treated using growing rods. A retrospective review of the medical records and radiographs was conducted and the following data collected: complications, pre- and postoperative Cobb angles at time of initial surgery (growing rod placement), pre- and postoperative Cobb angles at time of final surgery (growing rod removal and definitive fusion), total spine length as measured from T1-S1, % correction since initiation of treatment and at definitive fusion, total number of surgeries, and number of patients found to have autofusion at the time of device removal. The rate of autofusion in children treated with growing rods was 89%. The average percent of the Cobb angle correction obtained at definitive fusion was 44%. On average, 7 osteotomies per patient were required at the time of definitive fusion due to autofusion. Although growing rods have efficacy in the control of deformity within the growing spine, they also have adverse effects on the spine. Immature spines treated with a growing rod have high rates of unintended autofusion which can possibly lead to difficult and only moderate correction at the time of definitive fusion.

Casting technology for manufacturing metal rods from simulated metallic spent fuels

NASA Astrophysics Data System (ADS)

Leeand, Y. S.; Lee, D. B.; Kim, C. K.; Shin, Y. J.; Lee, J. H.

2000-09-01

A uranium metal rod 13.5 mm in diameter and 1,150 mm long was produced from simulated metallic spent fuels with advanced casting equipment using the directional-solidification method. A vacuum casting furnace equipped with a four-zone heater to prevent surface oxidation and the formation of surface shrinkage holes was designed. By controlling the axial temperature gradient of the casting furnace, deformation by the surface shrinkage phenomena was diminished, and a sound rod was manufactured. The cooling behavior of the molten uranium was analyzed using the computer software package MAGMAsoft.

High frequency magnetostrictive transducers for waveguide applications

DOE Office of Scientific and Technical Information (OSTI.GOV)

Daw, Joshua Earl; Taylor, Steven Cheney; Rempe, Joy Lynn

A high frequency magnetostrictive transducer includes a magnetostrictive rod or wire inserted co-axially into a driving coil, wherein the driving coil includes a coil arrangement with a plurality of small coil segments along the magnetostrictive rod or wire; wherein frequency operation of the high frequency magnetostrictive transducer is controlled by a length of the small coil segments and a material type of the magnetostrictive rod or wire. This design of the high frequency magnetostrictive transducer retains the beneficial aspects of the magnetostrictive design, while reducing its primary drawback, lower frequency operation.

Supercell Depletion Studies for Prismatic High Temperature Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

J. Ortensi

2012-10-01

The traditional two-step method of analysis is not accurate enough to represent the neutronic effects present in the prismatic high temperature reactor concept. The long range coupling of the various regions in high temperature reactors poses a set of challenges that are not seen in either LWRs or fast reactors. Unlike LWRs, which exhibit large, localized effects, the dominant effects in PMRs are, for the most part, distributed over larger regions, but with lower magnitude. The 1-D in-line treatment currently used in pebble bed reactor analysis is not sufficient because of the 2-D nature of the prismatic blocks. Considerable challengesmore » exist in the modeling of blocks in the vicinity of reflectors, which, for current small modular reactor designs with thin annular cores, include the majority of the blocks. Additional challenges involve the treatment of burnable poisons, operational and shutdown control rods. The use of a large domain for cross section preparation provides a better representation of the neutron spectrum, enables the proper modeling of BPs and CRs, allows the calculation of generalized equivalence theory parameters, and generates a relative power distribution that can be used in compact power reconstruction. The purpose of this paper is to quantify the effects of the reflector, burnable poison, and operational control rods on an LEU design and to delineate an analysis approach for the Idaho National Laboratory. This work concludes that the use of supercells should capture these long-range effects in the preparation of cross sections and along with a set of triangular meshes to treat BPs, and CRs a high fidelity neutronics computation is attainable.« less

Surfactant mediated hydrothermal synthesis, characterization and luminescent properties of GdPO{sub 4}: Ce{sup 3+}/Tb{sup 3+} @ GdPO{sub 4} core shell nanorods

DOE Office of Scientific and Technical Information (OSTI.GOV)

Khajuria, Heena; Ladol, Jigmet; Khajuria, Sonika

Highlights: • Core shell nanorods were synthesised by surfactant assisted hydrothermal method. • Morphology of core shell nanorods resembles those of core nanorods indicating coating of shell on cores. • More uniform and non-aggregated core shell nanorods were prepared in presence of surfactants. • Surfactant assisted prepared core shell nanorods show intense emission as compared to uncoated core nanorods. - Abstract: Core shell GdPO{sub 4}: Ce{sup 3+}/Tb{sup 3+} @ GdPO{sub 4} nanorods were synthesized via hydrothermal route in the presence of different surfactants [cetyltrimethyl ammonium bromide (CTAB) and Sodium dodecyl sulphate (SDS)]. The nanorods were characterized by powder X-ray diffractionmore » (PXRD), fourier transform infrared spectroscopy (FTIR), field emission scanning electron microscopy (FE-SEM), transmission electron microscopy (TEM), energy dispersive spectroscopy (EDS) and photoluminescence (PL) studies. The X-ray diffraction results indicate good crystallinity and effective doping in core and core shell nanorods. SEM and TEM micrographs show that all of the as prepared gadolinium phosphate products have rod like shape. The compositional analysis of GdPO{sub 4}: Ce{sup 3+}/Tb{sup 3+} core was done by EDS. The emission intensity of the GdPO{sub 4}: Ce{sup 3+}/Tb{sup 3+} @ GdPO{sub 4} core shell increased significantly with respect to those of GdPO{sub 4}: Ce{sup 3+}/Tb{sup 3+} core nanorods. The effect of surfactant on the uniformity, thickness and luminescence of the core shell nanorods was investigated.« less

Plasmon-Exciton Interactions Probed Using Spatial Coentrapment of Nanoparticles by Topological Singularities.

PubMed

Ackerman, Paul J; Mundoor, Haridas; Smalyukh, Ivan I; van de Lagemaat, Jao

2015-12-22

We study plasmon-exciton interaction by using topological singularities to spatially confine, selectively deliver, cotrap and optically probe colloidal semiconductor and plasmonic nanoparticles. The interaction is monitored in a single quantum system in the bulk of a liquid crystal medium where nanoparticles are manipulated and nanoconfined far from dielectric interfaces using laser tweezers and topological configurations containing singularities. When quantum dot-in-a-rod particles are spatially colocated with a plasmonic gold nanoburst particle in a topological singularity core, its fluorescence increases because blinking is significantly suppressed and the radiative decay rate increases by nearly an order of magnitude owing to the Purcell effect. We argue that the blinking suppression is the result of the radiative rate change that mitigates Auger recombination and quantum dot ionization, consequently reducing nonradiative recombination. Our work demonstrates that topological singularities are an effective platform for studying and controlling plasmon-exciton interactions.

Plasmon–Exciton Interactions Probed Using Spatial Coentrapment of Nanoparticles by Topological Singularities

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ackerman, Paul J.; Mundoor, Haridas; Smalyukh, Ivan I.

2015-12-22

We study plasmon-exciton interaction by using topological singularities to spatially confine, selectively deliver, cotrap and optically probe colloidal semiconductor and plasmonic nanoparticles. The interaction is monitored in a single quantum system in the bulk of a liquid crystal medium where nanoparticles are manipulated and nanoconfined far from dielectric interfaces using laser tweezers and topological configurations containing singularities. When quantum dot-in-a-rod particles are spatially colocated with a plasmonic gold nanoburst particle in a topological singularity core, its fluorescence increases because blinking is significantly suppressed and the radiative decay rate increases by nearly an order of magnitude owing to the Purcell effect.more » We argue that the blinking suppression is the result of the radiative rate change that mitigates Auger recombination and quantum dot ionization, consequently reducing nonradiative recombination. Our work demonstrates that topological singularities are an effective platform for studying and controlling plasmon-exciton interactions.« less

The controlled growth of GaN microrods on Si(111) substrates by MOCVD

NASA Astrophysics Data System (ADS)

Foltynski, Bartosz; Garro, Nuria; Vallo, Martin; Finken, Matthias; Giesen, Christoph; Kalisch, Holger; Vescan, Andrei; Cantarero, Andrés; Heuken, Michael

2015-03-01

In this paper, a selective area growth (SAG) approach for growing GaN microrods on patterned SiNx/Si(111) substrates by metal-organic chemical vapor deposition (MOCVD) is studied. The surface morphology, optical and structural properties of vertical GaN microrods terminated by pyramidal shaped facets (six { 10 1 bar 1} planes) were characterized using scanning electron microscopy (SEM), room temperature photoluminescence (PL) and Raman spectroscopy, respectively. Measurements revealed high-quality GaN microcolumns grown with silane support. Characterized structures were grown nearly strain-free (central frequency of Raman peak of 567±1 cm-1) with crystal quality comparable to bulk crystals (FWHM=4.2±1 cm-1). Such GaN microrods might be used as a next-generation device concept for solid-state lighting (SSL) applications by realizing core-shell InGaN/GaN multi-quantum wells (MQWs) on the n-GaN rod base.

ON CRITICAL MASS ANALYSIS OF JRR-2

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

1961-01-01

The critica mass of the JRR-2 was found to be 15 fuel elements, instead of 8 as expected, when the reactor reached criticaity. The critica mass was analyzed by AMF and JAERI a few years ago, but afterwards some modifications have been made of the stucture for the reinforcement, for example, during the construction. The critical mass is recalculated perfectly and the difference bctween 15 and S fuel elements is discussed. The deviation of the critical mass is mainly caused by the effects of control rods, fuel elcments, grid-plate, etc., in the reflector; only heavy water or light water wasmore » conaidered as the reflector in the previous calculation. A simple method is used to calculate the critical mass. The effective multiplication factor for the core with 15 fuel elements is obtained about 2% higher than the experimental value. This difference is also discussed in detail. (auth)« less

Collisionless relaxation in spiral galaxy models

NASA Technical Reports Server (NTRS)

Hohl, F.

1974-01-01

The increase in random kinetic energy of stars by rapidly fluctuating gravitational fields (collisionless or violent relaxation) in disk galaxy models is investigated for three interaction potentials of the stars corresponding to (1) point stars, (2) rod stars of length 2 kpc, and (3) uniform density spherical stars of radius 2 kpc. To stabilize the galaxy against the large scale bar forming instability, a fixed field corresponding to a central core or halo component of stars was added with the stars containing at most 20 percent of the total mass of the galaxy. Considerable heating occurred for both the point stars and the rod stars, whereas the use of spherical stars resulted in a very low heating rate. The use of spherical stars with the resulting low heating rate will be desirable for the study of large scale galactic stability or density wave propagation, since collective heating effects will no longer mask the phenomena under study.

Nuclear radiation actuated valve

DOEpatents

Christiansen, David W.; Schively, Dixon P.

1985-01-01

A nuclear radiation actuated valve for a nuclear reactor. The valve has a valve first part (such as a valve rod with piston) and a valve second part (such as a valve tube surrounding the valve rod, with the valve tube having side slots surrounding the piston). Both valve parts have known nuclear radiation swelling characteristics. The valve's first part is positioned to receive nuclear radiation from the nuclear reactor's fuel region. The valve's second part is positioned so that its nuclear radiation induced swelling is different from that of the valve's first part. The valve's second part also is positioned so that the valve's first and second parts create a valve orifice which changes in size due to the different nuclear radiation caused swelling of the valve's first part compared to the valve's second part. The valve may be used in a nuclear reactor's core coolant system.

Thermal Oxidation of a Carbon Condensate Formed in High-Frequency Carbon and Carbon-Nickel Plasma Flow

NASA Astrophysics Data System (ADS)

Churilov, G. N.; Nikolaev, N. S.; Cherepakhin, A. V.; Dudnik, A. I.; Tomashevich, E. V.; Trenikhin, M. V.; Bulina, N. G.

2018-02-01

We have reported on the comparative characteristics of thermal oxidation of a carbon condensate prepared by high-frequency arc evaporation of graphite rods and a rod with a hollow center filled with nickel powder. In the latter case, along with different forms of nanodisperse carbon, nickel particles with nickel core-carbon shell structures are formed. It has been found that the processes of the thermal oxidation of carbon condensates with and without nickel differ significantly. Nickel particles with the carbon shell exhibit catalytic properties with respect to the oxidation of nanosized carbon structures. A noticeable difference between the temperatures of the end of the oxidation process for various carbon nanoparticles and nickel particles with the carbon shell has been established. The study is aimed at investigations of the effect of nickel nanoparticles on the dynamics of carbon condensate oxidation upon heating in the argon-oxygen flow.

Technical Achievements in Communist China’s Electrical Equipment Industry

DTIC Science & Technology

1960-09-15

products has also been , developed, including long rod type insulating porcelains and a new series of line porcelains . In 1958, oil sockets for 330...is now aimed at the creation of high intensity, high insulating , and small-size high-tension porcelain products. During the past 10 years, our...of lead-covered oil-immersed paper- insulated cables of 55 kilovolts and less, rubber-sheathed cables of 6,000 volts and less, and aluminum core

CNGA3 Deficiency Affects Cone Synaptic Terminal Structure and Function and Leads to Secondary Rod Dysfunction and Degeneration

PubMed Central

Xu, Jianhua; Morris, Lynsie M.; Michalakis, Stylianos; Biel, Martin; Fliesler, Steven J.; Sherry, David M.

2012-01-01

Purpose. To investigate rod function and survival after cone dysfunction and degeneration in a mouse model of cone cyclic nucleotide-gated (CNG) channel deficiency. Methods. Rod function and survival in mice with cone CNG channel subunit CNGA3 deficiency (CNGA3−/− mice) were evaluated by electroretinographic (ERG), morphometric, and Western blot analyses. The arrangement, integrity, and ultrastructure of photoreceptor terminals were investigated by immunohistochemistry and electron microscopy. Results. The authors found loss of cone function and cone death accompanied by impairment of rods and rod-driven signaling in CNGA3−/− mice. Scotopic ERG b-wave amplitudes were reduced by 15% at 1 month, 30% at 6 months, and 40% at 9 months and older, while scotopic a-wave amplitudes were decreased by 20% at 9 months, compared with ERGs of age-matched wild-type mice. Outer nuclear layer thickness in CNGA3−/− retina was reduced by 15% at 12 months compared with age-matched wild-type controls. This was accompanied by a 30%–40% reduction in expression of rod-specific proteins, including rhodopsin, rod transducin α-subunit, and glutamic acid-rich protein (GARP). Cone terminals in the CNGA3−/− retina showed a progressive loss of neurochemical and ultrastructural integrity. Abnormalities were observed as early as 1 month. Disorganized rod terminal ultrastructure was noted by 12 months. Conclusions. These findings demonstrate secondary rod impairment and degeneration after cone degeneration in mice with cone CNG channel deficiency. Loss of cone phototransduction accompanies the compromised integrity of cone terminals. With time, rod synaptic structure, function, and viability also become compromised. PMID:22247469

Variable gas leak rate valve

DOEpatents

Eernisse, Errol P.; Peterson, Gary D.

1976-01-01

A variable gas leak rate valve which utilizes a poled piezoelectric element to control opening and closing of the valve. The gas flow may be around a cylindrical rod with a tubular piezoelectric member encircling the rod for seating thereagainst to block passage of gas and for reopening thereof upon application of suitable electrical fields.

Inhibitory masking controls the threshold sensitivity of retinal ganglion cells

PubMed Central

Pan, Feng; Toychiev, Abduqodir; Zhang, Yi; Atlasz, Tamas; Ramakrishnan, Hariharasubramanian; Roy, Kaushambi; Völgyi, Béla; Akopian, Abram

2016-01-01

Key points Retinal ganglion cells (RGCs) in dark‐adapted retinas show a range of threshold sensitivities spanning ∼3 log units of illuminance.Here, we show that the different threshold sensitivities of RGCs reflect an inhibitory mechanism that masks inputs from certain rod pathways.The masking inhibition is subserved by GABAC receptors, probably on bipolar cell axon terminals.The GABAergic masking inhibition appears independent of dopaminergic circuitry that has been shown also to affect RGC sensitivity.The results indicate a novel mechanism whereby inhibition controls the sensitivity of different cohorts of RGCs. This can limit and thereby ensure that appropriate signals are carried centrally in scotopic conditions when sensitivity rather than acuity is crucial. Abstract The responses of rod photoreceptors, which subserve dim light vision, are carried through the retina by three independent pathways. These pathways carry signals with largely different sensitivities. Retinal ganglion cells (RGCs), the output neurons of the retina, show a wide range of sensitivities in the same dark‐adapted conditions, suggesting a divergence of the rod pathways. However, this organization is not supported by the known synaptic morphology of the retina. Here, we tested an alternative idea that the rod pathways converge onto single RGCs, but inhibitory circuits selectively mask signals so that one pathway predominates. Indeed, we found that application of GABA receptor blockers increased the sensitivity of most RGCs by unmasking rod signals, which were suppressed. Our results indicate that inhibition controls the threshold responses of RGCs under dim ambient light. This mechanism can ensure that appropriate signals cross the bottleneck of the optic nerve in changing stimulus conditions. PMID:27350405

In vitro degradation, flexural, compressive and shear properties of fully bioresorbable composite rods.

PubMed

Felfel, R M; Ahmed, I; Parsons, A J; Walker, G S; Rudd, C D

2011-10-01

Several studies have investigated self-reinforced polylactic acid (SR-PLA) and polyglycolic acid (SR-PGA) rods which could be used as intramedullary (IM) fixation devices to align and stabilise bone fractures. This study investigated totally bioresorbable composite rods manufactured via compression moulding at ~100 °C using phosphate glass fibres (of composition 50P(2)O(5)-40CaO-5Na(2)O-5Fe(2)O(3) in mol%) to reinforce PLA with an approximate fibre volume fraction (v(f)) of 30%. Different fibre architectures (random and unidirectional) were investigated and pure PLA rods were used as control samples. The degradation profiles and retention of mechanical properties were investigated and PBS was selected as the degradation medium. Unidirectional (P50 UD) composite rods had 50% higher initial flexural strength as compared to PLA and 60% higher in comparison to the random mat (P50 RM) composite rods. Similar initial profiles for flexural modulus were also seen comparing the P50 UD and P50 RM rods. Higher shear strength properties were seen for P50 UD in comparison to P50 RM and PLA rods. However, shear stiffness values decreased rapidly (after a week) whereas the PLA remained approximately constant. For the compressive strength studies, P50 RM and PLA rods remained approximately constant, whilst for the P50 UD rods a significantly higher initial value was obtained, which decreased rapidly after 3 days immersion in PBS. However, the mechanical properties decreased after immersion in PBS as a result of the plasticisation effect of water within the composite and degradation of the fibres. The fibres within the random and unidirectional composite rods (P50 RM and P50 UD) degraded leaving behind microtubes as seen from the SEM micrographs (after 28 days degradation) which in turn created a porous structure within the rods. This was the main reason attributed for the increase seen in mass loss and water uptake for the composite rods (~17% and ~16%, respectively). Copyright © 2011 Elsevier Ltd. All rights reserved.

Measurement station for interim inspections of Lightbridge metallic fuel rods at the Halden Boiling Water Reactor

NASA Astrophysics Data System (ADS)

Hartmann, C.; Totemeier, A.; Holcombe, S.; Liverud, J.; Limi, M.; Hansen, J. E.; Navestad, E. AB(; )

2018-01-01

Lightbridge Corporation has developed a new Uranium-Zirconium based metallic fuel. The fuel rods aremanufactured via a co-extrusion process, and are characterized by their multi-lobed (cruciform-shaped) cross section. The fuel rods are also helically-twisted in the axial direction. Two experimental fuel assemblies, each containing four Lightbridge fuel rods, are scheduled to be irradiated in the Halden Boiling Water Reactor (HBWR) starting in 2018. In addition to on-line monitoring of fuel rod elongation and critical assembly conditions (e.g. power, flow rates, coolant temperatures, etc.) during the irradiation, several key parameters of the fuel will be measured out-of-core during interim inspections. An inspection measurement station for use in the irradiated fuel handling compartment at the HBWR has therefore been developed for this purpose. The multi-lobed cladding cross section combined with the spiral shape of the Lightbridge metallic fuel rods requires a high-precision guiding system to ensure good position repeatability combined with low-friction guiding. The measurement station is equipped with a combination of instruments and equipment supplied from third-party vendors and instruments and equipment developed at Institute for Energy Technology (IFE). Two sets of floating linear voltage differential transformer (LVDT) pairs are used to measure swelling and diameter changes between the lobes and the valleys over the length of the fuel rods. Eddy current probes are used to measure the thickness of oxide layers in the valleys and on the lobe tips and also to detect possible surface cracks/pores. The measurement station also accommodates gamma scans. Additionally, an eddy-current probe has been developed at IFE specifically to detect potential gaps or discontinuities in the bonding layer between the metallic fuel and the Zirconium alloy cladding. Potential gaps in the bonding layer will be hidden behind a 0.5-1.0 mm thick cladding wall. It has therefore been necessary to perform a careful design study of the probe geometry. For this, finite element analysis (FEA) has been performed in combination with practical validation tests on representative fuel dummies with machined flaws to find the probe geometry that best detects a hidden flaw. Tests performed thus far show that gaps down to 25 μm thickness can be detected with good repeatability and good discrimination from lift-off signals.

Immunomodulation-accelerated neuronal regeneration following selective rod photoreceptor cell ablation in the zebrafish retina.

PubMed

White, David T; Sengupta, Sumitra; Saxena, Meera T; Xu, Qingguo; Hanes, Justin; Ding, Ding; Ji, Hongkai; Mumm, Jeff S

2017-05-02

Müller glia (MG) function as inducible retinal stem cells in zebrafish, completely repairing the eye after damage. The innate immune system has recently been shown to promote tissue regeneration in which classic wound-healing responses predominate. However, regulatory roles for leukocytes during cellular regeneration-i.e., selective cell-loss paradigms akin to degenerative disease-are less well defined. To investigate possible roles innate immune cells play during retinal cell regeneration, we used intravital microscopy to visualize neutrophil, macrophage, and retinal microglia responses to induced rod photoreceptor apoptosis. Neutrophils displayed no reactivity to rod cell loss. Peripheral macrophage cells responded to rod cell loss, as evidenced by morphological transitions and increased migration, but did not enter the retina. Retinal microglia displayed multiple hallmarks of immune cell activation: increased migration, translocation to the photoreceptor cell layer, proliferation, and phagocytosis of dying cells. To test function during rod cell regeneration, we coablated microglia and rod cells or applied immune suppression and quantified the kinetics of ( i ) rod cell clearance, ( ii ) MG/progenitor cell proliferation, and ( iii ) rod cell replacement. Coablation and immune suppressants applied before cell loss caused delays in MG/progenitor proliferation rates and slowed the rate of rod cell replacement. Conversely, immune suppressants applied after cell loss had been initiated led to accelerated photoreceptor regeneration kinetics, possibly by promoting rapid resolution of an acute immune response. Our findings suggest that microglia control MG responsiveness to photoreceptor loss and support the development of immune-targeted therapeutic strategies for reversing cell loss associated with degenerative retinal conditions.

Immunomodulation-accelerated neuronal regeneration following selective rod photoreceptor cell ablation in the zebrafish retina

PubMed Central

White, David T.; Sengupta, Sumitra; Saxena, Meera T.; Xu, Qingguo; Hanes, Justin; Ding, Ding; Ji, Hongkai

2017-01-01

Müller glia (MG) function as inducible retinal stem cells in zebrafish, completely repairing the eye after damage. The innate immune system has recently been shown to promote tissue regeneration in which classic wound-healing responses predominate. However, regulatory roles for leukocytes during cellular regeneration—i.e., selective cell-loss paradigms akin to degenerative disease—are less well defined. To investigate possible roles innate immune cells play during retinal cell regeneration, we used intravital microscopy to visualize neutrophil, macrophage, and retinal microglia responses to induced rod photoreceptor apoptosis. Neutrophils displayed no reactivity to rod cell loss. Peripheral macrophage cells responded to rod cell loss, as evidenced by morphological transitions and increased migration, but did not enter the retina. Retinal microglia displayed multiple hallmarks of immune cell activation: increased migration, translocation to the photoreceptor cell layer, proliferation, and phagocytosis of dying cells. To test function during rod cell regeneration, we coablated microglia and rod cells or applied immune suppression and quantified the kinetics of (i) rod cell clearance, (ii) MG/progenitor cell proliferation, and (iii) rod cell replacement. Coablation and immune suppressants applied before cell loss caused delays in MG/progenitor proliferation rates and slowed the rate of rod cell replacement. Conversely, immune suppressants applied after cell loss had been initiated led to accelerated photoreceptor regeneration kinetics, possibly by promoting rapid resolution of an acute immune response. Our findings suggest that microglia control MG responsiveness to photoreceptor loss and support the development of immune-targeted therapeutic strategies for reversing cell loss associated with degenerative retinal conditions. PMID:28416692

The MAGEC system for spinal lengthening in children with scoliosis: A NICE Medical Technology Guidance.

PubMed

Jenks, Michelle; Craig, Joyce; Higgins, Joanne; Willits, Iain; Barata, Teresa; Wood, Hannah; Kimpton, Christine; Sims, Andrew

2014-12-01

Scoliosis-structural lateral curvature of the spine-affects around four children per 1,000. The MAGEC system comprises a magnetically distractible spinal rod implant and an external remote controller, which lengthens the rod; this system avoids repeated surgical lengthening. Rod implants brace the spine internally and are lengthened as the child grows, preventing worsening of scoliosis and delaying the need for spinal fusion. The Medical Technologies Advisory Committee at the National Institute for Health and Care Excellence (NICE) selected the MAGEC system for evaluation in a NICE medical technologies guidance. Six studies were identified by the sponsor (Ellipse Technologies Inc.) as being relevant to the decision problem. Meta-analysis was used to compare the clinical evidence results with those of one conventional growth rod study, and equal efficacy of the two devices was concluded. The key weakness was selection of a single comparator study. The External Assessment Centre (EAC) identified 16 conventional growth rod studies and undertook meta-analyses of relevant outcomes. Its critique highlighted limitations around study heterogeneity and variations in baseline characteristics and follow-up duration, precluding the ability to draw firm conclusions. The sponsor constructed a de novo costing model showing that MAGEC rods generated cost savings of £9,946 per patient after 6 years, compared with conventional rods. The EAC critiqued and updated the model structure and inputs, calculating robust cost savings of £12,077 per patient with MAGEC rods compared with conventional rods over 6 years. The year of valuation was 2012. NICE issued a positive recommendation as supported by the evidence (Medical Technologies Guidance 18).

Clip gage attachment for frictionless measurement of displacement during high-temperature mechanical testing

DOEpatents

Alexander, D.J.

1994-01-04

An attachment for placement between a test specimen and a remote clip gage extensometer providing improved fracture toughness tests of materials at elevated temperature is described. Using a cylindrical tube and axial rod in new relationship, the device transfers the displacement signal of the fracture toughness test specimen directly to a clip gage extensometer located outside the high temperature furnace. Virtually frictionless operation is assured by having the test specimen center one end of the rod in one end of the tube, while the clip gage extensometer arms center the other end of the rod in the other end of the tube. By providing positive control over both ends of both rod and tube, the attachment may be operated in orientations other than vertical. 1 figure.

BISON Fuel Performance Analysis of FeCrAl cladding with updated properties

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sweet, Ryan; George, Nathan M.; Terrani, Kurt A.

2016-08-30

In order to improve the accident tolerance of light water reactor (LWR) fuel, alternative cladding materials have been proposed to replace zirconium (Zr)-based alloys. Of these materials, there is a particular focus on iron-chromium-aluminum (FeCrAl) alloys due to much slower oxidation kinetics in high-temperature steam than Zr-alloys. This should decrease the energy release due to oxidation and allow the cladding to remain integral longer in the presence of high temperature steam, making accident mitigation more likely. As a continuation of the development for these alloys, suitability for normal operation must also be demonstrated. This research is focused on modeling themore » integral thermo-mechanical performance of FeCrAl-cladded fuel during normal reactor operation. Preliminary analysis has been performed to assess FeCrAl alloys (namely Alkrothal 720 and APMT) as a suitable fuel cladding replacement for Zr-alloys, using the MOOSE-based, finite-element fuel performance code BISON and the best available thermal-mechanical and irradiation-induced constitutive properties. These simulations identify the effects of the mechanical-stress and irradiation response of FeCrAl, and provide a comparison with Zr-alloys. In comparing these clad materials, fuel rods have been simulated for normal reactor operation and simple steady-state operation. Normal reactor operating conditions target the cladding performance over the rod lifetime (~4 cycles) for the highest-power rod in the highest-power fuel assembly under reactor power maneuvering. The power histories and axial temperature profiles input into BISON were generated from a neutronics study on full-core reactivity equivalence for FeCrAl using the 3D full core simulator NESTLE. Evolution of the FeCrAl cladding behavior over time is evaluated by using steady-state operating conditions such as a simple axial power profile, a constant cladding surface temperature, and a constant fuel power history. The fuel rod designs and operating conditions used are based off the Peach Bottom BWR and design consideration was given to minimize the neutronic penalty of the FeCrAl cladding by changing fuel enrichment and cladding thickness. As this study progressed, systematic parametric analysis of the fuel and cladding creep responses were also performed.« less

Learned arbitrary responses to light in mice without rods or cones

NASA Astrophysics Data System (ADS)

Mrosovsky, N.; Salmon, Peggy

2002-10-01

The aim of this investigation was to discover whether mice lacking classical photoreceptors (rods and cones) can nevertheless be trained to respond to light. Mice with the coneless (cl) transgene have an attenuated diphtheria toxin fused to a cone opsin promotor. Mutant mice homozygous for the retinal degeneration (rd) gene undergo loss of their rods. By mating these two strains, mice lacking both cones and rods can be generated (Lucas et al. 1999). Such coneless-rodless mice were able to use light as a signal to make a behavioural response to avoid impending shock. Nevertheless, especially initially, they used the light as a cue less often than wildtype controls, indicating that normally the rods and cones are used for such responses. However, other photoreceptors are able to take over this role to some extent. When the lights were covered with opaque material, the performance of rodless-coneless mice dropped to chance level, indicating that they had been using the light as a cue for avoidance.