Nuclear reactor internals and control rod handling device

International Nuclear Information System (INIS)

Betancourt, G.N.; Etzel, W.W.

1981-01-01

A method and apparatus for removing, in an essentially continuous operation, the control rods and the upper guide structure from a nuclear reactor vessel during refueling. The apparatus includes a rigid frame which is secured to the upper guide structure after the vessel head is removed. A platform is vertically reciprocable within the frame and is adapted to engage and lift simultaneously all control rod drive shafts to a maximum elevation within the frame. A mechanical interface between the platform and the frame is provided so that continuation of the lifting force on the platform transfers the lift force to the frame whereby the upper guide structure is lifted out of the vessel. Automatically operated stop means are provided to lock the platform and rods in the maximum elevation within the frame in order to prevent accidental dropping of the rods during transfer of the upper guide structure and control rods to a temporary storage area

Reactor core with rod-shaped fuel cells

International Nuclear Information System (INIS)

Dworak, A.

1975-01-01

Power distribution in a high-temperature gas-cooled reactor is optimized. Especially the axial as well as the radial power distribution is kept constant, the core consisting of several consecutive rod-shaped fuel cells. To this end, the dwell times of the fuel cells are fitted to the given power distribution. Fuel cells with equal dwell times, seen in flow direction, are arranged side by side, and those with the shortest dwell times are placed in areas with the greatest power release. These areas ly on the coolant inlet side. To keep the power distribution constant, fuel cells with neutron poison or absorber rods with absorbing rates decreasing in flow direction can also be inserted. (RW/PB) [de

Gray rod for a nuclear reactor

International Nuclear Information System (INIS)

Francis, T.A.; Cerni, Samuel.

1986-01-01

The invention relates to an improved gray rod for insertion in a nuclear fuel assembly having an array of fuel rods. The gray rod includes a thin-walled cladding tube a first longitudinal section of which is positioned within, and a second longitudinal section of which is positioned essentially without, the array of fuel rods when the gray rod is inserted in the fuel assembly. The first longitudinal section defines a pellet-receiving space having detained therein a stack of annular pellets with an outer diameter sufficient to lend radial support to the wall of the first longitudinal tube section. The second longitudinal section defines a hollow space devoid of pellets and having means to resist radial collapse under external pressure. This means may be a partially compressed spiral spring which serves the dual purpose of retaining the stack of pellets in the pellet-receiving space and of lending radial support to the wall of the second longitudinal tube section or it may be holes through the wall to allow pressure equalisation. The cladding tube is composed of stainless-steel material having a low neutron-capture cross-section, and the annular pellets preferably being composed of Zircaloy or Zirconia material. (author)

Control rod

International Nuclear Information System (INIS)

Kawakami, Kazuo; Shimoshige, Takanori; Nishimura, Akira

1979-01-01

Purpose: A control rod has been developed, which provided a plurality of through-holes in the vicinity of the sheath fitting position, in order to flatten burn-up, of fuel rods in positions confronting a control rod. Thereby to facilitate the manufacture of the control rods and prevent fuel rod failures. Constitution: A plurality of through-holes are formed in the vicinity of the sheath fitting position of a central support rod to which a sheath for the control rod is fitted. These through-holes are arranged in the axial direction of the central support rod. Accordingly, burn-up of fuel rods confronting the control rods can be reduced by through-holes and fuel rod failures can be prevented. (Yoshino, Y.)

Firmware development and testing of the ATLAS Pixel Detector / IBL ROD card

International Nuclear Information System (INIS)

Gabrielli, A.; Balbi, G.; Falchieri, D.; Lama, L.; Travaglini, R.; Backhaus, M.; Bindi, M.; Chen, S.P.; Hauck, S.; Hsu, S.C.; Flick, T.; Wensing, M.; Kretz, M.; Kugel, A.

2015-01-01

The ATLAS Experiment is reworking and upgrading systems during the current LHC shut down. In particular, the Pixel detector has inserted an additional inner layer called the Insertable B-Layer (IBL). The Readout-Driver card (ROD), the Back-of-Crate card (BOC), and the S-Link together form the essential frontend data path of the IBL's off-detector DAQ system. The strategy for IBL ROD firmware development was three-fold: keeping as much of the Pixel ROD datapath firmware logic as possible, employing a complete new scheme of steering and calibration firmware, and designing the overall system to prepare for a future unified code version integrating IBL and Pixel layers. Essential features such as data formatting, frontend-specific error handling, and calibration are added to the ROD data path. An IBL DAQ test bench using a realistic front-end chip model was created to serve as an initial framework for full offline electronic system simulation. In this document, major firmware achievements concerning the IBL ROD data path implementation, test on the test bench and ROD prototypes, will be reported. Recent Pixel collaboration efforts focus on finalizing hardware and firmware tests for the IBL. The plan is to approach a complete IBL DAQ hardware-software installation by the end of 2014

Fuel rod design by statistical methods for MOX fuel

International Nuclear Information System (INIS)

Heins, L.; Landskron, H.

2000-01-01

Statistical methods in fuel rod design have received more and more attention during the last years. One of different possible ways to use statistical methods in fuel rod design can be described as follows: Monte Carlo calculations are performed using the fuel rod code CARO. For each run with CARO, the set of input data is modified: parameters describing the design of the fuel rod (geometrical data, density etc.) and modeling parameters are randomly selected according to their individual distributions. Power histories are varied systematically in a way that each power history of the relevant core management calculation is represented in the Monte Carlo calculations with equal frequency. The frequency distributions of the results as rod internal pressure and cladding strain which are generated by the Monte Carlo calculation are evaluated and compared with the design criteria. Up to now, this methodology has been applied to licensing calculations for PWRs and BWRs, UO 2 and MOX fuel, in 3 countries. Especially for the insertion of MOX fuel resulting in power histories with relatively high linear heat generation rates at higher burnup, the statistical methodology is an appropriate approach to demonstrate the compliance of licensing requirements. (author)

Aerobic Glycolysis Is Essential for Normal Rod Function and Controls Secondary Cone Death in Retinitis Pigmentosa.

Science.gov (United States)

Petit, Lolita; Ma, Shan; Cipi, Joris; Cheng, Shun-Yun; Zieger, Marina; Hay, Nissim; Punzo, Claudio

2018-05-29

Aerobic glycolysis accounts for ∼80%-90% of glucose used by adult photoreceptors (PRs); yet, the importance of aerobic glycolysis for PR function or survival remains unclear. Here, we further established the role of aerobic glycolysis in murine rod and cone PRs. We show that loss of hexokinase-2 (HK2), a key aerobic glycolysis enzyme, does not affect PR survival or structure but is required for normal rod function. Rods with HK2 loss increase their mitochondrial number, suggesting an adaptation to the inhibition of aerobic glycolysis. In contrast, cones adapt without increased mitochondrial number but require HK2 to adapt to metabolic stress conditions such as those encountered in retinitis pigmentosa, where the loss of rods causes a nutrient shortage in cones. The data support a model where aerobic glycolysis in PRs is not a necessity but rather a metabolic choice that maximizes PR function and adaptability to nutrient stress conditions. Copyright © 2018 The Author(s). Published by Elsevier Inc. All rights reserved.

Repulsion between oppositely charged rod-shaped macromolecules: Role of overcharging and ionic confinement

Science.gov (United States)

Antila, Hanne S.; Van Tassel, Paul R.; Sammalkorpi, Maria

2017-09-01

The interaction between two oppositely charged rod-shaped macro-ions in a micro-ion solution is investigated via Monte Carlo simulations of the primitive model. The focus is on the asymmetry in rod and/or ion charge, i.e., conditions where oppositely charged objects can repel one another. For equally and oppositely charged rods with asymmetric z:1 micro-ions, repulsion may be induced by overcharging one of the rods with the z valent ions. For asymmetrically charged rods in a symmetric z:z micro-ion solution, a repulsive interaction—at separation of the order of one ion diameter—can arise via an unbalanced osmotic pressure contribution from the ionic atmosphere in the inter-rod space, and an attractive interaction—at a smaller separation—may occur due to a "squeezing out" of the micro-ions from the space between the rods (with a consequent gain in entropy). The thermodynamics of each mechanism is investigated in terms of rod charge and size and micro-ion valence, size, and concentration. Our findings contribute to the understanding of the complex role of charge asymmetry on the interaction of, for example, oppositely charged polyelectrolytes, functionalized nanotubes, and rod-like biomolecules, e.g., viruses.

Control rod

International Nuclear Information System (INIS)

Igarashi, Takao; Sugawara, Satoshi; Yoshimoto, Yuichiro; Saito, Shozo; Fukumoto, Takashi.

1987-01-01

Purpose: To reduce the weight and thereby obtain satisfactory operationability of control rods by combining absorbing nuclear chain type neutron absorbers and conventional type neutron absorbers in the axial direction of blades. Constitution: Neutron absorber rods and long life type neutron absorber rods are disposed in a tie rod and a sheath. The neutron absorber rod comprises a poison tube made of stainless steels and packed with B 4 C powder. The long life type neutron absorber rod is prepared by packing B-10 enriched boron carbide powder into a hafnium metal rod, hafnium pipe, europium and stainless made poison tube. Since the long life type absorber rod uses HF as the absorbing nuclear chain type neutron absorber, it absorbs neutrons to form new neutron absorbers to increase the nuclear life. (Yoshino, Y.)

Control rods

International Nuclear Information System (INIS)

Maruyama, Hiromi.

1984-01-01

Purpose: To realize effective utilization, cost reduction and weight reduction in neutron absorbing materials. Constitution: Residual amount of neutron absorbing material is averaged between the top end region and other regions of a control rod upon reaching to the control rod working life, by using a single kind of neutron absorbing material and increasing the amount of the neutron absorber material at the top end region of the control rod as compared with that in the other regions. Further, in a case of a control rod having control rod blades such as in a cross-like control rod, the amount of the neutron absorbing material is decreased in the middle portion than in the both end portions of the control rod blade along the transversal direction of the rod, so that the residual amount of the neutron absorbing material is balanced between the central region and both end regions upon reaching the working life of the control rod. (Yoshihara, H.)

Control rod calibration including the rod coupling effect

International Nuclear Information System (INIS)

Szilard, R.; Nelson, G.W.

1984-01-01

In a reactor containing more than one control rod, which includes all reactors licensed in the United States, there will be a 'coupling' or 'shadowing' of control rod flux at the location of a control rod as a result of the flux depression caused by another control rod. It was decided to investigate this phenomenon further, and eventually to put calibration table data or formulae in a small computer in the control room, so once could insert the positions of the three control rods and receive the excess reactivity without referring to separate tables. For this to be accomplished, a 'three control- rod reactivity function' would be used which would include the flux coupling between the rods. The function is design and measured data was fitted into it to determine the calibration constants. The input data for fitting the trial functions consisted of 254 data points, each consisting of the position of the reg, shim, and transient rods, and the total excess reactivity. (About 200 of these points were 'critical balance points', that is the rod positions for which reactor was critical, and the remainder were determined by positive period measurements.) Although this may be unrealistic from a physical viewpoint, the function derived gave a very accurate recalculation of the input data, and thus would faithfully give the excess reactivity for any possible combination of the locations of the three control rods. The next step, incorporation of the three-rod function into the minicomputer, will be pursued in the summer and fall of 1984

A cw 4-rod RFQ linac

International Nuclear Information System (INIS)

Fujisawa, Hiroshi

1994-01-01

A cw 4-rod RFQ linac system has been designed, constructed, and tested as an accelerator section of a MeV-class ion implanter system. The tank diameter is only 60 cm for 34 MHz operating frequency. An equally spaced arrangement of the RFQ electrode supporting plates is proved to be suitable for a low resonant frequency 4-rod RFQ structure. The RFQ electrode cross section is not circular but rectangular to make the handling and maintenance of the electrodes easier. The machining of the electrode is done three dimensionally. Second order corrections in the analyzing magnet of the LEBT (Low Energy Beam Transport) section assure a better transmission through and the matching to the RFQ. A new approach is introduced to measure the rf characteristics of the 4-rod RFQ. This method requires only a few capacitors and a network analyzer. Both the rf and thermal stability of the 4-rod RFQ are tested up to cw 50 kW. Beam experiments with several ions confirm the acceleration of beams to the goal energy of 83 keV/u. The ion beam intensities obtained at the RFQ output for He + , N 2+ , and C + are 32, 13, and 220 pμA, respectively. The measured beam transmissions of >80% agree with the PARMTEQ calculations. The ion implantation method also gives definitive information on the energies of an RFQ output beam. ((orig.))

On-line fuel and control rod integrity management in BWRs

International Nuclear Information System (INIS)

Larsson, Irina; Sihver, Lembit

2011-01-01

Surveillance of fuel and control rod integrity in a BWR core is essential to maintain a safe and reliable operation of a nuclear power plant. An accurate and prompt way to monitor fuel integrity in a reactor core during reactor operation is by using on-line measurements of the gamma emitting noble gas activities in the off-gas system. The integrity of control rods can be efficiently followed by on-line measurements of the helium (He) concentration in the off-gases. This method also gives information about fuel rod failures since He is used as a fill gas in the fuel rods. To survey fuel and control rod integrity during reactor operation, a system consisting of combined gamma and He on-line measurements in the off-gases should be used. Such a system can detect and follow the behavior of fuel and control rod failures. In addition, it can separate fuel failures from control rod failures since fuel rods contain both He and gamma emitting noble gases, while control rods only contain He. Moreover, the system is able to distinguish primary fuel failures from degradation of already existing ones. In this paper we present a combined system for on-line measurements of He and gamma emitting noble gases in the reactor off-gas system and measuring experiences from different BWRs. (author)

Key developments of a rod control system - 15101

International Nuclear Information System (INIS)

Pouillot, M.; Jegou, H.; Duthou, A.

2015-01-01

The aim of the Rod Control System is to carry out the insertion and withdrawal of control rod clusters to provide the power required by the grid (G-mode control), to control the temperature of the reactor, or to provide negative reactivity margin when the reactor is shut down. The rod control system is not classified important for safety, but its correct operation is essential for the availability of the reactor, as the spurious drop of a single cluster usually results in a reactor trip. Rolls-Royce has been designing, manufacturing and providing rod control systems since 1977, in France, China, Belgium, Korea, and South Africa, as an original manufacturer and for modernization projects. All the corresponding nuclear units share the following features, key points for the system design: -) The power source is a three-phased 260 Vac with neutral, provided by zigzag-coupled alternators; -) The Control Rod Drive Mechanisms (CRDM) are 'three-coil type': Stationary Gripper (SG), Movable Gripper (MG) and Lift Coil (LC); -) Rod clusters are arranged in banks and sub-banks, the bank being composed of one or two sub-banks and a sub-bank is a set of 4 clusters moved simultaneously, the central cluster being an exception; and -) Most of those reactors are operated in G-mode (load following). (authors)

Apparatus for loading fuel rods into grids of nuclear fuel assemblies

International Nuclear Information System (INIS)

Shallenberger, J.M.; Ferlan, S.J.

1989-01-01

For use with a nuclear fuel assembly including support grids having cells for receiving fuel rods and with detents disposed within the respective cells for resiliently engaging and laterally supporting the fuel rods received therein, an apparatus is described for facilitating scratchless insertion of each fuel rod into cells of the support rids. The apparatus consists of: a thin-walled metallic tubular member which is long enough to extend through at least a majority of support grids, and is positionable so as to have its thin wall interposed, during insertion of each fuel rod, between the latter and the detents within the cells receiving it, the thin-walled tubular member having a substantially uniform wall thickness of not more than about 0.008 inch, an as-formed inner diameter substantially equal to the outer diameter of the fuel rod, and a longitudinal slit formed in the wall of the tubular member so as to render the wall resiliently deflectable in a diameter-reducing sense, the longitudinal slit having a width sufficient to preclude overlapping of the edges of the wall along the slit, and insufficient for any of the detents to enter the slit when the wall of the tubular member is in position between the detents and the fuel rod

Gelation And Mechanical Response of Patchy Rods

Science.gov (United States)

Kazem, Navid; Majidi, Carmel; Maloney, Craig

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

Evaluation of the Effect of Fixation Angle between Polyaxial Pedicle Screw Head and Rod on the Failure of Screw-Rod Connection

Directory of Open Access Journals (Sweden)

Engin Çetin

2015-01-01

Full Text Available Introduction. Polyaxial screws had been only tested according to the ASTM standards (when they were perpendicularly positioned to the rod. In this study, effects of the pedicle screws angled fixation to the rod on the mechanical properties of fixation were investigated. Materials and Method. 30 vertically fixed screws and 30 screws fixed with angle were used in the study. Screws were used in three different diameters which were 6.5 mm, 7.0 mm, and 7.5 mm, in equal numbers. Axial pull-out and flexion moment tests were performed. Test results compared with each other using appropriate statistical methods. Results. In pull-out test, vertically fixed screws, in 6.5 mm and 7.0 mm diameter, had significantly higher maximum load values than angled fixed screws with the same diameters (P<0.01. Additionally, vertically fixed screws, in all diameters, had significantly greater stiffness according to corresponding size fixed with angle (P<0.005. Conclusion. Fixing the pedicle screw to the rod with angle significantly decreased the pull-out stiffness in all diameters. Similarly, pedicle screw instrumentation fixed with angle decreased the minimum sagittal angle between the rod and the screw in all diameters for flexion moment test but the differences were not significant.

Control rod drives

International Nuclear Information System (INIS)

Nakamura, Akira.

1984-01-01

Purpose: To enable to monitor the coupling state between a control rod and a control rod drive. Constitution: After the completion of a control rod withdrawal, a coolant pressure is applied to a control rod drive being adjusted so as to raise only the control rod drive and, in a case where the coupling between the control rod drive and the control rod is detached, the former is elevated till it contacts the control rod and then stopped. The actual stopping position is detected by an actual position detection circuit and compared with a predetermined position stored in a predetermined position detection circuit. If both of the positions are not aligned with each other, it is judged by a judging circuit that the control rod and the control rod drives are not combined. (Sekiya, K.)

Loudspeaker Equalization with Post-Processing

Directory of Open Access Journals (Sweden)

Ser Wee

2002-01-01

Full Text Available Loudspeaker equalization is an essential technique in audio system design. A well-known equalization scheme is based on the deconvolution of the desired equalized response with the measured impulse response of the loudspeaker. In this paper, a post-processing scheme is combined with the deconvolution-based algorithm to provide a better equalization effect. Computer simulation results are given to demonstrate the significant improvement that can be achieved using this method.

Control rod assembly

International Nuclear Information System (INIS)

Takahashi, Akio.

1982-01-01

Purpose: To enable reliable insertion and drops of control rods, as well as insure a sufficient flow rate of coolants flowing through the control rods for attaining satisfactory cooling thereof to enable relexation of thermal stress resulted to rectifying mechanisms or the likes. Constitution: To the outer circumference of a control rod contained vertically movably within a control rod guide tube, resistive members are retractably provided in such a way as to project to close the gap between outer circumference of the control rod and the inner surface of the control rod guide tube upon engagement of a gripper of control rod drives, and retract upon release of the engagement of the gripper. Thus, since the resistive members project to provide a greater resistance to the coolants flowing between them and the control rod guide tube in the normal operation where the gripper is engaged to drive the control rod by the control rod drives, a major part of the coolant flowing into the control rod guide tube flows into the control rod. This enables to cool the control rod effectively and make the temperature distribution uniform for the coolant flowing from the upper end of the control rod guide tube to thereby attain the relaxation of the thermal stress resulted in the rectifying mechanisms or the likes. (Moriyama, K.)

Device for coupling a control rod and control rod drive

International Nuclear Information System (INIS)

Nishioka, Kazuya.

1975-01-01

Object: To obtain simple and reliable coupling between a control rod and control rod drive by equipping the lower end of the control rod with an extension provided with lateral protuberances and forming the upper end of an index tube with a recess provided with lateral holes. Structure: The tapering central extension of the control rod is inserted into the recess by lowering the control rod, and then it is further inserted by causing frictional movement of the inclined surfaces of lateral protuberances in frictional contact with guide surfaces. When the lateral protuberances are brought into contact with a stepped portion, the control rod is rotated to fit the lateral protuberances into the lateral holes. In this way, the control rod is coupled to the index tube of the control rod drive. (Yoshino, Y.)

Effect analysis of air introduced by pressurization on fuel rod performances

International Nuclear Information System (INIS)

Ren Qisen; Liu Tong; Sheng Guofu

2012-01-01

In the process of pressurization and seal welding, it is common practice to vacuumize before gas filling for the sake of preventing introducing air and other impurities, which would affect the gas composition inside of the fuel rod. However, vacuumization during pressurization is likely not being required sometimes in order to simplify the fabrication procedure. In the present work, based on the AFA3G fuel rod design with 2 MPa of filling gas, analyses on fuel rod performances were carried out under the condition of pressurization with and without vacuumization, respectively. Furthermore, the effect on hydrogen content in fuel rod was preliminarily discussed. Results indicate that the impacts of air composition introduced by pressurization on fuel rod thermal-mechanical performances, such as internal pressure and fuel center temperature, were extremely slight. The gap conductance varies to some extent as a result of the change of gas composition due to air introduced in fuel rod. The impact of humidity on water content in fuel rod is negligible at a low temperature of around 25℃. However, at higher temperature, it is essential to pay attention on the control of fabrication process, and prevent much moisture entering into the fuel rod and increasing the probability of hydriding failure. (authors)

Control rod displacement

International Nuclear Information System (INIS)

Nakazato, S.

1987-01-01

This patent describes a nuclear reactor including a core, cylindrical control rods, a single support means supporting the control rods from their upper ends in spaced apart positions and movable for displacing the control rods in their longitudinal direction between a first end position in which the control rods are fully inserted into the core and a second end position in which the control rods are retracted from the core, and guide means contacting discrete regions of the outer surface of each control rod at least when the control rods are in the vicinity of the second end position. The control rods are supported by the support means for longitudinal movement without rotation into and out of the core relative to the guide means to thereby cause the outer surface of the control rods to experience wear as a result of sliding contact with the guide means. The support means are so arranged with respect to the core and the guide means that it is incapable of rotation relative to the guide means. The improvement comprises displacement means being operatively coupled to a respective one of the control rods for periodically rotating the control rod in a single angular direction through an angle selected to change the locations on the outer surfaces of the control rods at which the control rods are contacted by the guide means during subsequent longitudinal movement of the control rods

Morphoelastic rods. Part I: A single growing elastic rod

KAUST Repository

Moulton, D.E.

2013-02-01

A theory for the dynamics and statics of growing elastic rods is presented. First, a single growing rod is considered and the formalism of three-dimensional multiplicative decomposition of morphoelasticity is used to describe the bulk growth of Kirchhoff elastic rods. Possible constitutive laws for growth are discussed and analysed. Second, a rod constrained or glued to a rigid substrate is considered, with the mismatch between the attachment site and the growing rod inducing stress. This stress can eventually lead to instability, bifurcation, and buckling. © 2012 Elsevier Ltd. All rights reserved.

Morphoelastic rods. Part I: A single growing elastic rod

KAUST Repository

Moulton, D.E.; Lessinnes, T.; Goriely, A.

2013-01-01

A theory for the dynamics and statics of growing elastic rods is presented. First, a single growing rod is considered and the formalism of three-dimensional multiplicative decomposition of morphoelasticity is used to describe the bulk growth of Kirchhoff elastic rods. Possible constitutive laws for growth are discussed and analysed. Second, a rod constrained or glued to a rigid substrate is considered, with the mismatch between the attachment site and the growing rod inducing stress. This stress can eventually lead to instability, bifurcation, and buckling. © 2012 Elsevier Ltd. All rights reserved.

Firmware development and testing of the ATLAS Pixel Detector / IBL ROD card

CERN Document Server

Balbi, G; The ATLAS collaboration; Gabrielli, A; Lama, L; Travaglini, R; Backhaus, M; Bindi, M; Chen, S-P; Flick, T; Kretz, M; Kugel, A; Wensing, M

2014-01-01

The ATLAS Experiment is reworking and upgrading systems during the current LHC shut down. In particular, the Pixel detector has inserted an additional inner layer called Insertable B-Layer (IBL). The Readout-Driver card (ROD), the Back-of-Crate card (BOC), and the S-Link together form the essential frontend data path of the IBL’s off-detector DAQ system. The strategy for IBLROD firmware development was three-fold: keeping as much of the PixelROD datapath firmware logic as possible, employing a complete new scheme of steering and calibration firmware and designing the overall system to prepare for a future unified code version integrating IBL and Pixel layers. Essential features such as data formatting, frontend-specific error handling, and calibration are added to the ROD data path. An IBLDAQ testbench using realistic frontend chip model was created to serve as an initial framework for full offline electronic system simulation. In this document, major firmware achievements concerning the IBLROD data path im...

Control rod drive

International Nuclear Information System (INIS)

Hawke, B.C.

1986-01-01

A reactor core, one or more control rods, and a control rod drive are described for selectively inserting and withdrawing the one or more control rods into and from the reactor core, which consists of: a support structure secured beneath the reactor core; control rod positioning means supported by the support structure for movably supporting the control rod for movement between a lower position wherein the control rod is located substantially beneath the reactor core and an upper position wherein at least an upper portion of the control rod extends into the reactor core; transmission means; primary drive means connected with the control rod positioning means by the transmission means for positioning the control rod under normal operating conditions; emergency drive means for moving the control rod from the lower position to the upper position under emergency conditions, the emergency drive means including a weight movable between an upper and a lower position, means for movably supporting the weight, and means for transmitting gravitational force exerted on the weight to the control rod positioning means to move the control rod upwardly when the weight is pulled downwardly by gravity; the transmission means connecting the control rod positioning means with the emergency drive means so that the primary drive means effects movement of the weight and the control rod in opposite directions under normal conditions, thus providing counterbalancing to reduce the force required for upward movement of the control rod under normal conditions; and restraint means for restraining the fall of the weight under normal operating conditions and disengaging the primary drive means to release the weight under emergency conditions

Data report of a tight-lattice rod bundle thermal-hydraulic tests (1). Base case test using 37-rod bundle simulated water-cooled breeder reactor (Contract research)

International Nuclear Information System (INIS)

Kureta, Masatoshi; Tamai, Hidesada; Liu, Wei; Akimoto, Hajime; Sato, Takashi; Watanabe, Hironori; Ohnuki, Akira

2006-03-01

Japan Atomic Energy Agency has been performing tight-lattice rod bundle thermal-hydraulic tests to realize essential technologies for the technological and engineering feasibility of super high burn-up water-cooled breeder reactor featured by a high breeding ratio and super high burn-up by reducing the core water volume in water-cooled reactor. The tests are performing to make clear the fundamental subjects related to the boiling transition (BT) (Subjects: BT criteria under a highly tight-lattice rod bundle, effects of gap-width between rods and of rod-bowing) using 37-rod bundles (Base case test section (1.3mm gap-width), Two parameter effect test sections (Gap-width effect one (1.0mm) and Rod-bowing one)). In the present report, we summarize the test results from the base case test section. The thermal-hydraulic characteristics using the large scale test section were obtained for the critical power, the pressure drop and the wall heat transfer under a wide range of pressure, flow rate, etc. including normal operational conditions of the designed reactor. Effects of local peaking factor on the critical power were also obtained. (author)

Replacement rod

International Nuclear Information System (INIS)

Hatfield, S.C.

1989-01-01

This patent describes in an elongated replacement rod for use with fuel assemblies of the type having two end fittings connected by guide tubes with a plurality of rod and guide tube cell defining spacer grids containing rod support features and mixing vanes. The grids secured to the guide tubes in register between the end fittings at spaced intervals. The fuel rod comprising: an asymmetrically beveled tip; a shank portion having a straight centerline; and a permanently diverging portion between the tip and the shank portion

Tokay gecko photoreceptors achieve rod-like physiology with cone-like proteins.

Science.gov (United States)

Zhang, Xue; Wensel, Theodore G; Yuan, Ching

2006-01-01

The retinal photoreceptors of the nocturnal Tokay gecko (Gekko gekko) consist exclusively of rods by the criteria of morphology and key features of their light responses. Unlike cones, they display robust photoresponses and have relatively slow recovery times. Nonetheless, the major and minor visual pigments identified in gecko rods are of the cone type by sequence and spectroscopic behavior. In the ongoing search for the molecular bases for the physiological differences between cones and rods, we have characterized the molecular biology and biochemistry of the gecko rod phototransduction cascade. We have cloned cDNAs encoding all or part of major protein components of the phototransduction cascade by RT-PCR with degenerate oligonucleotides designed to amplify cone- or rod-like sequences. For all proteins examined we obtained only cone-like and never rod-like sequences. The proteins identified include transducin alpha (Galphat), phosphodiesterase (PDE6) catalytic and inhibitory subunits, cyclic nucleotide-gated channel (CNGalpha) and arrestin. We also cloned cDNA encoding gecko RGS9-1 (Regulator of G Protein Signaling 9, splice variant 1), which is expressed in both rods and cones of all species studied but is typically found at 10-fold higher concentrations in cones, and found that gecko rods contain slightly lower RGS9-1 levels than mammalian rods. Furthermore, we found that the levels of GTPase accelerating protein (GAP) activity and cyclic GMP (cGMP) phosphodiesterase activity were similar in gecko and mammalian rods. These results place substantial constraints on the critical changes needed to convert a cone into a rod in the course of evolution: The many features of phototransduction molecules conserved between those expressed in gecko rods and those expressed in cones cannot explain the physiological differences, whereas the higher levels of RGS9-1 and GAP activity in cones are likely among the essential requirements for the rapid photoresponses of cones.

Water rod

International Nuclear Information System (INIS)

Kashiwai, Shin-ichi; Yokomizo, Osamu; Orii, Akihito.

1992-01-01

In a reactor core of a BWR type reactor, the area of a flow channel in a lower portion of a downcoming pipe for downwardly releasing steams present at the top portion in a water rod is increased. Further, a third coolant flow channel (an inner water rod) is disposed in an uprising having an exit opened near the inlet of the water rod and an inlet opened at the outside near the top portion of the water and having an increase flow channel area in the upper portion. The downcoming pipe in the water rod is filled with steams, and the void ratio is increased by so much as the flow channel area of the downcoming pipe is increased. Since the pressure difference between the inlet and the exit of the inner water rod is greater than the pressure difference between the inlet and the exit of the water rod, most of water flown into the inner water rod is discharged out of the exit in the form of water as it is. Since the area of the flow channel is increased in the portion of the inner water rod, void efficiency in the upper portion of the reactor core is decreased by so much. Since the void ratio is thus increased in the lower portion and the void efficiency is decreased in the upper portion of the reactor core, axial void distribution can be flattened. (N.H.)

Equality in Sport for Women.

Science.gov (United States)

Geadelmann, Patricia L.; And Others

The subject of equal rights and opportunities for women in the field of physical education is discussed in nine articles. The major emphasis is on the legal aspects of sex discrimination. Defining equality, knowing the laws regarding enforcement, understanding the court procedures, and realizing the avenues for change are the essential tools…

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

Science.gov (United States)

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

2015-09-23

In the retina, rod and cone photoreceptors form distinct connections with different classes of downstream bipolar cells. However, the molecular mechanisms responsible for their selective connectivity are unknown. Here we identify a cell-adhesion protein, ELFN1, to be essential for the formation of synapses between rods and rod ON-bipolar cells in the primary rod pathway. ELFN1 is expressed selectively in rods where it is targeted to the axonal terminals by the synaptic release machinery. At the synapse, ELFN1 binds in trans to mGluR6, the postsynaptic receptor on rod ON-bipolar cells. Elimination of ELFN1 in mice prevents the formation of synaptic contacts involving rods, but not cones, allowing a dissection of the contributions of primary and secondary rod pathways to retinal circuit function and vision. We conclude that ELFN1 is necessary for the selective wiring of rods into the primary rod pathway and is required for high sensitivity of vision. Copyright © 2015 Elsevier Inc. All rights reserved.

Control rod

International Nuclear Information System (INIS)

Igarashi, Takao; Yoshimoto, Yuichiro; Sugawara, Satoshi; Fukumoto, Takashi; Endo, Zen-ichiro; Saito, Shozo; Shinpo, Katsutoshi; Nishimura, Akira; Ozawa, Michihiro

1988-01-01

Purpose: To provide a sufficient shutdown margin upon reactor shutdown, prevent sheath deformation without decreasing neutron absorbents and prevent impact shocks exerted to structural materials. Constitution: The control rod of the present invention comprises a neutron absorption region, a sheath deformation means attached to the side wall and means for restricting and supporting axial movement of the neutron absorbent rod. Then, the amount of absorptive nuclei chained absorbents in the lower region is reduced than that in the upper region. In this way, effective neutron absorbing performance can be obtained relative to the neutron importance distribution during reactor shutdown. In addition, since the operationability is improved by reducing the weight of the control rod and the absorptive nuclei chained neutron abosrbers are used, mechanical nuclear life of the control rod can be increased. Thus, it is possible to prevent the outward deformation of the sheath, as well as prevent collision between the neutron absorber rod and the structural material on the side of inserting the control rod generated upon reactor scram by a simple structure. (Kamimura, M.)

Study of corium radial spreading between fuel rods in a PWR core

International Nuclear Information System (INIS)

Roche, S.; Gatt, J.M.

1996-01-01

In the framework of severe accident studies for PWR like Three Mile Island Unit 2 (TMI-2), the reactor core essentially constituted of fuel rods begins to heat and then to melt. During the early degradation phase, a melt (essentially UO2 and ZrO2) that constitutes the corium flows first along the rods, and after a blockage formation, may radially propagate towards the core periphery. A simplified model has been elaborated to study the corium freezing phenomena during its crossflow between the fuel rods. The corium spreads on an horizontal support made, of either a corium crust, or a grid assembly. The model solves numerically the interface energy balance equation at the solid-liquid corium interface and the monodimensional heat balance equation in transient process with convective terms and heat source (residual power). ''Zukauskas'' correlations are used to calculate heat transfer coefficients. The model can be integrated in severe accident codes like ICARE II (IPSN) describing the in-vessel degradation scenarios. (author). 5 refs, 10 figs

Data report of tight-lattice rod bundle thermal-hydraulic tests (2). Gap-width effect test using 37-rod bundle simulated water-cooled breeder reactor (Contract research)

International Nuclear Information System (INIS)

Tamai, Hidesada; Kureta, Masatoshi; Liu, Wei; Akimoto, Hajime; Sato, Takashi; Watanabe, Hironori; Ohnuki, Akira

2006-11-01

Japan Atomic Energy Agency has been performing tight-lattice rod bundle thermal-hydraulic tests to realize essential technologies for the technological and engineering feasibility of super high burn-up water-cooled breeder reactor featured by a high breeding ratio and super high burn-up by reducing the core water volume in water-cooled reactor. The tests are performing to make clear the fundamental subjects related to the boiling transition (BT) (Subjects: BT criteria under a highly tight-lattice rod bundle, effects of gap-width between rods and of rod-bowing) using 37-rod bundles (Base case test section (1.3mm gap-width), Two parameter effect test sections (Gap-width effect one (1.0mm) and Rod-bowing one)). In the present report, we summarize the test results from the gap-width effect test section. The thermal-hydraulic characteristics were obtained for the critical power under the steady-state and transient conditions, the pressure drop and the wall heat transfer within a wide range of pressure, flow rate, etc. including normal operational conditions of the designed reactor. Then the gap-width effects were also obtained from the comparison between the results using the base case test section and the gap-width effect one. (author)

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

Science.gov (United States)

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). 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

Control rod drives

International Nuclear Information System (INIS)

Futatsugi, Masao.

1980-01-01

Purpose: To secure the reactor operation safety by the provision of a fluid pressure detecting section for control rod driving fluid and a control rod interlock at the midway of the flow pass for supplying driving fluid to the control rod drives. Constitution: Between a driving line and a direction control valve are provided a pressure detecting portion, an alarm generating device, and a control rod inhibition interlock. The driving fluid from a driving fluid source is discharged by way of a pump and a manual valve into the reactor in which the control rods and reactor fuels are contained. In addition, when the direction control valve is switched and the control rods are inserted and extracted by the control rod drives, the pressure in the driving line is always detected by the pressure detection section, whereby if abnormal pressure is resulted, the alarm generating device is actuated to warn the abnormality and the control rod inhibition interlock is actuated to lock the direction control valve thereby secure the safety operation of the reactor. (Seki, T.)

Control rod drives

International Nuclear Information System (INIS)

Oonuki, Koji.

1981-01-01

Purpose: To increase the driving speed of control rods at rapid insertion with an elongate control rod and an extension pipe while ensuring sufficient buffering performance in a short buffering distance, by providing a plurality of buffers to an extension pipe between a control rod drive source and a control rod in LMFBR type reactor. Constitution: First, second and third buffers are respectively provided to an acceleration piston, an extension pipe and a control rod respectively and the insertion positions for each of the buffers are displaced orderly from above to below. Upon disconnection of energizing current for an electromagnet, the acceleration piston, the extension pipe and the control rod are rapidly inserted in one body. The first, second and third buffers are respectively actuated at each of their falling strokes upon rapid insertion respectively, and the acceleration piston, the extension pipe and the control rod receive the deceleration effect in the order correspondingly. Although the compression force is applied to the control rod only near the stroke end, it does not cause deformation. (Kawakami, Y.)

An investigation on diffusion bonding of aluminum to copper using equal channel angular extrusion process

OpenAIRE

Eslami, P.; Taheri, A. Karimi

2011-01-01

A new method for production of bimetallic rods, utilizing the equal channel angular extrusion (ECAE) process has been introduced before by previous researchers, but no attempt has been made to assess the effect of different temperatures and holding times in order to achieve a diffusional bond between the mating surfaces. In present research copper sheathed aluminum rods have been ECAEed at room temperature and subsequently held at a constant ECAE pressure, at different temperatures and holdin...

Control rod cluster with removable rods for nuclear fuel assembly

International Nuclear Information System (INIS)

Denizou, J.P.

1989-01-01

For each removable control rod, the open end section of the sleeve has a certain length of reduced diameter with openings in its wall. The top end of the rod is joined to an extension tube that surrounds the shaft over part of its lenght. This extension tube fits over the reduced part of the sleeve when the shaft is screwed into the bore of the sleeve. Rotation of the rod in the sleeve is prevented by deforming the extension tube locally in the openings of the end part of the sleeve. The rod is dismantled by exerting a torque on it using a gripping area near the end of the rod [fr

Experimental investigations of single and two-phase flow in a heated rod bundle

International Nuclear Information System (INIS)

Barthel, Frank; Franz, Ronald; Hampel, Uwe; Technische Univ. Dresden

2013-01-01

An experimental facility for the study of boiling flows in a 3 x 3 rod bundle geometry was setup. The bundle resembles in essential geometrical parts the geometry in a pressurized water reactor fuel element. The facility is operated with a refrigerant fluid. Beside standard instrumentation for temperature, pressure and flow rate we employed particle image velocimetry for single phase flow studies, gamma ray densitometry for integral gas fraction measurement sand ultrafast X-ray tomography for the study of the void dynamics in the cross-section. Moreover extensive thermo-instrumentation allows axial rod surface temperature measurements for the central heated rod. First results will be discussed in this article. (orig.)

Control rod drive mechanism

International Nuclear Information System (INIS)

Nakamura, Akira.

1981-01-01

Purpose: To ensure the scram operation of a control rod by the reliable detection for the position of control rods. Constitution: A permanent magnet is provided to the lower portion of a connecting rod in engagement with a control rod and a tube having a plurality of lead switches arranged axially therein in a predetermined pitch is disposed outside of the control rod drives. When the control rod moves upwardly in the scram operation, the lead switches are closed successively upon passage of the permanent magnet to operate the electrical circuit provided by way of each of the lead switches. Thus, the position for the control rod during the scram can reliably be determined and the scram characteristic of the control rod can be recognized. (Furukawa, Y.)

Method of inserting fuel rod

International Nuclear Information System (INIS)

Kamimoto, Shuji; Imoo, Makoto; Tsuchida, Kenji.

1991-01-01

The present invention concerns a method of inserting a fuel rod upon automatic assembling, automatic dismantling and reassembling of a fuel assembly in a light water moderated reactor, as well as a device and components used therefor. That is, a fuel rod is inserted reliably to an aimed point of insertion by surrounding the periphery of the fuel rod to be inserted with guide rods, and thereby suppressing the movement of the fuel rod during insertion. Alternatively, a fuel rod is inserted reliably to a point of insertion by inserting guide rods at the periphery of the point of insertion for the fuel rod to be inserted thereby surrounding the point of insertion with the guide rods or fuel rods. By utilizing fuel rods already present in the fuel assembly as the guide rods described above, the fuel rod can be inserted reliably to the point of insertion with no additional devices. Dummy fuel rods are previously inserted in a fuel assembly which are then utilized as the above-mentioned guide rods to accurately insert the fuel rod to the point of insertion. (I.S.)

Control rod drives

International Nuclear Information System (INIS)

Hayakawa, Hiroyasu.

1979-01-01

Purpose: To enable rapid control in a simple circuit by providing a motor control device having an electric capacity capable of simultaneously driving all of the control rods rapidly only in the inserting direction as well as a motor controlling device capable of fine control for the insertion and extraction at usual operation. Constitution: The control rod drives comprise a first motor control device capable of finely controlling the control rods both in inserting and extracting directions, a second motor control device capable of rapidly driving the control rods only in the inserting direction, and a first motor switching circuit and a second motor switching circuit switched by switches. Upon issue of a rapid insertion instruction for the control rods, the second motor switching circuit is closed by the switch and the second motor control circuit and driving motors are connected. Thus, each of the control rod driving motors is driven at a high speed in the inserting direction to rapidly insert all of the control rods. (Yoshino, Y.)

Process and equipment for pressure build-up in nuclear reactor fuel rods

International Nuclear Information System (INIS)

Heer, W.F.; Carli, E.V. de.

1976-01-01

The equipment makes possible the build-up of inert gas pressure in a filled and closed fuel can, i.e. in a complete fuel rod. Handling is simple, it is suitable for mass production and only causes low processing costs. The quality, e.g. the degree of purity of the contents of the rod, remains unchangedin processing. The equipment consists of a vacuum-tight space, into which the equally vacuum tight fuel rod is introduced, and can be fixed so that its position can be reproduced unmistakeably. The vacuum space contains a connection for the inert gases and a laser arrangement. After inserting a fuel rod into the facility, this is evacuated and the fuel can has a hole bored in it by a laser beam. After fast equalisation of pressure, an inert gas at the required pressure is introduced into the chamber and the fuel rod. After the filling process is completed, the fuel can is closed again with the same laser beam. The quality of the seal obtained, i.e the leak-tightness of the fuel can, can be checked after reduction of the inert gas pressure and before taking out the fuel rod, by repeated evacuation of the chamber. Laser light energies between 13,000 and 110,000 Joule/sq cm are sufficient. Optimum results were obtained for a Zircaloy fuel can with about 52,000 Joule/sq cm. (TK) [de

Laboratory demonstration of lightning strike pattern on different roof tops installed with Franklin Rods

Science.gov (United States)

Ullah, Irshad; Baharom, MNR; Ahmed, H.; Luqman, HM.; Zainal, Zainab

2017-11-01

Protection against lightning is always a challenging job for the researcher. The consequences due to lightning on different building shapes needs a comprehensive knowledge in order to provide the information to the common man. This paper is mainly concern with lightning pattern when it strikes on the building with different shape. The work is based on the practical experimental work in high voltage laboratory. Different shapes of the scaled structures have been selected in order to investigate the equal distribution of lightning voltage. The equal distribution of lightning voltage will provide the maximum probability of lightning strike on air terminal of the selected shapes. Building shapes have a very important role in lightning protection. The shapes of the roof tops have different geometry and the Franklin rod installation is also varies with changing the shape of the roof top. According to the ambient weather condition of Malaysia high voltage impulse is applied on the lightning rod installed on different geometrical shape. The equal distribution of high voltage impulse is obtained as the geometry of the scaled structure is identical and the air gap for all the tested object is kept the same. This equal distribution of the lightning voltage also proves that the probability of lightning strike is on the corner and the edges of the building structure.

Fission reactor control rod

International Nuclear Information System (INIS)

Irie, Tomoo.

1991-01-01

The present invention concerns a control rod in a PWR type reactor. A control rod has an inner cladding tube and an outer cladding tube disposed coaxially, and a water draining hole is formed at the inside of the inner cladding tube. Neutron absorbers are filled in an annular gap between the outer cladding tube and the inner cladding tube. The water draining hole opens at the lower end thereof to the top end of the control rod and at the upper end thereof to the side of the upper end plug of the control rod. If the control rod is dropped to a control rod guide thimble for reactor scram, coolants from the control rod guide thimble are flown from the lower end of the water draining hole and discharged from the upper end passing through the water draining hole. In this way, water from the control rod guide thimble is removed easily when the control rod is dropped. Further, the discharging amount of water itself is reduced by the provision of the water draining hole. Accordingly, sufficient control rod dropping speed can be attained. (I.N.)

Fuel rod simulator effects in flooding experiments single rod tests

International Nuclear Information System (INIS)

Nishida, M.

1984-09-01

The influence of a gas filled gap between cladding and pellet on the quenching behavior of a PWR fuel rod during the reflood phase of a LOCA has been investigated. Flooding experiments were conducted with a short length electrically heated single fuel rod simulator surrounded by glass housing. The gap of 0.05 mm width between the Zircaloy cladding and the internal Al 2 O 3 pellets of the rod was filled either wit helium or with argon to vary the radial heat resistance across the gap. This report presents some typical data and an evaluation of the reflood behavior of the fuel rod simulator used. The results show that the quench front propagates faster for increasing heat resistance in the gap between cladding and heat source of the rod. (orig.) [de

Maximum/minimum asymmetric rod detection

International Nuclear Information System (INIS)

Huston, J.T.

1990-01-01

This patent describes a system for determining the relative position of each control rod within a control rod group in a nuclear reactor. The control rod group having at least three control rods therein. It comprises: means for producing a signal representative of a position of each control rod within the control rod group in the nuclear reactor; means for establishing a signal representative of the highest position of a control rod in the control rod group in the nuclear reactor; means for establishing a signal representative of the lowest position of a control rod in the control rod group in the nuclear reactor; means for determining a difference between the signal representative of the position of the highest control rod and the signal representative of the position of the lowest control rod; means for establishing a predetermined limit for the difference between the signal representative of the position of the highest control rod and the signal representative of the position of the lowest control rod; and means for comparing the difference between the signals with the predetermined limit. The comparing means producing an output signal when the difference between the signals exceeds the predetermined limit

MreB Drives De Novo Rod Morphogenesis in Caulobacter crescentus via Remodeling of the Cell Wall

OpenAIRE

Takacs, Constantin N.; Poggio, Sebastian; Charbon, Godefroid; Pucheault, Mathieu; Vollmer, Waldemar; Jacobs-Wagner, Christine

2010-01-01

MreB, the bacterial actin-like cytoskeleton, is required for the rod morphology of many bacterial species. Disruption of MreB function results in loss of rod morphology and cell rounding. Here, we show that the widely used MreB inhibitor A22 causes MreB-independent growth inhibition that varies with the drug concentration, culture medium conditions, and bacterial species tested. MP265, an A22 structural analog, is less toxic than A22 for growth yet equally efficient for disrupting the MreB cy...

Morphogenic Protein RodZ Interacts with Sporulation Specific SpoIIE in Bacillus subtilis.

Science.gov (United States)

Muchová, Katarína; Chromiková, Zuzana; Bradshaw, Niels; Wilkinson, Anthony J; Barák, Imrich

2016-01-01

The first landmark in sporulation of Bacillus subtilis is the formation of an asymmetric septum followed by selective activation of the transcription factor σF in the resulting smaller cell. How the morphological transformations that occur during sporulation are coupled to cell-specific activation of transcription is largely unknown. The membrane protein SpoIIE is a constituent of the asymmetric sporulation septum and is a crucial determinant of σF activation. Here we report that the morphogenic protein, RodZ, which is essential for cell shape determination, is additionally required for asymmetric septum formation and sporulation. In cells depleted of RodZ, formation of asymmetric septa is disturbed and σF activation is perturbed. During sporulation, we found that SpoIIE recruits RodZ to the asymmetric septum. Moreover, we detected a direct interaction between SpoIIE and RodZ in vitro and in vivo, indicating that SpoIIE-RodZ may form a complex to coordinate asymmetric septum formation and σF activation. We propose that RodZ could provide a link between the cell shape machinery and the coordinated morphological and developmental transitions required to form a resistant spore.

Evaluation of droplet deposition in rod bundle

International Nuclear Information System (INIS)

Ji, W.; Gu, C.Y.; Anglart, H.

1997-01-01

Deposition model for droplets in gas droplet two-phase flow in rod bundle is developed in this work using the Lagrangian method. The model is evaluated in a 9-rod bundle geometry. The deposition coefficient in the bundle geometry are compared with that in round tube. The influences of the droplet size and gas mass flow rate on deposition coefficient are investigated. Furthermore, the droplet motion is studied in more detail by dividing the bundle channel into sub-channels. The results show that the overall deposition coefficient in the bundle geometry is close to that in the round tube with the diameter equal to the bundle hydraulic diameter. The calculated deposition coefficient is found to be higher for higher gas mass flux and smaller droplets. The study in the sub-channels show that the ratio between the local deposition coefficient for a sub-channel and the averaged value for the whole bundle is close to a constant value, deviations from the mean value for all the calculated cases being within the range of ±13%. (author)

A mathematical method for boiling water reactor control rod programming

International Nuclear Information System (INIS)

Tokumasu, S.; Hiranuma, H.; Ozawa, M.; Yokomi, M.

1985-01-01

A new mathematical programming method has been developed and utilized in OPROD, an existing computer code for automatic generation of control rod programs as an alternative inner-loop routine for the method of approximate programming. The new routine is constructed of a dual feasible direction algorithm, and consists essentially of two stages of iterative optimization procedures Optimization Procedures I and II. Both follow almost the same algorithm; Optimization Procedure I searches for feasible solutions and Optimization Procedure II optimizes the objective function. Optimization theory and computer simulations have demonstrated that the new routine could find optimum solutions, even if deteriorated initial control rod patterns were given

Control rod drives

International Nuclear Information System (INIS)

Hayakawa, Hiroyasu; Kawamura, Atsuo.

1979-01-01

Purpose: To reduce pellet-clad mechanical interactions, as well as improve the fuel safety. Constitution: In the rod drive of a bwr type reactor, an electric motor operated upon intermittent input such as of pulse signals is connected to a control rod. A resolver for converting the rotational angle of the motor to electric signals is connected to the rotational shaft of the motor and the phase difference between the output signal from the resolver and a reference signal is adapted to detect by a comparator. Based on the detection result, the controller is actuated to control a motor for control rod drive so that fine control for the movement of the control rod is made possible. This can reduce the moving distance of the control rod, decrease the thermal stress applied to the control rod and decrease the pellet clad mechanical interaction failures due to thermal expansion between the cladding tube and the pellets caused by abrupt changes in the generated power. (Furukawa, Y.)

Status of rod consolidation

International Nuclear Information System (INIS)

Bailey, W.J.

1985-04-01

Two of the factors that need to be taken into account with rod consolidation are (1) the effects on rods from their removal from the fuel assembly and (2) the effects on rods as a result of the consolidation process. Potential components of both factors are described in the report. Discussed under (1) are scratches on the fuel rod surfaces, rod breakage, crud, extended burnup, and possible cladding embrittlement due to hydrogen injection at BWRs. Discussed under (2) are the increased water temperature (less than 10 0 C) because of closer packing of the rods, formation of crevices between rods in the close-packed mode, contact with dissimilar metals, and the potential for rapid heating of fuel rods following the loss of water from a spent fuel storage pool. Another factor that plays an important role in rod consolidation is the cost of disposal of the nonfuel-bearing components of the fuel assembly. Also, the dose rate from the components - especially Inconel spacer grids - can affect the handling procedures. Several licensing issues that exist are described. A list of recommendations is provided. 98 refs., 5 figs., 5 tabs

Bacterial actin MreB assembles in complex with cell shape protein RodZ.

Science.gov (United States)

van den Ent, Fusinita; Johnson, Christopher M; Persons, Logan; de Boer, Piet; Löwe, Jan

2010-03-17

Bacterial actin homologue MreB is required for cell shape maintenance in most non-spherical bacteria, where it assembles into helical structures just underneath the cytoplasmic membrane. Proper assembly of the actin cytoskeleton requires RodZ, a conserved, bitopic membrane protein that colocalises to MreB and is essential for cell shape determination. Here, we present the first crystal structure of bacterial actin engaged with a natural partner and provide a clear functional significance of the interaction. We show that the cytoplasmic helix-turn-helix motif of Thermotoga maritima RodZ directly interacts with monomeric as well as filamentous MreB and present the crystal structure of the complex. In vitro and in vivo analyses of mutant T. maritima and Escherichia coli RodZ validate the structure and reveal the importance of the MreB-RodZ interaction in the ability of cells to propagate as rods. Furthermore, the results elucidate how the bacterial actin cytoskeleton might be anchored to the membrane to help constrain peptidoglycan synthesis in the periplasm.

Control rod drive

International Nuclear Information System (INIS)

Okutani, Tetsuro.

1988-01-01

Purpose: To provide a simple and economical control rod drive using a control circuit requiring no pulse circuit. Constitution: Control rods in a BWR type reactor are driven by hydraulic pressure and inserted or withdrawn in the direction of applying the hydraulic pressure. The direction of the hydraulic pressure is controlled by a direction control valve. Since the driving for the control rod is extremely important in view of the operation, a self diagnosis function is disposed for rapid inspection of possible abnormality. In the present invention, two driving contacts are disposed each by one between the both ends of a solenoid valve of the direction control valve for driving the control rod and the driving power source, and diagnosis is conducted by alternately operating them. Therefore, since it is only necessary that the control circuit issues a driving instruction only to one of the two driving contacts, the pulse circuit is no more required. Further, since the control rod driving is conducted upon alignment of the two driving instructions, the reliability of the control rod drive can be improved. (Horiuchi, T.)

Measurement and analysis of flow wall shear stress in an interior subchannel of triangular array rods

International Nuclear Information System (INIS)

Fakori-Monazah, M.R.; Todreas, N.E.

1977-08-01

A simulated model of triangular array rods with pitch to diameter ratio of 1.10 (as a test section) and air as the fluid flow was used to study the LMFBR hydraulic parameters. The wall shear stress distribution around the rod periphery, friction factors, static pressure distributions and turbulence intensity corresponding to various Reynolds numbers ranging from 4140 to 36170 in the central subchannel were measured. Various approaches for measurement of wall shear stress were compared. The measurement was performed using the Preston tube technique with the probe outside diameter equal to 0.014 in

Control rod withdrawal monitoring device

International Nuclear Information System (INIS)

Ebisuya, Mitsuo.

1984-01-01

Purpose: To prevent the power ramp even if a plurality of control rods are subjected to withdrawal operation at a time, by reducing the reactivity applied to the reactor. Constitution: The control rod withdrawal monitoring device is adapted to monitor and control the withdrawal of the control rods depending on the reactor power and the monitoring region thereof is divided into a control rod group monitoring region a transition region and a control group monitoring not interfere region. In a case if the distance between a plurality of control rods for which the withdrawal positions are selected is less than a limiting value, the coordinate for the control rods, distance between the control rods and that the control rod distance is shorter are displayed on a display panel, and the withdrawal for the control rods are blocked. Accordingly, even if a plurality of control rods are subjected successively to the withdrawal operation contrary to the control rod withdrawal sequence upon high power operation of the reactor, the power ramp can be prevented. (Kawakami, Y.)

Study on the improved evaluation of radioactivity of activated control rods in PWR

International Nuclear Information System (INIS)

Waki, Toshikazu; Yamada, Motoyuki; Horikawa, Yoshihiko; Miyake, Yusuke; Sakashita, Akira

2009-01-01

The evaluation method of radioactivity of activated materials has been developed as ORIGEN code. However, it is difficult to precisely evaluate the radioactivity of neutron absorption materials such as control rods. A control rod in PWR is made of Ag-In-Cd alloy that absorbs neutron greatly and the thermal neutron flux decreases rapidly in and around it. This phenomenon is called depression effect. The consideration of depression effect is necessary to evaluate radioactivity of the control rod. In this study we improved the reliability of the cross-section value of Ag-107(n,γ) Ag-108m by the irradiation examination in JRR3. In addition, we calculated (1) the neutron spectrum and neutron flux with depression effect by MCNP of Monte Carlo method and (2) the radioactivity of the activated control rod. The pieces of control rod were irradiated at JMTR of JAERI. As a result of the accuracy of the measurement data calculation results, we developed the method of evaluation for the radioactivity of activated control rod. The radioactivity of activated control rod in PWR was evaluated and compared with the measurement data, resulting in positive accuracy. Of special significance was confirmation of the value of Ag-108m, as an essential nuclide for long term dose estimation of disposal facility. The cross-section value of Ag-107(n,γ) Ag-108m was about one forty of existent library. This method was accurately confirmed and developed for evaluating activated control rods reasonably. (author)

Control rod effects with plutonium recycle in a PWR

International Nuclear Information System (INIS)

Nash, G.; Muehl, G.J.; Gibson, I.H.

1979-03-01

A study has been made on a PWR loaded partly and wholly with plutonium to determine the changes in shutdown margin compared with an enriched uranium core. Lattice calculations are used to generate cell constants for core calculations. Three fuel loadings were considered, all uranium, 30% (approximately) of the assemblies plutonium in natural uranium, and all plutonium. The equilibrium fuel management schemes adopted in each case are based on the standard three cycle equal size batch scheme. Detailed calculations of power and irradiation distributions through the cycles have been carried out to provide a starting point for the control rod worth and requirement calculations. Control rod worths are reduced in a plutonium core because of the harder spectrum and higher fuel absorption cross sections. Furthermore, the control rod requirements for shutdown increase because of the increase in fuel and moderator temperature coefficients. This results in a reduction in shutdown margin. The magnitude of these changes is fully analysed in the report. The significance of these reductions depends on the detail of the safety argument but reductions of these sizes are unlikely to be acceptable. The data provided in this report could be used to give a first estimate of the plutonium loading acceptable given the safety assessment of the normal uranium core. (U.K.)

To the problem of interchannel mixing of a coolant in finned-rod bundles

International Nuclear Information System (INIS)

Dzyubenko, B.V.

1979-01-01

Suggested is the methods of experimental data processing on interchannel mixing in finned-rod bundles, based on the use of P-theorem of similarity and dimension theory and physical considerations. The obtained general conclusions of dependence for calculation of effective diffusion coefficients and coefficients of interchannel mixing permit to lock the systems of differential equations, describina the flow of homogenized media and the flow in a real rod bundle correspondingly. The carried out analysis of the experimental data obtained by different authors, who applied the method of central rod heating during the investigation of interchannel change, permitted to find essential influence of the suggested similarity criteria on mixing coefficient, characterizing the effect of centrifugal forces on the flow and a reformed longitudinal coordinate

Experimental investigation of the turbulent flow through a wall subchannel of a rod bundle

International Nuclear Information System (INIS)

Rehme, K.

1977-04-01

An experimental investigation was performed to establish reliable information on the transport properties of turbulent flow through subchannels of rod bundles. Detailed data were measured of the distributions of the time-mean velocity, the turbulence intensities in all directions and, thus, the kinetic energy of turbulence, of the shear stresses in the directions normal and parallel to the walls, and of the wall shear stresses for a wall subchannel of a rod bundle of four rods in parallel. The pitch-to-diameter ratio of the rods equal to the wall-to-diameter ratio was 1.07, the Reynolds number of this investigation was Re = 8.7 x 10 4 . On the basis of the data measured the eddy viscosities in the directions normal and parallel to the walls were calculated. Thus, detailed data of the eddy viscosities in direction parallel to the walls in rod bundels were obtained for the first time. The experimental results were compared with predictions by the VELASCO-code. There are considerable differences between calculated and measured data of the time-mean velocity and the wall shear stresses. Attempts to adjust the VELASCO-code against the measurements were not successful. The reasons of the discrepancies are discussed. (orig.) [de

Mechanical Properties Analysis of an Al-Mg Alloy Connecting Rod with Submicrometric Structure

Directory of Open Access Journals (Sweden)

Javier León

2015-07-01

Full Text Available Over these last few years, there has been a growing interest in developing mechanical components from submicrometric materials due to the significant improvement that these materials present compared to their original state. This present research work deals with the study of the mechanical properties of a connecting rod isothermally forged from different starting materials. These materials are as follows: annealed aluminum alloy (AA 5754, the same alloy previously deformed through equal channel angular pressing (ECAP and a third case where the previously ECAP-processed material is subjected to a recovery heat treatment. A comparison is made between finite volume (FV simulations and experimental tests with respect to hardness, plastic strain and forging force. Furthermore, the improvement in the mechanical properties of the connecting rod forged from predeformed material is evaluated in comparison to the connecting rod forged with annealed material. The microstructure of both cases is also compared at the end of the manufacturing process.

Nuclear fuel rod loading apparatus

International Nuclear Information System (INIS)

King, H.B.; Macivergan, R.; Mckenzie, G.W.

1980-01-01

An apparatus incorporating a microprocessor control is provided for automatically loading nuclear fuel pellets into fuel rods commonly used in nuclear reactor cores. The apparatus comprises a split ''v'' trough for assembling segments of fuel pellets in rows and a shuttle to receive the fuel pellets from the split ''v'' trough when the two sides of the split ''v'' trough are opened. The pellets are weighed while in the shuttle, and the shuttle then moves the pellets into alignment with a fuel rod. A guide bushing is provided to assist the transfer of the pellets into the fuel rod. A rod carousel which holds a plurality of fuel rods presents the proper rod to the guide bushing at the appropriate stage in the loading sequence. The bushing advances to engage the fuel rod, and the shuttle advances to engage the guide bushing. The pellets are then loaded into the fuel rod by a motor operated push rod. The guide bushing includes a photocell utilized in conjunction with the push rod to measure the length of the row of fuel pellets inserted in the fuel rod

Control rod shutdown system

International Nuclear Information System (INIS)

Miyamoto, Yoshiyuki; Higashigawa, Yuichi.

1996-01-01

The present invention provides a control rod terminating system in a BWR type nuclear power plant, which stops an induction electric motor as rapidly as possible to terminate the control rods. Namely, the control rod stopping system controls reactor power by inserting/withdrawing control rods into a reactor by driving them by the induction electric motor. The system is provided with a control device for controlling the control rods and a control device for controlling the braking device. The control device outputs a braking operation signal for actuating the braking device during operation of the control rods to stop the operation of the control rods. Further, the braking device has at least two kinds of breaks, namely, a first and a second brakes. The two kinds of brakes are actuated by receiving the brake operation signals at different timings. The brake device is used also for keeping the control rods after the stopping. Even if a stopping torque of each of the breaks is small, different two kinds of brakes are operated at different timings thereby capable of obtaining a large stopping torque as a total. (I.S.)

Fuel rod leak detector

International Nuclear Information System (INIS)

Womack, R.E.

1978-01-01

A typical embodiment of the invention detects leaking fuel rods by means of a radiation detector that measures the concentration of xenon-133 ( 133 Xe) within each individual rod. A collimated detector that provides signals related to the energy of incident radiation is aligned with one of the ends of a fuel rod. A statistically significant sample of the gamma radiation (γ-rays) that characterize 133 Xe is accumulated through the detector. The data so accumulated indicates the presence of a concentration of 133 Xe appropriate to a sound fuel rod, or a significantly different concentration that reflects a leaking fuel rod

Fuel rods

International Nuclear Information System (INIS)

Hattori, Shinji; Kajiwara, Koichi.

1980-01-01

Purpose: To ensure the safety for the fuel rod failures by adapting plenum springs to function when small forces such as during transportation of fuel rods is exerted and not to function the resilient force when a relatively great force is exerted. Constitution: Between an upper end plug and a plenum spring in a fuel rod, is disposed an insertion member to the lower portion of which is mounted a pin. This pin is kept upright and causes the plenum spring to function resiliently to the pellets against the loads due to accelerations and mechanical vibrations exerted during transportation of the fuel rods. While on the other hand, if a compression force of a relatively high level is exerted to the plenum spring during reactor operation, the pin of the insertion member is buckled and the insertion member is inserted to the inside of the plenum spring, whereby the pellets are allowed to expand freely and the failures in the fuel elements can be prevented. (Moriyama, K.)

Investigating The Integral Control Rod Worth Of The Miniature Neutron Source Reactor MNSR

International Nuclear Information System (INIS)

Nguyen Hoang Hai; Do Quang Binh

2011-01-01

Determining control rod characteristics is an essential problem of nuclear reactor analysis. In this research, the integral control rod worth of the miniature neutron source reactor MNSR is investigated. Some other parameters of the nuclear reactor, such as core excess reactivity, shut down margin, are also calculated. Group constants for all reactor components are generated by the WIMSD code and then are used in the CITATION code to solve the neutron diffusion equations. The maximum relative error of the calculated results compared with the measurement data is about 3.5%. (author)

Multiple fuel rod gripper

International Nuclear Information System (INIS)

Shields, E.P.

1987-01-01

An apparatus is described for gripping an array of rods comprising: (a) gripping members grippingly engageable with the rods, each of which has a hollow portion terminating in an open end for receiving the end of one of the rods; (b) a closing means for causing the hollow portion of each of the gripping members to apply substantially the same gripping force onto the end of its respective rod, including (i) a locking plate having a plurality of tapered holes registrable with the array of rods, wherein the exterior of each of the gripping members is tapered and nested within one of the tapered holes, (ii) a withdrawing means having a hydraulic plunger operatively connected to each of the gripping members for applying a substantially identical withdrawing force on each of the gripping members, whereby the hollow portion of each of the gripping members applies substantially the same gripping force on its respective rod, and (c) means for detecting whether each of the gripping members has grippingly engaged its respective rod

High-yield production of hydrophobins RodA and RodB from Aspergillus fumigatus in Pichia pastoris

DEFF Research Database (Denmark)

Pedersen, Mona Højgaard; Borodina, Irina; Moresco, Jacob Lange

2011-01-01

A as well as rRodB were able to convert a glass surface from hydrophilic to hydrophobic similar to native RodA, but only rRodB was able to decrease the hydrophobicity of a Teflon-like surface to the same extent as native RodA, while rRodA showed this ability to a lesser extent. Recombinant RodA and native...

Cone dystrophy with "supernormal" rod ERG: psychophysical testing shows comparable rod and cone temporal sensitivity losses with no gain in rod function.

Science.gov (United States)

Stockman, Andrew; Henning, G Bruce; Michaelides, Michel; Moore, Anthony T; Webster, Andrew R; Cammack, Jocelyn; Ripamonti, Caterina

2014-02-10

We report a psychophysical investigation of 5 observers with the retinal disorder "cone dystrophy with supernormal rod ERG," caused by mutations in the gene KCNV2 that encodes a voltage-gated potassium channel found in rod and cone photoreceptors. We compared losses for rod- and for cone-mediated vision to further investigate the disorder and to assess whether the supernormal ERG is associated with any visual benefit. L-cone, S-cone, and rod temporal acuity (critical flicker fusion frequency) were measured as a function of target irradiance; L-cone temporal contrast sensitivity was measured as a function of temporal frequency. Temporal acuity measures revealed that losses for vision mediated by rods, S-cones, and L-cones are roughly equivalent. Further, the gain in rod function implied by the supernormal ERG provides no apparent benefit to near-threshold rod-mediated visual performance. The L-cone temporal contrast sensitivity function in affected observers was similar in shape to the mean normal function but only after the mean function was compressed by halving the logarithmic sensitivities. The name of this disorder is potentially misleading because the comparable losses found across rod and cone vision suggest that the disorder is a generalized cone-rod dystrophy. Temporal acuity and temporal contrast sensitivity measures are broadly consistent with the defect in the voltage-gated potassium channel producing a nonlinear distortion of the photoreceptor response but after otherwise normal transduction processes.

Control rod drive for vertical movement

International Nuclear Information System (INIS)

Suskov, I.I.; Gorjunov, V.S.; Zajcev, B.I.; Derevjankin, N.E.; Petrov, V.A.; Istomin, S.D.; Kovalencik, D.I.; Archipov, E.A.; Serebrjakov, V.I.; Kacalin, V.S.

1982-01-01

The control of the rod repositioning gear unit and the control unit of the profile grab of the control rod drive for the alkali metal-cooled fast breeder reactor is achieved by an electromotor being arranged outside the hermetic drive casing. The guide tube is directly repositioned by the rod repositioning gear unit. Coupling control of the drive with the control rod is done in the lower operative position of the control rod and that because of the interaction of the tie rod arranged on the spring-mounted control rod with the induction transmitter for the lower position of the control rod. In the transfer position the rod is fixed within the guide tube. (orig.)

Safety rod driving device

International Nuclear Information System (INIS)

Murakami, Kiyonobu; Kurosaki, Akira.

1988-01-01

Purpose: To rapidly insert safety rods for a criticality experiment device into a reactor core container to stop the criticality reaction thereby prevent reactivity accidents. Constitution: A cylinder device having a safety rod as a cylinder rod attached with a piston at one end is constituted. The piston is elevated by pressurized air and attracted and fixed by an electromagnet which is a stationary device disposed at the upper portion of the cylinder. If the current supply to the electromagnet is disconnected, the safety rod constituting the cylinder rod is fallen together with the piston to the lower portion of the cylinder. Since the cylinder rod driving device has neither electrical motor nor driving screw as in the conventional device, necessary space can be reduced and the weight is decreased. In addition, since the inside of the nuclear reactor can easily be shielded completely from the external atmosphere, leakage of radioactive materials can be prevented. (Horiuchi, T.)

A new physics design of control safety rods for prototype fast breeder reactor

International Nuclear Information System (INIS)

Devan, K.; Riyas, A.; Alagan, M.; Mohanakrishnan, P.

2008-01-01

The absorber rods of 500 MWe prototype fast breeder reactor (PFBR), which is under construction at Kalpakkam, have been designed to provide sufficient shutdown margin during normal and accidental conditions for ensuring the safe shut down. There are nine control and safety rods (CSR) and 3 diverse safety rods (DSR). Absorber material used is initially 65% enriched B 4 C. Based on the reported experiments in PHENIX reactor and design of absorber rods in SUPERPHENIX, the design of CSR is modified by introducing 20 cm length natural B 4 C at the top and bottom of absorber column and maintaining the remaining portion with 65% enriched B 4 C. This design ensures sufficient shutdown margin (SDM) during normal operation and also during the one stuck rod condition. For comparison of the above two designs, a CSR of 57% of enrichment was considered which gives the same worth as the revised CSR design with natural B 4 C sections in top and bottom. There is significant savings in the initial inventory of enriched B 4 C for CSR. The annual requirement of enriched boron also reduces. This new CSR can last for about 5 cycles, based on its clad life. But, it is planned to be replaced after every 3 cycles (1 cycle equals 180 efpd) of operation due to radiation damage effects in hexcan D9 steel. Use of ferritic steel for hexcan can extend the life of CSR to 5 cycles

Testing device for control rod drives

International Nuclear Information System (INIS)

Hayakawa, Toshifumi.

1992-01-01

A testing device for control rod drives comprises a logic measuring means for measuring an output signal from a control rod drive logic generation circuit, a control means for judging the operation state of a control rod and a man machine interface means for outputting the result of the judgement. A driving instruction outputted from the control rod operation device is always monitored by the control means, and if the operation instruction is stopped, a testing signal is outputted to the control rod control device to simulate a control rod operation. In this case, the output signal of the control rod drive logic generation circuit is held in a control rod drive memory means and intaken into a logic analysis means for measurement and an abnormality is judged by the control means. The stopping of the control rod drive instruction is monitored and the operation abnormality of the control rod is judged, to mitigate the burden of an operator. Further, the operation of the control rod drive logic generation circuit can be confirmed even during a nuclear plant operation by holding the control rod drive instruction thereby enabling to improve maintenance efficiency. (N.H.)

Turbulent flow through a wall subchannel of a rod bundle

International Nuclear Information System (INIS)

Rehme, K.

1978-04-01

The turbulent flow through a wall subchannel of a rod bundle was investigated experimentally by means of hotwires und Pitot-tubes. The aim of this investigation was to get experimental information on the transport properties of turbulent flow especially on the momentum transport. Detailed data were measured of the distributions of the time-mean velocity, the turbulence intensities and, hence the kinetic of turbulence, of the shear stresses in the directions normal and parallel to the walls, and of the wall shear stresses. The pitch-to-diameter ratio of the rods equal to the wall-to-diameter ratio was 1.15, the Reynolds number of this investigation was Re = 1.23.10 5 . On the basis of the measurements the eddy viscosities normal and parallel to the walls were calculated. The eddy viscosities observed showed a considerable deviation from the data known up-to-now and from the assumptions introduced in the codes. (orig.) [de

Failed fuel rod detector

Energy Technology Data Exchange (ETDEWEB)

Uchida, Katsuya; Matsuda, Yasuhiko

1984-05-02

The purpose of the project is to enable failed fuel rod detection simply with no requirement for dismantling the fuel assembly. A gamma-ray detection section is arranged so as to attend on the optional fuel rods in the fuel assembly. The fuel assembly is adapted such that a gamma-ray shielding plate is detachably inserted into optional gaps of the fuel rods or, alternatively, the fuel assembly can detachably be inserted to the gamma-ray shielding plate. In this way, amount of gaseous fission products accumulated in all of the plenum portions in the fuel rods as the object of the measurement can be determined without dismantling the fuel assembly. Accordingly, by comparing the amounts of the gaseous fission products, the failed fuel rod can be detected.

RodPilotR - The Innovative and Cost-Effective Digital Control Rod Drive Control System for PWRs

International Nuclear Information System (INIS)

Baron, Clemens

2008-01-01

With RodPilot, AREVA NP offers an innovative and cost-effective system for controlling control rods in Pressurized Water Reactors. RodPilot controls the three operating coils of the control rod drive mechanism (lift, moveable gripper and stationary gripper coil). The rods are inserted into or withdrawn from the core as required by the Reactor Control System. The system combines modern components, state-of-the-art logic and a proven electronic control rod drive control principle to provide enhanced reliability and lower maintenance costs. (author)

Integrated control rod monitoring device

International Nuclear Information System (INIS)

Saito, Katsuhiro

1997-01-01

The present invention provides a device in which an entire control rod driving time measuring device and a control rod position support device in a reactor building and a central control chamber are integrated systematically to save hardwares such as a signal input/output device and signal cables between boards. Namely, (1) functions of the entire control rod driving time measuring device for monitoring control rods which control the reactor power and a control rod position indication device are integrated into one identical system. Then, the entire devices can be made compact by the integration of the functions. (2) The functions of the entire control rod driving time measuring device and the control rod position indication device are integrated in a central operation board and a board in the site. Then, the place for the installation of them can be used in common in any of the cases. (3) The functions of the entire control rod driving time measuring device and the control rod position indication device are integrated to one identical system to save hardware to be used. Then, signal input/output devices and drift branching panel boards in the site and the central operation board can be saved, and cables for connecting both of the boards is no more necessary. (I.S.)

Control rods

International Nuclear Information System (INIS)

Koga, Isao; Masuoka, Ryuzo.

1979-01-01

Purpose: To prevent fuel element failures during power conditioning by removing liquid absorbents in poison tubes of control rods in a fast power up step and extracting control rods to slightly increase power in a medium power up step. Constitution: A plurality of poison tubes are disposed in a coaxial or plate-like arrangement and divided into a region capable of compensating the reactivity from the initial state at low temperature to 40% power operation and a region capable of compensating the reactivity in the power up above 40% power operation. Soluble poisons are used as absorbers in the poison tubes corresponding to above 40% power operation and they are adapted to be removed independently from the driving of control rods. The poison tubes filled with the soluble absorbers are responsible for the changes in the reactivity from the initial state at low temperature to the medium power region and the reactivity control is conducted by the elimination of liquid absorbers from the poison tubes. In the succeeding slight power up region above the medium power, power up is proceeding by extracting the control rods having remaining poison tubes filled with solid or liquid absorbers. (Seki, T.)

Simulation of nuclear fuel rods by using process computer-controlled power for indirect electrically heated rods

International Nuclear Information System (INIS)

Malang, S.

1975-11-01

An investigation was carried out to determine how the simulation of nuclear fuel rods with indirect electrically heated rods could be improved by use of a computer to control the electrical power during a loss-of-coolant accident (LOCA). To aid in the experiment, a new version of the HETRAP code was developed which simulates a LOCA with heater rod power controlled by a computer that adjusts rod power during a blowdown to minimize the difference in heat flux of the fuel and heater rods. Results show that without computer control of heater rod power, only the part of a blowdown up to the time when the heat transfer mode changes from nucleate boiling to transition or film boiling can be simulated well and then only for short times. With computer control, the surface heat flux and temperature of an electrically heated rod can be made nearly identical to that of a reactor fuel rod with the same cooling conditions during much of the LOCA. A small process control computer can be used to achieve close simulation of a nuclear fuel rod with an indirect electrically heated rod

Reactor core with rod-shaped fuel cells

International Nuclear Information System (INIS)

Dworak, A.

1976-01-01

The proposal refers to the optimization of the power distribution in a reactor core which is provided with several successive rod-shaped fuel cells. A uniform power output - especially in radial direction - is aimed at. This is achieved by variation of the dwelling periods of the fuel cells, which have, for this purpose, a fuel mixture changing from layer to layer. The fuel cells with the shortest dwelling period are arranged near the coolant inlet side of the reactor core. The dwelling periods of the fuel cells are adapted to the given power distribution. As neighboring cells have equal dwelling periods, the exchange can be performed much easier then with the composition currently known. (UWI) [de

Fuel rod technology

International Nuclear Information System (INIS)

Bezold, H.; Romeiser, H.J.

1979-07-01

By extensive mechanization and automation of the fuel rod production, also at increasing production numbers, an efficient production shall be secured, simultaneously corresponding to the high quality standard of the fuel rods. The works done up to now concentrated on the lay out of a rough concept for a mechanized production course. Detail-studies were made for the problems of fuel rod humidity, filling and resistance welding. Further promotion of this project and thus further report will be stopped, since the main point of these works is the production technique. (orig.) [de

Control rod guide tube assembly

International Nuclear Information System (INIS)

Jabsen, F.S.

1982-01-01

An improved fuel assembly is described as consisting of a sleeve that engages one end of a control rod guide tube essentially fixing the guide tube to one of the fuel assembly end structures. The end of the sleeve protrudes above the surface of the end fitting. The outer surface of the sleeve has a peripheral groove that engages the resilient sides of a cellular grid or lattice shaped lock. This lock fixes the sleeve in position between the various elements that comprise the end fitting, thereby eliminating a profusion of costly and potentially troublesome nuts, threaded studs and the like that are frequently employed in the fuel assemblies that are presently in use

Digital control rod blocking monitor

International Nuclear Information System (INIS)

Funayama, Yoshio.

1996-01-01

The present invention system is used for monitoring of a power region of a reactor, and used for monitoring of simultaneous withdrawal of a plurality of control rods without increasing the size or complicating the system. Namely, the system processes signals from a neutron flux detectors at the periphery of control rods controlled for withdrawal. As a result of the processing, the digital monitoring system generates an alarm when the reactor power at the periphery of the control rods fluctuates exceeding an allowable range. In the system, a control rod information forming means prepares frame data comprising front data, positions of the control rods to be withdrawn, frame numbers and completion data. A serial data transmitting means transmits the frame data successively as repeating frame data rows. A control rod information receiving means takes up the frame data of each of control rods to be withdrawn from the transmitted frame data rows. Since the system of the present invention can monitor the withdrawal of a plurality of control rods simultaneously without increasing the size or complicating the system, cost can be saved and the maintenance can be improved. (I.S.)

Evolution of fuel rod support under irradiation impact on the mechanical behaviour of fuel assemblies

International Nuclear Information System (INIS)

Billerey, Antoine; Waeckel, Nicolas

2005-01-01

New fuel management targets imply to increase fuel assembly discharge burnup. Therefore, the prediction of the mechanical behaviour of the irradiated fuel assembly is essential such as excessive fuel assembly distortion induce incomplete Rod Cluster Control Assembly insertion problems (safety issue) or fuel rod vibration induced wear leading to leaking rods (plant operation problems). Within this framework, one of the most important parameter is the knowledge of the fuel rod support in the grid cell because it directly governs the mechanical behaviour of the fuel assembly and consequently allows to predict the behaviour of irradiated structures in terms of (1) axial and lateral deformation (global behaviour of the assembly) and (2) rod vibration induced wear (local behaviour of the rod). Generally, fuel rod support is provided by a spring-dimple system fixed to the grid. During irradiation, the spring force decreases and a gap between the rod and the spring may occur. This phenomenon is due to (1) stress relieving in the spring and in the dimples, (2) grid growth and (3) reduction of the rod diameter. Two models have been developed to predict the behaviour of the rod in the cell. The first model is dedicated to the evaluation of the spring force relaxation during irradiation. The second one can assess the rotation characteristic of the fuel rod in the cell, function of the spring force. The main input parameters are (1) the creep laws of the grid materials, (2) the growth law of the grid, (3) the evolution of rod diameter and (4) the design of the fuel rod support. The aim of this paper is to: (1) evaluate the consequences of grid support design modifications on the rod vibration sensitivity in terms of predicted rod to grid maximum gap during irradiation and time in operation with an open rod to grid gap, (2) evaluate, using a linear or non-linear Finite Element assembly model, the impact of the evolution of grid support under irradiation on the overall mechanical

Measurement of the anti reactivity of a control rod of G1, by a slow oscillation method

International Nuclear Information System (INIS)

Breton, D.; Leroy, J.; Vidal, R.

1957-01-01

It is possible to determine the effect of the end of a control rod on the reactivity of the pile by measuring the modulation induced in the neutron flux by the slow oscillation of this control rod. The total effect of the control rod can be deduced, given certain hypothesis and corrections, from the experimental curve giving the effect of the end of the rod as a function of its position. This method has the advantage of permitting the measurement of very large anti reactivities, such as p= 10 -2 for example, which would not be possible by other kinetic methods. Thus the control rod B 3 , in the low position, brings about a reduction in reactivity equal to 1130 p.c.m. ± 30 in the pile charged with 518 fuel elements, on one side only of the slit. We have compared the oscillation method with the classical divergence method, in the fields where the two measurements were possible: a satisfactory agreement was found. We have established that the phase displacement between the oscillation of the rod and the modulation of the flux varied greatly with the position of the rod. This variation cannot be explained on the basis of the dynamic model independent of space; we have attributed it to the influence of spatial harmonics of the flux distribution, and have determined a correction which frees the measurements of this influence. (author) [fr

Research on TIG weld machine of the upper side ring slot of Gd-rod technology reconstruct

International Nuclear Information System (INIS)

Fang Shixiang; Lan Zhibing; Cui Quhu

2010-01-01

The research on TIG weld machine of the upper side ring slot of Gd-rod existent matter: seal electrical source got up difficulty; control system had graveness aging; space between was adjusted precision lowness; welding torch lay mode and structure were not in reason. carried through all around technology reconstruct: had chosen the best of TIG weld machine; designed ignite arc device, designed optics imaging device, designed tungsten mighty axis direction auto conditioning, was provided with arc slot, adopted PLC to control the whole system and realization auto control. After TIG weld machine of the upper side ring slot of Gd-rod technology reconstruct research , provided with arc slot the first time in the Gd-rod of nuclear fuel, optimized the weld technics, improved welding line melt width and deep equality, stability, and great breadth advanced nuclear fuel product line technology and throughput. (authors)

The BWR Hybrid 4 control rod

International Nuclear Information System (INIS)

Gross, H.; Fuchs, H.P.; Lippert, H.J.; Dambietz, W.

1988-01-01

The service life of BWR control rods designed in the past has been unsatisfactory. The main reason was irradiation assisted stress corrosion cracking of B 4 C rods caused by external swelling of the B 4 C powder. By this reason KWU developed an improved BWR control rod (Hybrid 4 control rod) with extended service life and increased control rod worth. It also allows the procedure for replacing and rearranging fuel assemblies to be considerably simplified. A complete set of Hydbrid 4 control rods is expected to last throughout the service life of a plant (assumption: ca. 40 years) if an appropriate control rod reshuffling management program is used. (orig.)

Burnable poison rod

International Nuclear Information System (INIS)

Natsume, Tomohiro.

1988-01-01

Purpose: To decrease the effect of water elimination and the effect of burn-up residue boron, thereby reduce the effect of burnable poison rods as the neutron poisons at the final stage of reactor core lifetime. Constitution: In a burnable poison rod according to the present invention, a hollow burnable poison material is filled in an external fuel can, an inner fuel can mounted with a carbon rod is inserted to the hollow portion of the burnable poison material and helium gases are charged in the outer fuel can. In such a burnable poison rod, the reactivity worths after the burning are reduced to one-half as compared with the conventional case. Accordingly, since the effect of the burnable poison as the neutron poisons is reduced at the final stage of the reactor core of lifetime, the excess reactivity of the reactor core is increased. (Horiuchi, T.)

Biomechanics of lumbar cortical screw-rod fixation versus pedicle screw-rod fixation with and without interbody support.

Science.gov (United States)

Perez-Orribo, Luis; Kalb, Samuel; Reyes, Phillip M; Chang, Steve W; Crawford, Neil R

2013-04-15

Seven different combinations of posterior screw fixation, with or without interbody support, were compared in vitro using nondestructive flexibility tests. To study the biomechanical behavior of a new cortical screw (CS) fixation construct relative to the traditional pedicle screw (PS) construct. The CS is an alternative to the PS for posterior fixation of the lumbar spine. The CS trajectory is more sagittally and cranially oriented than the PS, being anchored in the pars interarticularis. Like PS fixation, CS fixation uses interconnecting rods fastened with top-locking connectors. Stability after bilateral CS fixation was compared with stability after bilateral PS fixation in the setting of intact disc and with direct lateral interbody fixation (DLIF) or transforaminal lateral interbody fixation (TLIF) support. Standard nondestructive flexibility tests were performed in cadaveric lumbar specimens, allowing non-paired comparisons of specific conditions from 28 specimens (4 groups of 7) within a larger experiment of multiple hardware configurations. Condition tested and group from which results originated were as follows: (1) intact (all groups); (2) with L3-L4 bilateral PS-rods (group 1); (3) with bilateral CS-rods (group 2); (4) with DLIF (group 3); (5) with DLIF + CS-rods (group 4); (6) with DLIF + PS-rods (group 3); (7) with TLIF + CS-rods (group 2), and (8) with TLIF + PS-rods (group 2). To assess spinal stability, the mean range of motion, lax zone, and stiff zone at L3-L4 were compared during flexion-extension, lateral bending, and axial rotation. With intact disc, stability was equivalent after PS-rod and CS-rod fixation, except that PS-rod fixation was stiffer during axial rotation. With DLIF support, there was no significant difference in stability between PS-rod and CS-rod fixation. With TLIF support, PS-rod fixation was stiffer than CS-rod fixation during lateral bending. Bilateral CS-rod fixation provided about the same stability in cadaveric specimens

Freely suspended rod fall dampener, especially for control rod of liquid-cooled nuclear reactor

International Nuclear Information System (INIS)

Becvar, J.; Saroch, V.

1977-01-01

A shock absorber is described whose advantage is that the space required for the movement of the shock absorber in the operating travel of the system suspension rod-control rod bundle may be reduced. The design allows the automatic disconnection of the system and the removal of the suspension rod from the reactor without dismantling. The braking force reaction is transmitted to the structure above the core. The system fall energy is absorbed on the side of the suspension rod which has a bigger mass. (J.B.)

RodPilot{sup R} - The Innovative and Cost-Effective Digital Control Rod Drive Control System for PWRs

Energy Technology Data Exchange (ETDEWEB)

Baron, Clemens [AREVA NP GmbH, NLEE-G, Postfach 1199, 91001 Erlangen (Germany)

2008-07-01

With RodPilot, AREVA NP offers an innovative and cost-effective system for controlling control rods in Pressurized Water Reactors. RodPilot controls the three operating coils of the control rod drive mechanism (lift, moveable gripper and stationary gripper coil). The rods are inserted into or withdrawn from the core as required by the Reactor Control System. The system combines modern components, state-of-the-art logic and a proven electronic control rod drive control principle to provide enhanced reliability and lower maintenance costs. (author)

Control-rod driving mechanism

International Nuclear Information System (INIS)

Jodoi, Takashi.

1976-01-01

Purpose: To prevent falling of control rods due to malfunction. Constitution: The device of the present invention has a scram function in particular, and uses principally a fluid pressure as a scram accelerating means. The control rod is held by upper and lower holding devices, which are connected by a connecting mechanism. This connecting mechanism is designed to be detachable only at the lower limit of driving stroke of the control rod so that there occurs no erroneous scram resulting from careless disconnection of the connecting mechanism. Further, scramming operation due to own weight of the scram operating portion such as control rod driving shaft may be effected to increase freedom. (Kamimura, M.)

REACTOR CONTROL ROD OPERATING SYSTEM

Science.gov (United States)

Miller, G.

1961-12-12

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

Design of diverse safety rod and its drive mechanism of PFBR

International Nuclear Information System (INIS)

Vijayashree, R.; Govindarajan, S.; Chetal, S.C.

1997-01-01

In Prototype Fast Breeder Reactor (PFBR), there are two types of absorber rods for control and shutdown of the reactor in the event of any abnormal event. They are: (i) Control and Safety Rod (CSR) and (ii) the Diverse Safety Rod (DSR). Of these, the former (CSR) caters to the control function of the reactor during normal operating conditions and to the shutdown during abnormal situations. The DSR, on the other hand is meant essentially for the reactor shutdown to take care of any abnormal transient. It is rather important to note that functionally the DSR is independent of CSR in the sense, that it can bring the reactor to a cold shutdown state and maintain it even under the hypothetical condition of the failure of CSR. From the design point of view, this stipulates a failure probability of less than 10 -4 per demand. The DSR is normally parked above the core by the Diverse Safety Rod Drive Mechanism (DSRDM). On receiving a scram signal it gets released from the holding electromagnet and falls under the gravity into the core. Diverse features are incorporated both in the absorber rods and in the drive mechanisms to avoid common mode failures. This paper discusses the salient features of DSR and DSRDM. A brief account of detailed design, analysis and development of two important subassemblies viz. electromagnet and sodium dash pot is also presented. In addition, a brief comparison between CSR and DSR including their drive mechanisms is also provided. (author)

Control rod velocity limiter

International Nuclear Information System (INIS)

Cearley, J.E.; Carruth, J.C.; Dixon, R.C.; Spencer, S.S.; Zuloaga, J.A. Jr.

1986-01-01

This patent describes a velocity control arrangement for a reciprocable, vertically oriented control rod for use in a nuclear reactor in a fluid medium, the control rod including a drive hub secured to and extending from one end therefrom. The control device comprises: a toroidally shaped control member spaced from and coaxially positioned around the hub and secured thereto by a plurality of spaced radial webs thereby providing an annular passage for fluid intermediate the hub and the toroidal member spaced therefrom in coaxial position. The side of the control member toward the control rod has a smooth generally conical surface. The side of the control member away from the control rod is formed with a concave surface constituting a single annular groove. The device also comprises inner and outer annular vanes radially spaced from one another and spaced from the side of the control member away from the control rod and positioned coaxially around and spaced from the hub and secured thereto by spaced radial webs thereby providing an annular passage for fluid intermediate the hub and the vanes. The vanes are angled toward the control member, the outer edge of the inner vane being closer to the control member and the inner edge of the outer vane being closer to the control member. When the control rod moves in the fluid in the direction toward the drive hub the vanes direct a flow of fluid turbulence which provides greater resistance to movement of the control rod in the direction toward the drive hub than in the other direction

Control rod position control device

International Nuclear Information System (INIS)

Ubukata, Shinji.

1997-01-01

The present invention provides a control rod position control device which stores data such as of position signals and driving control rod instruction before and after occurrence of abnormality in control for the control rod position for controlling reactor power and utilized the data effectively for investigating the cause of abnormality. Namely, a plurality of individual control devices have an operation mismatching detection circuit for outputting signals when difference is caused between a driving instruction given to the control rod position control device and the control rod driving means and signals from a detection means for detecting an actual moving amount. A general control device collectively controls the individual control devices. In addition, there is also disposed a position storing circuit for storing position signals at least before and after the occurrence of the control rod operation mismatching. With such procedures, the cause of the abnormality can be determined based on the position signals before and after the occurrence of control rod mismatching operation stored in the position storing circuit. Accordingly, the abnormality cause can be determined to conduct restoration in an early stage. (I.S.)

Control rod selecting and driving device

International Nuclear Information System (INIS)

Isobe, Hideo.

1981-01-01

Purpose: To simultaneously drive a predetermined number of control rods in a predetermined mode by the control of addresses for predetermined number of control rods and read or write of driving codified data to and from the memory by way of a memory controller. Constitution: The system comprises a control rod information selection device for selecting predetermined control rods from a plurality of control rods disposed in a reactor and outputting information for driving them in a predetermined mode, a control rod information output device for codifying the information outputted from the above device and outputting the addresses to the predetermined control rods and driving mode coded data, and a driving device for driving said predetermined control rods in a predetermined mode in accordance with the codified data outputted from the above device, said control rod infromation output device comprising a memory device capable of storing a predetermined number of the codified data and a memory control device for storing the predetermined number of data into the above memory device at a predetermined timing while successively outputting the thus stored predetermined number of data at a predetermined timing. (Seki, T.)

NEUTRONIC REACTOR CONTROL ROD DRIVE APPARATUS

Science.gov (United States)

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.

RODDRP - A FORTRAN program for use in control rod calibration by the rod drop method

International Nuclear Information System (INIS)

Wilson, W.E.

1972-01-01

The different methods to measure reactivity which are applicable to control rod calibration are discussed. They include: 1) the positive period method, 2) the rod drop method, 3) the source-jerk method, 4) the rod oscillation method, and 5) the pulsed neutron method. The instrument setup used at WSU for rod drop measurements is presented. To speed up the analysis of power fall-off trace, a FORTRAN IV program called RODDRP was written to simultaneously solve the in-hour equation and relative neutron flux. The procedure for calculating the worth of the rod that produced the power trace is given. The reactivity for each time relative flux point is obtained. Conclusions about the status of the equipment are made

Simulation of vibration modes of the fuel rod damaged due to the grid-to-rod fretting wear

International Nuclear Information System (INIS)

Kim, Kyu Tae; Kim, Kyeong Koo; Jang, Young Ki; Lee, Kyou Seok

1997-01-01

The flow-induced fuel fretting wear observed in some PWRs mainly proceeds in the grid-to-rod contact positions. The grid-to-rod fretting wear in the PWR fuel assembly depends on grid-to-rod gap size, its axial profile and flow-induced vibration. This paper describes the GRIDFORCE program which generates the axially dependent grid-to-rod gap size as a function of burnup. The axially dependent grid-to-rod gap profiles are employed to predict the fuel rod vibration mode shapes by the ANSYS code. With the help of the Paidousis empirical formula, this paper also calculates the fuel rod vibration amplitudes under various supporting conditions, which indicates that the increase of the number of unsupported mid-grids will increase the fuel rod vibration amplitude. On the other hand, the comparison of the predicted vibration mode shapes and the observed mid-grid fretting wear pattern indicates that the 1st and 6th vibration mode shapes under the supporting inactive condition at the mid-grids can simulate the observed mid-grid fretting wear profile. This paper also proposes design guidelines against the grid-to-rod fretting wear. (author). 3 refs., 8 figs

Automated nuclear fuel rod pattern loading system

International Nuclear Information System (INIS)

Lambert, D.V.; Nylund, T.W.; Byers, J.W.; Haley, D.E. Jr.; Cioffi, J.V.

1991-01-01

This patent describes a method for loading fuel rods in a desired pattern. It comprises providing a supply of fuel rods of known enrichments; providing a magazine defining a matrix of elongated slots open at their forward ends for receiving fuel rods; defining a fuel rod feed path; receiving successively one at a time along the feed path fuel rods selected from the supply thereof; verifying successively one at a time along the feed path the identity of the selected fuel rods, the verifying including blocking passage of each selected fuel rod along the feed path until the identity of each selected fuel rod is confirmed as correct; feeding to the magazine successively one at a time along the feed path the selective and verified fuel rods; and supporting and moving the magazine along X-Y axes to successively align one at a time selected ones of the slots with the feed path for loading in the magazine the successive fuel rods in a desired enrichment pattern

Control rod drives

International Nuclear Information System (INIS)

Ikakura, Hiroaki.

1986-01-01

Purpose: To enable to direct disconnection of control rods upon abnormal temperature rise in the reactor thereby improve the reliability for the disconnecting operation in control rod drives for FBR type reactors upon emergency. Constitution: A diaphragm is disposed to the upper opening of a sealing vessel inserted to the hollow portion of an electromagnet and a rod is secured to the central position of the upper surface. A spring contacts are attached by way of an insulator to the inner surface at the lower portion of an extension pipe and connected with cables for supplying electric power sources respectively to a magnet. If the temperature in the reactor abnormally rises, liquid metals in the sealing vessel are expanded tending to extend the bellows downwardly. However, since they are attracted by the electromagnet, the thermal expansion of the liquid metals exert on the diaphragm prior to the bellows. Thus, the switch between the spring contacts is made open to attain the deenergized state to thereby disconnect the control rod and shutdown the neclear reactor. (Horiuchi, T.)

Method for compacting spent nuclear reactor fuel rods

International Nuclear Information System (INIS)

Wachter, W.J.

1988-01-01

In a nuclear reactor system which requires periodic physical manipulation of spent fuel rods, the method of compacting fuel rods from a fuel rod assembly is described. The method consists of: (1) removing the top end from the fuel rod assembly; (2) passing each of multiple fuel rod pulling elements in sequence through a fuel rod container and thence through respective consolidating passages in a fuel rod directing chamber; (3) engaging one of the pulling elements to the top end of each of the fuel rods; (4) drawing each of the pulling elements axially to draw the respective engaged fuel rods in one axial direction through the respective the passages in the chamber to thereby consolidate the fuel rods into a compacted configuration of a cross-sectional area smaller than the cross-sectional area occupied thereby within the fuel rod assembly; and (5) drawing all of the engaged fuel rods concurrently and substantially parallel to one another in the one axial direction into the fuel rod container while maintaining the compacted configuration whereby the fuel rods are aligned within the container in a fuel rod density of the the fuel rod assembly

Subchannel analysis and correlation of the Rod Bundle Heat Transfer (RBHT) steam cooling experimental data

Energy Technology Data Exchange (ETDEWEB)

Riley, M.P.; Mohanta, L.; Miller, D.J.; Cheung, F.B. [Pennsylvania State Univ., University Park, PA (United States); Bajorek, S.M.; Tien, K.; Hoxie, C.L. [U.S. Nuclear Regulatory Commission, Washington, DC (United States). Office of Nuclear Regulatory Research

2016-07-15

A subchannel analysis of the steam cooling data obtained in the Rod Bundle Heat Transfer (RBHT) test facility has been performed in this study to capture the effect of spacer grids on heat transfer. The RBHT test facility has a 7 x 7 rod bundle with heater rods and with seven spacer grids equally spaced along the length of the rods. A method based on the concept of momentum and heat transport analogy has been developed for calculating the subchannel bulk mean temperature from the measured steam temperatures. Over the range of inlet Reynolds number, the local Nusselt number was found to exhibit a minimum value between the upstream and downstream spacer grids. The presence of a spacer grid not only affects the local Nusselt number downstream of the grid but also affects the local Nusselt number upstream of the next grid. A new correlation capturing the effect of Reynolds number on the local flow restructuring downstream as well as upstream of the spacer grids was proposed for the minimum Nusselt number. In addition, a new enhancement factor accounting for the effects of the upstream as well as downstream spacer grids was developed from the RBHT data. The new enhancement factor was found to compare well with the data from the ACHILLLES test facility.

Control rod drives

International Nuclear Information System (INIS)

Asano, Hiromitsu.

1979-01-01

Purpose: To drive control rods at an optimum safety speed corresponding to the reactor core output. Constitution: The reactor power is detected by a neutron detector and the output signal is applied to a process computer. The process computer issues a signal representing the reactor core output, which is converted through a function generator into a signal representing the safety speed of control rods. The converted signal is further supplied to a V/F converter and converted into a pulse signal. The pulse signal is inputted to a step motor driving circuit, which actuates a step motor to operate the control rods always at a safety speed corresponding to the reactor core power. (Furukawa, Y.)

Individual nuclear fuel rod weighing system

International Nuclear Information System (INIS)

Fogg, J. L.; Howell, C. A.; Smith, J. H.; Vining, G. E.

1985-01-01

An individual nuclear fuel rod weighing system for rods carried on a tray which moves along a materials handling conveyor. At a first tray position on the conveyor, a lifting device raises the rods off the tray and places them on an overhead ramp. A loading mechanism conveys the rods singly from the overhead ramp onto an overhead scale for individual weighing. When the tray is at a second position on the conveyor, a transfer apparatus transports each weighed rod from the scale back onto the tray

Measuring device for control rod driving time

International Nuclear Information System (INIS)

Tanaka, Kazuhiko; Hanabusa, Masatoshi.

1993-01-01

The present invention concerns a measuring device for control driving time having a function capable of measuring a selected control rod driving time and measuring an entire control rod driving time simultaneously. A calculation means and a store means for the selected rod control rod driving time, and a calculation means and a store means for the entire control rod driving time are disposed individually. Each of them measures the driving time and stores the data independent of each other based on a selected control rod insert ion signal and an entire control rod insertion signal. Even if insertion of selected and entire control rods overlaps, each of the control rod driving times can be measured reliably to provide an advantageous effect capable of more accurately conducting safety evaluation for the nuclear reactor based on the result of the measurement. (N.H.)

Status of rod consolidation, 1988

International Nuclear Information System (INIS)

Bailey, W.J.

1989-01-01

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

Inspecting method for fuel rods

International Nuclear Information System (INIS)

Watanabe, Masaaki; Kogure, Sumio.

1976-01-01

Purpose: To precisely detect the response of flaw in clad tube and submerged fuel pellets from a relationship between the surface of fuel rod and internal signal. Constitution: Ultrasonic reflected waves from the surface of fuel rods and the interior are detected and either one of fuel rod or ultrasonic flaw detecting contact is rotated to thereby precisely detect the response of the flaw of clad tube and submerged fuel pellets from a relationship between said surface and the interior. It will be noted that the ultrasonic flaw detecting contact used is of the line-focus type, the incident angle of ultrasonic wave from the ultrasonic flaw detecting contact relative to the fuel rod is the angle of skew, that is, the ultrasonic flaw detecting contact is not perpendicular to a center axis of the fuel rod but is slightly displace. That is, the use of the aforesaid contact may facilitate discrimination between the surface flaw of the fuel rod and the response of submergence, and in addition, the employment of the aforesaid incident angle makes it hard to receive reflected waves from the surface of the fuel rod which is great in terms of energy to facilitate discrimination of waves responsive to submergence. (Kawakami, Y.)

Control rod position detection device

International Nuclear Information System (INIS)

Akita, Haruo; Ogiwara, Sakae.

1996-01-01

The device of the present invention is used in a back-up shut down system of an LMFBR type reactor which is easy for maintenance, has high reliability and can recognize the position of control rods accurately. Namely, a permanent magnet is disposed to a control rod extension tube connected to the lower portion of the control rod. The detector guide tube is disposed in the vicinity of the control rod extension tube. A detector having a detection coil is inserted into a detector tube. With such constitution, the control rod can be detected at one position using the following method. (1) the movement of the magnetic field of the permanent magnet is detected by the detection coil. (2) a plurality of grooves are formed on the control rod extension tube, and the movement of the grooves is detected. In addition, the detection coil is inserted into the detector guide tube, and the signals from the detection coil are inputted to a signal processing circuit disposed at the outside of the reactor vessel using an MI cable to enable the maintenance of the detector. Further, if the detector comprises a detection coil and an excitation coil, the position of a dropped control rod can be recognized at a plurality of points. (I.S.)

Rope wind-up type control rod

International Nuclear Information System (INIS)

Tsuji, Teruaki; Watanabe, Shigeru.

1979-01-01

Purpose: To hold a control rod at a certain position even if the sealed cover of the rod drive mechanism should fail. Constitution: A plurality of friction plates, engaging wheels and a threaded shaft are provided to the wind-up drum for winding up a rope which moves the control rod up and down. While the control rod is adapted to drop by its own weight upon insertion, it is adapted to stop at a predetermined position exactly with no shocks by gradually increasing braking force by the sliding friction caused from the friction plates or the like. A ratch mechanism is provided to the upper portion of the control rod so that the top of the ratch piece may automatically engage the guide passage wall of the control rod upon uncontrolled running of the control rod to prevent further uncontrolled running thereof. (Ikeda, J.)

ELECTROMAGNETIC APPARATUS FOR MOVING A ROD

Science.gov (United States)

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.

Individual nuclear fuel rod weighing system

International Nuclear Information System (INIS)

Fogg, J.L.; Smith, J.H.; Vining, G.E.; Howell, C.A.

1985-01-01

An individual nuclear fuel rod weighing system for rods carried on a tray which moves along a materials handling conveyor is discussed. At a first tray position on the conveyor, a lifting device raises the rods off the tray and places them on an overhead ramp. A loading mechanism conveys the rods singly from the overhead ramp onto an overhead scale for individual weighing. When the tray is at a second position on the conveyor, a transfer apparatus transports each weighed rod from the scale back onto the tray

Control rods in LMFBRs: a physics assessment

International Nuclear Information System (INIS)

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

1982-08-01

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

THE USE OF NITINOL RODS IN SURGICAL TREATMENT OF DEGENERATIVE SCOLIOSIS. 2.5-YEAR FOLLOW-UP

Directory of Open Access Journals (Sweden)

Natalia Sergeyevna Morozova

2016-03-01

Full Text Available ABSTRACT Objectives: To compare the outcomes of surgical treatment with lumbar fixation using nitinol rods without fusion and with standard lumbar fixation with titanium rods and interbody fusion. Methods: Treatment results of 70 patients with degenerative lumbar scoliosis aged 40 to 82 were analyzed. In all cases pedicle screws and nitinol rods with a diameter of 5.5 mm were used. Thirty patients underwent fixation at L1-S1 and 40 patients underwent fixation at L1-L5. Spinal fusion was not performed. All patients had radiography, CT and MRI performed. The results were assessed according to the Oswestry scale, SRS 22, SF 36 and VAS. The minimum follow-up period for all patients was 2.5 years. For the control group, consisting of 72 patients, pedicle fixation with titanium rods and interbody fusion in the lumbosacral region were performed. Results: The average level of deformity correction equaled 25° (10° - 38°. The analysis of X-ray and CT-scans revealed a single patient with implant instability, two patients with bone resorption around the screws and one patient with rod fractures. Functional radiography 2.5 years after surgery showed an average mobility of the lumbar spine of 21° (15° - 30°. There were no problems at the adjacent levels. Conclusions: The use of nitinol rods in spinal deformity surgery is promising. This technology is an alternative to rigid fixation. Continued gathering of clinical data and its further evaluation is necessary.

Self-Assembly of Rod-Coil Block Copolymers

National Research Council Canada - National Science Library

Jenekhe, S

1999-01-01

... the self-assembly of new rod-coil diblock, rod- coil-rod triblock, and coil-rod-coil triblock copolymers from solution and the resulting discrete and periodic mesostmctares with sizes in the 100...

Frontalis Sling Operation using Silicone Rod Compared with Autogenous Fascia Lata for Simple Congenital Ptosis

Directory of Open Access Journals (Sweden)

Purnima Rajkarnikar Sthapit

2014-09-01

Full Text Available Introduction: To evaluate the cosmetic results and recurrence of unilateral frontalis sling surgery using a silicone rod compared with autogenous fascia lata in cases of simple congenital ptosis. Methods: This is a retrospective comparative study of 59 patients who underwent a frontalis sling operation for congenital ptosis. Patients were divided into two groups according to the sling material used; an autogeneous fascia lata (fl group (n = 24 and a silicone rod group (sl (n = 35. Cosmetic results and recurrence rates were compared between these 2 groups. The cosmetic results of the frontalis sling operation were assessed as good, fair, or poor based on the difference between the Margin Reflex Distance of both eyelids and graded as good if the difference in two eyes was ≤1mm and poor if it was 2mm or more. Recurrence was defined as the conversion of the cosmetic result from good or fair to poor category. Results: At postoperative day seven and 30, MRD of both the groups were good but on three months follow-up MRD of silicon rod group dropped, however it was not statistically significant .Lid contour was good in both the groups, however, lid symmetry was poor in two cases of fascia lata at three months follow-up. Repeat surgery for poor outcome was done in 8.6% of cases in silicon rod and 8.3% of fascia lata group. Conclusions: The frontalis sling operation using either a silicone rod or autogenous fascia lata showed equally good cosmetic results and lower recurrence rate at three months follow up. Keywords: congenital ptosis; fascia lata; frontalis sling surgery; margin reflex distance; silicone rod.

Ejected control rod and rods drop measurements during Mochovce startup physical tests

International Nuclear Information System (INIS)

Minarcin, Miroslav; Elko, Marek

1998-01-01

Paper deals with measurements of asymmetric reactivity insertion into the reactor core that were carried out during physical startup tests of Mochovce Unit 1 in June 1998. Control rods worth measurements with one and two rods s tucked in upper limit and worth measurement of one control rod from group 6 'ejected' from the reactor core are discussed. During the experiments neutron flux was measured by four ionisation chambers (three of them were placed symmetrically around the reactor core). Results of measurements and influence of asymmetric reactivity influence on ionisation chambers response are presented in the paper. (Authors)

Quivers For Special Fuel Rods-Disposal Of Special Fuel Rods In CASTOR V Casks

International Nuclear Information System (INIS)

Bannani, Amin; Cebula, Wojciech; Buchmuller, Olga; Huggenberg, Roland; Helmut Kuhl

2015-01-01

While GNS casks of the CASTOR family are a suitable means to transfer fuel assemblies (FA) from the NPP to an interim dry storage site, Germanys phase-out of nuclear energy has triggered the demand for an additional solution to dispose of special fuel rods (SFR), normally remaining in the fuel pond until the final shutdown of the NPP. SFR are fuel rods that had to be removed from fuel assemblies mainly due to their special condition, e. g. damages in the cladding of the fuel rods which may have occurred during reactor operations. SFR are usually stored in the spent fuel pond after they are removed from the FA. The quiver for special fuel rods features a robust yet simple design, with a high mechanical stability, a reliable leak-tightness and large safety margins for future requirements on safety analysis. The quiver for special fuel rods can be easily adapted to a large variety of different damaged fuel rods and tailored to the specific need of the customer. The quiver for special fuel rods is adaptable e.g. in length and diameter for use in other types of transport and storage casks and is applicable in other countries as well. The overall concept presented here is a first of its kind solution for the disposal of SFRs via Castor V-casks. This provides an important precondition in achieving the status 'free from nuclear fuel' of the shut down German NPPs

Quivers For Special Fuel Rods-Disposal Of Special Fuel Rods In CASTOR V Casks

Energy Technology Data Exchange (ETDEWEB)

Bannani, Amin; Cebula, Wojciech; Buchmuller, Olga; Huggenberg, Roland [GNS, Essen (Germany); Helmut Kuhl [WTI, Julich (Germany)

2015-05-15

While GNS casks of the CASTOR family are a suitable means to transfer fuel assemblies (FA) from the NPP to an interim dry storage site, Germanys phase-out of nuclear energy has triggered the demand for an additional solution to dispose of special fuel rods (SFR), normally remaining in the fuel pond until the final shutdown of the NPP. SFR are fuel rods that had to be removed from fuel assemblies mainly due to their special condition, e. g. damages in the cladding of the fuel rods which may have occurred during reactor operations. SFR are usually stored in the spent fuel pond after they are removed from the FA. The quiver for special fuel rods features a robust yet simple design, with a high mechanical stability, a reliable leak-tightness and large safety margins for future requirements on safety analysis. The quiver for special fuel rods can be easily adapted to a large variety of different damaged fuel rods and tailored to the specific need of the customer. The quiver for special fuel rods is adaptable e.g. in length and diameter for use in other types of transport and storage casks and is applicable in other countries as well. The overall concept presented here is a first of its kind solution for the disposal of SFRs via Castor V-casks. This provides an important precondition in achieving the status 'free from nuclear fuel' of the shut down German NPPs.

Monitoring device for withdrawing control rods

International Nuclear Information System (INIS)

Higashigawa, Yuichi.

1985-01-01

Purpose: To improve the sensitivity and the responsivity to an equivalent extent to those in the case where local power range monitors are densely arranged near each of the control rods, with no actual but pseudo increase of the number of local power range monitors. Constitution: The monitor arrangement is patterned by utilizing the symmetricity of the reactor core and stored in a monitor designating device. The symmetricity of control rods to be selected and withdrawn by an operator is judged by a control rod symmetry monitoring device, while the symmetricity of the withdrawn control rods is judged by a control rod withdrawal state monitoring device. Then, only when both of the devices judge the symmetricity, the control rods are subjected to gang driving by the control rod drive mechanisms. In this way, monitoring at a high sensitivity and responsivity is enabled with no increase for the number of monitors. (Yoshino, Y.)

Nondestructive assay of HTGR fuel rods

International Nuclear Information System (INIS)

Menlove, H.O.

1974-01-01

Performance characteristics of three different radioactive source NDA systems are compared for the assay of HTGR fuel rods and stacks of rods. These systems include the fast neutron Sb-Be assay system, the 252 Cf ''Shuffler,'' and the thermal neutron PAPAS assay system. Studies have been made to determinethe perturbation on the measurements from particle size, kernel Th/U ratio, thorium content, and hydrogen content. In addition to the total 235 U determination, the pellet-to-pellet or rod-to-rod uniformity of HTGR fuel rod stacks has been measured by counting the delayed gamma rays with a NaI through-hole in the PAPAS system. These measurements showed that rod substitutions can be detected easily in a fuel stack, and that detailed information is available on the loading variations in a uniform stack. Using a 1.0 mg 252 Cf source, assay rates of 2 to 4 rods/s are possible, thus facilitating measurement of 100 percent of a plant's throughput. (U.S.)

Inspection system for Zircaloy clad fuel rods

International Nuclear Information System (INIS)

Yancey, M.E.; Porter, E.H.; Hansen, H.R.

1975-10-01

A description is presented of the design, development, and performance of a remote scanning system for nondestructive examination of fuel rods. Characteristics that are examined include microcracking of fuel rod cladding, fuel-cladding interaction, cladding thickness, fuel rod diameter variation, and fuel rod bowing. Microcracking of both the inner and outer fuel rod surfaces and variations in wall thickness are detected by using a pulsed eddy current technique developed by Argonne National Laboratory (ANL). Fuel rod diameter variation and fuel rod bowing are detected by using two linear variable differential transformers (LVDTs) and a signal conditioning system. The system's mechanical features include variable scanning speeds, a precision indexing system, and a servomechanism to maintain proper probe alignment. Initial results indicate that the system is a very useful mechanism for characterizing irradiated fuel rods

Seismic scrammability of HTTR control rods

International Nuclear Information System (INIS)

Nishiguchi, I.; Iyoku, T.; Ito, N.; Watanabe, Y.; Araki, T.; Katagiri, S.

1990-01-01

Scrammability tests on HTTR (High-Temperature Engineering Test Reactor) control rods under seismic conditions have been carried out and seismic conditions influences on scram time as well as functional integrity were examined. A control rod drive located in a stand-pipe at the top of a reactor vessel, raises and lowers a pair of control rods by suspension cables. Each flexible control rod consists of 10 neutron absorber sections held together by a metal spine passing through the center. It falls into a hole in graphite blocks due to gravity at scram. In the tests, a full scale control rod drive and a pair of control rods were employed with a column of graphite blocks in which holes for rods were formed. Blocks misalignment and contact with the hole surface during earthquakes were considered as major causes of disturbance in scram time. Therefore, the following parameters were set up in the tests: excitation direction, combination or horizontal and vertical excitation, acceleration, frequency and block to block gaps. Main results obtained from tests are as follow. 1) Every scram time obtained under the design conditions was within 6 seconds. On the contrary, the scram times were 5.2 seconds when there were no vibration. Therefore, it was concluded that the seismic effects on scram time were not significant. 2) Scram time became longer with increase in both acceleration and horizontal excitation frequency, and control rods fell very smoothly without any jerkiness. This suggests that collision between control rods and hole surface is the main disturbing factor of falling motion. 3) Mechanical and functional integrity of control rod drive mechanism, control rods and graphite blocks was confirmed after 140 seismic scrammability tests. (author). 10 figs, 1 tab

Comparisons of numerical simulations with ASTRID code against experimental results in rod bundle geometry for boiling flows

International Nuclear Information System (INIS)

Larrauri, D.; Briere, E.

1997-12-01

After different validation simulations of flows through cylindrical and annular channels, a subcooled boiling flow through a rod bundle has been simulated with ASTRID Steam-Water of software. The experiment simulated is called Poseidon. It is a vertical rectangular channel with three heating rods inside. The thermohydraulic conditions of the simulated flow were close to the DNB conditions. The simulation results were analysed and compared against the available measurements of liquid and wall temperatures. ASTRID Steam-Water produced satisfactory results. The wall and the liquid temperatures were well predicted in the different parts of the flow. The void fraction reached 40 % in the vicinity of the heating rods. The distribution of the different calculated variables showed that a three-dimensional simulation gives essential information for the analysis of the physical phenomena involved in this kind of flow. The good results obtained in Poseidon geometry will encourage future rod bundle flow simulations and analyses with ASTRID Steam-Water code. (author)

Control rod testing apparatus

International Nuclear Information System (INIS)

Gaunt, R.R.; Ashman, C.M.

1987-01-01

A control rod testing apparatus is described comprising: a first guide means having a vertical cylindrical opening for grossly guiding a control rod; a second guide means having a vertical cylindrical opening for grossly guiding a control rod. The first and second guide means are supported at axially spaced locations with the openings coaxial; and a substantially cylindrical subassembly having a vertical cylindrical opening therethrough. The subassembly is trapped coaxial with and between the first and second guide means, and the subassembly radially floats with respect to the first and second guide means

Development of the automatic control rod operation system for JOYO. Verification of automatic control rod operation guide system

International Nuclear Information System (INIS)

Terakado, Tsuguo; Suzuki, Shinya; Kawai, Masashi; Aoki, Hiroshi; Ohkubo, Toshiyuki

1999-10-01

The automatic control rod operation system was developed to control the JOYO reactor power automatically in all operation modes(critical approach, cooling system heat up, power ascent, power descent), development began in 1989. Prior to applying the system, verification tests of the automatic control rod operation guide system was conducted during 32nd duty cycles of JOYO' from Dec. 1997 to Feb. 1998. The automatic control rod operation guide system consists of the control rod operation guide function and the plant operation guide function. The control rod operation guide function provides information on control rod movement and position, while the plant operation guide function provide guidance for plant operations corresponding to reactor power changes(power ascent or power descent). Control rod insertion or withdrawing are predicted by fuzzy algorithms. (J.P.N.)

Control rod drive shaft latch

International Nuclear Information System (INIS)

Thorp, A.G. II.

1976-01-01

A latch mechanism is operated by differential pressure on a piston to engage the drive shaft for a control rod in a nuclear reactor, thereby preventing the control rod from being ejected from the reactor in case of failure of the control rod drive mechanism housing which is subjected to the internal pressure in the reactor vessel. 6 claims, 4 drawing figures

Reconstitutable control rod spider assembly

International Nuclear Information System (INIS)

Shallenberger, J.M.; Ferian, S.J.

1990-01-01

A reconstitutable control rod/spider assembly includes a hollow connecting finger of the spider having a pair of opposing flat segments formed on the interior thereof and engaging a pair of opposing flat sectors formed on the exterior of a stem extending form the upper end of control rod. The stem also has an externally-threaded portion engaging a nut and a pilot aligning portion for the nut. The nut has a radially flexible and expandable thread-defining element captured in its bore. The segments and sectors allow the rod to be removed and reattached after turning through 180 0 to allow more even wear on the rod. (author)

Estimation of irradiated control rod worth

International Nuclear Information System (INIS)

Varvayanni, M.; Catsaros, N.; Antonopoulos-Domis, M.

2009-01-01

When depleted control rods are planned to be used in new core configurations, their worth has to be accurately predicted in order to deduce key design and safety parameters such as the available shutdown margin. In this work a methodology is suggested for the derivation of the distributed absorbing capacity of a depleted rod, useful in the case that the level of detail that is known about the irradiation history of the control rod does not allow an accurate calculation of the absorber's burnup. The suggested methodology is based on measurements of the rod's worth carried out in the former core configuration and on corresponding calculations based on the original (before first irradiation) absorber concentration. The methodology is formulated for the general case of the multi-group theory; it is successfully tested for the one-group approximation, for a depleted control rod of the Greek Research Reactor, containing five neutron absorbers. The computations reproduce satisfactorily the irradiated rod worth measurements, practically eliminating the discrepancy of the total rod worth, compared to the computations based on the nominal absorber densities.

Rod cluster having improved vane configuration

International Nuclear Information System (INIS)

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

1989-01-01

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

Sucker rod motor

Energy Technology Data Exchange (ETDEWEB)

Radzalov, N N; Radzhabov, N A

1983-01-01

The motor consists of rollers mounted on the wellmouth and connected by a flexible rink. Reciprocating mechanism is in the form of a horizontal non-mobile single-side operation cylinder, inside which a plunger and rod are mounted. The working housing of the hydrocylinder is connected to a gas-hydr aulic batter, and when running is connected via plunger to the high pressure source; running in reverse it is connected with a safety valve and automatic control unit. The unit is equipped with a reducer and a mechanical transformer consisting of screw and nut, and which is shutoff with a single-side lining. The plunger rod consists of an auger-like unit. The high pressure source is provided by the injection line of the sucker rod that has been equipped with a reverse valve.

Duke Power Company's control rod wear program

International Nuclear Information System (INIS)

Culp, D.C.; Kitlan, M.S. Jr.

1990-01-01

Recent examinations performed at several foreign and domestic pressurized water reactors have identified significant control rod cladding wear, leading to the conclusion that previously believed control rod lifetimes are not attainable. To monitor control rod performance and reduce safety concerns associated with wear, Duke Power Company has developed a comprehensive control rod wear program for Ag-In-Cd and boron carbide (B 4 C) rods at the McGuire and Catawba nuclear stations. Duke Power currently uses the Westinghouse 17 x 17 Ag-In-Cd control rod design at McGuire Unit 1 and the Westinghouse 17 x 17 hybrid B 4 C control rod design with a Ag-In-Cd tip at McGuire Unit 2 and Catawba Units 1 and 2. The designs are similar, with the exception of the absorber material and clad thickness. There are 53 control rods per unit

Fuel rod pellet loading head

International Nuclear Information System (INIS)

Howell, T.E.

1975-01-01

An assembly for loading nuclear fuel pellets into a fuel rod comprising a loading head for feeding pellets into the open end of the rod is described. The pellets rest in a perforated substantially V-shaped seat through which air may be drawn for removal of chips and dust. The rod is held in place in an adjustable notched locator which permits alignment with the pellets

MreB drives de novo rod morphogenesis in Caulobacter crescentus via remodeling of the cell wall.

Science.gov (United States)

Takacs, Constantin N; Poggio, Sebastian; Charbon, Godefroid; Pucheault, Mathieu; Vollmer, Waldemar; Jacobs-Wagner, Christine

2010-03-01

MreB, the bacterial actin-like cytoskeleton, is required for the rod morphology of many bacterial species. Disruption of MreB function results in loss of rod morphology and cell rounding. Here, we show that the widely used MreB inhibitor A22 causes MreB-independent growth inhibition that varies with the drug concentration, culture medium conditions, and bacterial species tested. MP265, an A22 structural analog, is less toxic than A22 for growth yet equally efficient for disrupting the MreB cytoskeleton. The action of A22 and MP265 is enhanced by basic pH of the culture medium. Using this knowledge and the rapid reversibility of drug action, we examined the restoration of rod shape in lemon-shaped Caulobacter crescentus cells pretreated with MP265 or A22 under nontoxic conditions. We found that reversible restoration of MreB function after drug removal causes extensive morphological changes including a remarkable cell thinning accompanied with elongation, cell branching, and shedding of outer membrane vesicles. We also thoroughly characterized the composition of C. crescentus peptidoglycan by high-performance liquid chromatography and mass spectrometry and showed that MreB disruption and recovery of rod shape following restoration of MreB function are accompanied by considerable changes in composition. Our results provide insight into MreB function in peptidoglycan remodeling and rod shape morphogenesis and suggest that MreB promotes the transglycosylase activity of penicillin-binding proteins.

Rod displacement measurements by x-ray CT and its impact on thermal-hydraulics in tight-lattice rod bundle (Joint research)

International Nuclear Information System (INIS)

Mitsutake, Toru; Misawa, Takeharu; Kureta, Masatoshi; Akimoto, Hajime

2005-06-01

In tight-lattice simulated rod bundles with about 1 mm gap between rods, a rod displacement might affect thermal-hydraulic characteristics since the displacement has a strong impact on the flow area change along the heated section. It should be important to estimate how large the rod position displacement could quantitatively affect critical power for the tight-lattice rod bundle from the point of improvement of prediction capability of subchannel analysis. In the present study, the inside-structure observation of the simulated seven-rod bundle of Reduced Moderation Water Reactor (RMWR) was made through the whole length of the test assembly. Based on the measured rod position data, the relation between the rod position displacement and the heat transfer characteristics was investigated experimentally and through the two kinds of subchannel analysis, the nominal rod position case and the measured rod position case, the effect on the predicted critical power was estimated. The high-energy X-ray computer tomograph (CT) of Fuels Monitoring Facilities (FMF) at the O-arai Engineering Center in Japan Nuclear Cycle Institute (JNC) was applied for the inside-structure observation of the test assembly. The CT view of the cross sections within the test assembly assured the hexagonal rod position arrangement was almost the same as expected by design. The measured data with the X-ray CT facility showed that all rod displacements were small, 0.5 millimeters at maximum and 0.2 millimeters in average. In the heat transfer experiments for the seven-rod bundle, the boiling transition (BT) position and the rod surface temperature behavior was measured. All thermocouples on the center rod downstream from the BT-onset axial height showed almost simultaneous temperature increase due to BT. And the thermocouples located on the same axial heights showed quite similar time-variation behaviors in the vapor cooling heat transfer regime. These results demonstrated the effect of the

Dynamic rod worth measurements (''Rod Insertion''). Final report for the period 01 December 1994 - 30 November 1996

International Nuclear Information System (INIS)

Bogdan, G.

1996-12-01

Reload startup physics tests are performed for pressurized water reactors (PWR power plant) following a refuelling or other significant core alteration for which nuclear design calculations are required. Part of the reload startup physics tests are control rod group worths measurements. for this purpose a new so-called method ''Rod-Insertion'' was developed. It can also be used as an additional measuring instrument on the research reactor for education purposes. The principle of the rod-insertion method is to start from a critical reactor operating at low power and to measure the time-dependent reactivity change while a control rod is inserted into the core. Unlike in the rod-drop method, the measured control rod is inserted with the drive mechanism at normal speed. By analyzing the flux trace using point-kinetics, not only the total rod worth but also the differential and the integral rod worth curves are obtained. A high-quality electrometer is required for monitoring the neutron flux. The analysis is performed by transferring the data to an IBM PC compatible with some additional standard electronic board and the associated software. The new reactivity meter has been validated on the TRIGA Mark II reactors in Ljubljana and Vienna and at the Krsko Nuclear Power Plant during physics startup tests after reload. The results proved the high performance of the reactivity meter in the standard applications according to the existing procedures, as well as in the new rod-insertion technique of measuring the control rod group worths. This method drastically differs from others such as absence of any chemical control of reactivity (like boron exchange method), and minimizing a testing time and waste coolant production

Control rod ejection analysis during a depressurization accident and the development of a rod-ejection-preventing device

International Nuclear Information System (INIS)

Mitake, S.; Itoh, K.; Fukushima, H.; Inoue, T.

1982-01-01

The control rods used for the experimental VHTR are suspended in the core by means of flexible steel cables and it is conceivable that an accidental rod ejection could occur due to a depressurization accident. The computer code AFLADE was developed in order to analyze the possibility of accidental rod ejection, and several studies were performed. The parametric study results showed that the adopted design condition for the VHTR core will not cause a rod ejection accident. In parallel with these accident analyses, a rod-ejection-preventing device was developed in preparation for a hypothetical accident, and its function was verified by the component tests

Investigation of control rod worth and nuclear end of life of BWR control rods

International Nuclear Information System (INIS)

Magnusson, Per

2008-01-01

This work has investigated the Control Rod Worth (CRW) and Nuclear End of Life (NEOL) values for BWR control rods. A study of how different parameters affect NEOL was performed with the transport code PHOENIX4. It was found that NEOL, expressed in terms of 10 B depletion, can be generalized beyond the conditions for which the rod is depleted, such as different power densities and void fractions, the corresponding variation in the NEOL will be about 0.2-0.4% 10 B. It was also found that NEOL results for different fuel types and different fuel enrichments have a variation of about 2-3% in 10 B depletion. A comparative study on NHOL and CRW was made between PHOENIX4 and the stochastic Monte Carlo code MCNP. It was found that there is a significant difference, both due to differences in the codes and to limitations in the geometrical modeling in PHOENIX4. Since MCNP is considered more physically correct, a methodology was developed to calculate the nuclear end of life of BWR control rods with MCNP. The advantages of the methodology are that it does not require other codes to perform the depletion of the absorber material, it can describe control rods of any design and it can deplete the control rod absorber material without burning the fuel. The disadvantage of the method is that is it time-consuming

Preliminary nuclear design for test MOX Fuel rods

Energy Technology Data Exchange (ETDEWEB)

Joo, Hyung Kook; Kim, Taek Kyum; Jeong, Hyung Guk; Noh, Jae Man; Cho, Jin Young; Kim, Young Il; Kim, Young Jin; Sohn, Dong Seong

1997-10-01

As a part of activity for future fuel development project, test MOX fuel rods are going to be loaded and irradiated in Halden reactor core as a KAERI`s joint international program with Paul Scherrer Institute (PSI). PSI will fabricate test MOX rods with attrition mill device which was developed by KAERI. The test fuel assembly rig contains three MOX rods and three inert matrix rods. One of three MOX rods will be fabricated by BNFL, the other two MOX fuel rods will be manufacturing jointly by KAERI and PSI. Three inert matrix fuel rods will be fabricated with Zr-Y-Er-Pu oxide. Neutronic evaluation was preliminarily performed for test fuel assembly suggested by PSI. The power distribution of test fuel rod in test fuel assembly was analyzed for various fuel rods position in assembly and the depletion characteristic curve for test fuel was also determined. The fuel rods position in test fuel assembly does not effect the rod power distribution, and the proposal for test fuel rods suggested by PSI is proved to be feasible. (author). 2 refs., 13 tabs., 16 figs.

Single-phase convective heat transfer in rod bundles

International Nuclear Information System (INIS)

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

2008-01-01

The convective heat transfer for turbulent flow through rod bundles representative of nuclear fuel rods used in pressurized water reactors is examined. The rod bundles consist of a square array of parallel rods that are held on a constant pitch by support grids spaced axially along the rod bundle. Split-vane pair support grids, which create swirling flow in the rod bundle, as well as disc and standard support grids are investigated. Single-phase convective heat transfer coefficients are measured for flow downstream of support grids in a rod bundle. The rods are heated using direct resistance heating, and a bulk axial flow of air is used to cool the rods in the rod bundle. Air is used as the working fluid instead of water to reduce the power required to heat the rod bundle. Results indicate heat transfer enhancement for up to 10 hydraulic diameters downstream of the support grids. A general correlation is developed to predict the heat transfer development downstream of support grids. In addition, circumferential variations in heat transfer coefficients result in hot streaks that develop on the rods downstream of split-vane pair support grids

Single-phase convective heat transfer in rod bundles

Energy Technology Data Exchange (ETDEWEB)

Holloway, Mary V. [Mechanical Engineering Department, United States Naval Academy, 590 Holloway Rd., Annapolis, MD 21402 (United States)], E-mail: holloway@usna.edu; Beasley, Donald E. [Mechanical Engineering Department, Clemson University, Clemson, SC 29634 (United States); Conner, Michael E. [Westinghouse Nuclear Fuel, 5801 Bluff Road, Columbia, SC 29250 (United States)

2008-04-15

The convective heat transfer for turbulent flow through rod bundles representative of nuclear fuel rods used in pressurized water reactors is examined. The rod bundles consist of a square array of parallel rods that are held on a constant pitch by support grids spaced axially along the rod bundle. Split-vane pair support grids, which create swirling flow in the rod bundle, as well as disc and standard support grids are investigated. Single-phase convective heat transfer coefficients are measured for flow downstream of support grids in a rod bundle. The rods are heated using direct resistance heating, and a bulk axial flow of air is used to cool the rods in the rod bundle. Air is used as the working fluid instead of water to reduce the power required to heat the rod bundle. Results indicate heat transfer enhancement for up to 10 hydraulic diameters downstream of the support grids. A general correlation is developed to predict the heat transfer development downstream of support grids. In addition, circumferential variations in heat transfer coefficients result in hot streaks that develop on the rods downstream of split-vane pair support grids.

Fabrication of internally instrumented reactor fuel rods

International Nuclear Information System (INIS)

Schmutz, J.D.; Meservey, R.H.

1975-01-01

Procedures are outlined for fabricating internally instrumented reactor fuel rods while maintaining the original quality assurance level of the rods. Instrumented fuel rods described contain fuel centerline thermocouples, ultrasonic thermometers, and pressure tubes for internal rod gas pressure measurements. Descriptions of the thermocouples and ultrasonic thermometers are also contained

Vortex Noise from Rotating Cylindrical Rods

Science.gov (United States)

Stowell, E Z; Deming, A F

1935-01-01

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

Analysis of Crystallographic Textures in Aluminum Plates Processed by Equal Channel Angular Extrusion

DEFF Research Database (Denmark)

Li, Saiyi; Mishin, Oleg

2014-01-01

A modeling and experimental investigation has been conducted to explore the effect of processing route on texture evolution during equal channel angular extrusion (ECAE) of aluminum plate samples. It is found that although the textures in the plates develop along orientation fibers previously ide...... identified for ECAE-processed rods and bars, the main components and strength of these textures vary significantly with processing route, which may lead to considerable differences in the plastic anisotropy of the plates....

Some Equalities Are More Equal Than Others: Quality Equality Emerges Later Than Numerical Equality.

Science.gov (United States)

Sheskin, Mark; Nadal, Amber; Croom, Adam; Mayer, Tanya; Nissel, Jenny; Bloom, Paul

2016-09-01

By age 6, children typically share an equal number of resources between themselves and others. However, fairness involves not merely that each person receive an equal number of resources ("numerical equality") but also that each person receive equal quality resources ("quality equality"). In Study 1, children (N = 87, 3-10 years) typically split four resources "two each" by age 6, but typically monopolized the better two resources until age 10. In Study 2, a new group of 6- to 8-year-olds (N = 32) allocated resources to third parties according to quality equality, indicating that children in this age group understand that fairness requires both types of equality. © 2016 The Authors. Child Development © 2016 Society for Research in Child Development, Inc.

Vibrational characteristics and wear of fuel rods

International Nuclear Information System (INIS)

Schmugar, K.L.

1977-01-01

Fuel rod wear, due to vibration, is a continuing concern in the design of liquid-cooled reactors. In my report, the methodology and models that are used to predict fuel rod vibrational response and vibratory wear, in a light water reactor environment, are discussed. This methodology is being followed at present in the design of Westinghouse Nuclear Fuel. Fuel rod vibrations are expressed as the normal bending modes, and sources of rod vibration are examined with special emphasis on flow-induced mechanisms in the stable flow region. In a typical Westinghouse PWR fuel assembly design, each fuel rod is supported at multiple locations along the rod axis by a square-shaped 'grid cell'. For a fuel rod /grid support system, the development of small oscillatory motions, due to fluid flow at the rod/grid interface, results in material wear. A theoretical wear mode is developed using the Archard Theory of Adhesive Wear as the basis. Without question certainty, fretting wear becomes a serious problem if it progresses to the stage where the fuel cladding is penetrated and fuel is exposed to the coolant. Westinghouse fuel is designed to minimize fretting wear by limiting the relative motion between the fuel rod and its supports. The wear producing motion between the fuel rod and its supports occurs when the vibration amplitude exceeds the slippage threshold amplitude

Drive-in device for long thin rods into narrow cavitations, especially for control-shutdown rods e.g. of nuclear reactors

International Nuclear Information System (INIS)

Flessner, H.; Paeserack, U.

1974-01-01

The auxiliary device serves as holder for long and thin rods, e.g. control rods, transported hanging in bundles, when these are lowered into narrow cavities. It is constructed as a rod grab vertically movable at the end of a guide tube. A comb-shaped trap in connection with a guide rod serves for lateral support of the lower ends of the rods hanging on the grab. This guide rod can be moved in vertical direction by means of two pairs of convex rollers resting on the inner guide tube. In addition, the guide rod has a prolongation carrying a traverse by means of an abutment on the lower end. With these auxiliaries amongst others, the trap can be brought into a horizontal position by turning around an axis with the control rods meshing with the teeth of the trap while the parallelism of the rods is kept up during transport. (DG) [de

Control rod excess withdrawal prevention device

International Nuclear Information System (INIS)

Takayama, Yoshihito.

1992-01-01

Excess withdrawal of a control rod of a BWR type reactor is prevented. That is, the device comprises (1) a speed detector for detecting the driving speed of a control rod, (2) a judging circuit for outputting an abnormal signal if the driving speed is greater than a predetermined level and (3) a direction control valve compulsory closing circuit for controlling the driving direction of inserting and withdrawing a control rod based on an abnormal signal. With such a constitution, when the with drawing speed of a control rod is greater than a predetermined level, it is detected by the speed detector and the judging circuit. Then, all of the direction control valve are closed by way of the direction control valve compulsory closing circuit. As a result, the operation of the control rod is stopped compulsorily and the withdrawing speed of the control rod can be lowered to a speed corresponding to that upon gravitational withdrawal. Accordingly, excess withdrawal can be prevented. (I.S)

Hollow rods for the oil producing industry

Energy Technology Data Exchange (ETDEWEB)

Khalimova, L M; Elyasheva, M A

1970-01-01

Hollow sucker rods have several advantages over conventional ones. The hollow rods actuate the well pump and at the same time conduct produced fluids to surface. When paraffin deposition occurs, it can be minimized by injecting steam, hot oil or hot water into the hollow rod. Other chemicals, such as demulsifiers, scale inhibitors, corrosion inhibitors, etc., can also be placed in the well through the hollow rods. This reduces cost of preventive treatments, reduces number of workovers, increases oil production, and reduces cost of oil. Because the internal area of the rod is small, the passing liquids have a high velocity and thereby carry sand and dirt out of the well. This reduces pump wear between the piston and the plunger. Specifications of hollow rods, their operating characteristics, and results obtained with such rods under various circumstances are described.

ELECTRIC FIELD MEASUREMENT IN ROD-DISCONTINUED ...

African Journals Online (AJOL)

2014-06-30

Jun 30, 2014 ... the electrogeometrical model using a laboratory experimental rod-plane air gap arrangement with a lightning conductor (Franklin rod or horizontal conductor). The stepped leader could be represented by the rod electrode under a negative lightning impulse voltage having a level leading to breakdown with ...

Control rod

International Nuclear Information System (INIS)

Takeda, Toshikazu; Inoue, Kotaro.

1979-01-01

Purpose: To flatten the power distribution in the reactor core without impairing neutron economy by disposing pins containing elements of lower atomic number in the central region of a shroud and loading pins containing depleted uranium in the periphery region thereof. Constitution: The shroud has a layer of pins containing depleted uranium in the peripheral region and a layer of pins containing elements of lower atomic number such as beryllium in the central region. Heat removal from those pins containing depleted uranium and elements of lower atomic number (neutron moderator) is effected by sodium flow outside of the cladding material. The control rod operation is conducted by inserting or extracting the central portion (pins containing elements of lower atomic number such as beryllium) inside of the stainless pipe. Upon extraction of the control rod, the moderator in the central region is removed whereby high speed neutrons are no more deccelerated and the absorption rate to the depleted uranium is decreased. This can flatten the power distribution in the reactore core with the disposition of a plurality of control rods at a better neutron economy as compared with the use of neutron absorber such as boron. (Seki, T.)

Control rod

International Nuclear Information System (INIS)

Fukumoto, Takashi; Hirakawa, Hiromasa; Kawashima, Norio; Goto, Yasuyuki.

1994-01-01

Neutron absorbers are contained in a tubular member comprising, integrally a tubular portion and four corners disposed at the outer circumference of the tubular portion at every 90deg, to provide a neutron absorbing tube. A plurality of neutron absorbing tubes are arranged in parallel in the lateral direction, and adjacent corners are joined, into a blade to constitute a control rod. Such a control rod has a great structural strength, simple in the structure and relatively light in weight and can contain a great amount of neutron absorbers. Upon formation of the control rod by arranging the blades in a cross-like shape, at least a portion thereof is constituted with short neutron absorbing tubes shorter than the entire length of the blade, and gaps are formed at positions in adjacent in the axial direction. With such a constitution, there is no worry that a wing end of the blade collides against or be abraded with a fuel channel box or a fuel support. Even if fuel channels are vibrated upon scram of the reactor, such as occurrence of earthquakes, it can be inserted to the reactor easily. (N.H.)

Control rod housing alignment

International Nuclear Information System (INIS)

Dixon, R.C.; Deaver, G.A.; Punches, J.R.; Singleton, G.E.; Erbes, J.G.; Offer, H.P.

1990-01-01

This patent describes a process for measuring the vertical alignment between a hole in a core plate and the top of a corresponding control rod drive housing within a boiling water reactor. It comprises: providing an alignment apparatus. The alignment apparatus including a lower end for fitting to the top of the control rod drive housing; an upper end for fitting to the aperture in the core plate, and a leveling means attached to the alignment apparatus to read out the difference in angularity with respect to gravity, and alignment pin registering means for registering to the alignment pin on the core plate; lowering the alignment device on a depending support through a lattice position in the top guide through the hole in the core plate down into registered contact with the top of the control rod drive housing; registering the upper end to the sides of the hole in the core plate; registering the alignment pin registering means to an alignment pin on the core plate to impart to the alignment device the required angularity; and reading out the angle of the control rod drive housing with respect to the hole in the core plate through the leveling devices whereby the angularity of the top of the control rod drive housing with respect to the hole in the core plate can be determined

RODSWELL: a computer code for the thermomechanical analysis of fuel rods under LOCA conditions

International Nuclear Information System (INIS)

Casadei, F.; Laval, H.; Donea, J.; Jones, P.M.; Colombo, A.

1984-01-01

The present report is the user's manual for the computer code RODSWELL developed at the JRC-Ispra for the thermomechanical analysis of LWR fuel rods under simulated loss-of-coolant accident (LOCA) conditions. The code calculates the variation in space and time of all significant fuel rod variables, including fuel, gap and cladding temperature, fuel and cladding deformation, cladding oxidation and rod internal pressure. The essential characteristics of the code are briefly outlined here. The model is particularly designed to perform a full thermal and mechanical analysis in both the azimuthal and radial directions. Thus, azimuthal temperature gradients arising from pellet eccentricity, flux tilt, arbitrary distribution of heat sources in the fuel and the cladding and azimuthal variation of coolant conditions can be treated. The code combines a transient 2-dimensional heat conduction code and a 1-dimentional mechanical model for the cladding deformation. The fuel rod is divided into a number of axial sections and a detailed thermomechanical analysis is performed within each section in radial and azimuthal directions. In the following sections, instructions are given for the definition of the data files and the semi-variable dimensions. Then follows a complete description of the input data. Finally, the restart option is described

The Third ATLAS ROD Workshop

CERN Multimedia

Poggioli, L.

A new-style Workshop After two successful ATLAS ROD Workshops dedicated to the ROD hardware and held at the Geneva University in 1998 and in 2000, a new style Workshop took place at LAPP in Annecy on November 14-15, 2002. This time the Workshop was fully dedicated to the ROD-TDAQ integration and software in view of the near future integration activities of the final RODs for the detector assembly and commissioning. More precisely, the aim of this workshop was to get from the sub-detectors the parameters needed for T-DAQ, as well as status and plans from ROD builders. On the other hand, what was decided and assumed had to be stated (like EB decisions and URDs), and also support plans. The Workshop gathered about 70 participants from all ATLAS sub-detectors and the T-DAQ community. The quite dense agenda allowed nevertheless for many lively discussions, and for a dinner in the old town of Annecy. The Sessions The Workshop was organized in five main sessions: Assumptions and recommendations Sub-de...

Temperature actuated automatic safety rod release

Science.gov (United States)

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.

Shielding device for control rod in nuclear reactor

International Nuclear Information System (INIS)

Sakamaki, Kazuo; Tomatsu, Tsutomu.

1995-01-01

The device of the present invention shields radiation emitted from control rods to greatly reduce an operator's radiation exposure even if reactor water level is lowered and the upper portion of the control rod is exposed upon inspection of a BWR type reactor. Namely, a shield assembly has a structure comprising a set of four columnar shields in a two-row and two-column arrangement, which can be inserted into a control rod guide tube. Upon conducting inspection, the control rod is lowered into the control rod guide tube, and in this state, the columnar shields of the shield assembly are inserted to the control rod in the control rod guide tube. With such procedures, the upper portion of the control rod protruded from the control rod guide tube is covered with the shield assembly. As a result, radiation leaked from the control rod is shielded. Accordingly, irradiation in the reactor due to leaked radiation can be prevented thereby enabling to reduce an operator's radiation exposure. (I.S.)

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

Science.gov (United States)

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.

Burnable poison rod

International Nuclear Information System (INIS)

Natsume, Tomohiro.

1988-01-01

Purpose: To increase the reactor core lifetime by decreasing the effect of neutron absorption of burnable poison rods by using material with less neutron absorbing effect. Constitution: Stainless steels used so far as the coating material for burnable poison rods have relatively great absorption in the thermal neutral region and are not preferred in view of the neutron economy. Burnable poison rods having fuel can made of zirconium alloy shows absorption the thermal neutron region lower by one digit than that of stainless steels but they shows absorption in the resonance region and the cost is higher. In view of the above, the fuel can of the burnable poison material is made of aluminum or aluminu alloy. This can reduce the neutron absorbing effect by stainless steel fuel can and effectively utilize neutrons that have been wastefully absorbed and consumed in stainless steels. (Takahashi, M.)

Simulation of leaking fuel rods

International Nuclear Information System (INIS)

Hozer, Z.

2006-01-01

The behaviour of failed fuel rods includes several complex phenomena. The cladding failure initiates the release of fission product from the fuel and in case of large defect even urania grains can be released into the coolant. In steady state conditions an equilibrium - diffusion type - release is expected. During transients the release is driven by a convective type leaching mechanism. There are very few experimental data on leaking WWER fuel rods. For this reason the activity measurements at the nuclear power plants provide very important information. The evaluation of measured data can help in the estimation of failed fuel rod characteristics and the prediction of transient release dynamics in power plant transients. The paper deals with the simulation of leaking fuel rods under steady state and transient conditions and describes the following new results: 1) A new algorithm has been developed for the simulation of leaking fuel rods under steady state conditions and the specific parameters of the model for the Paks NPP has been determined; 2) The steady state model has been applied to calculation of leaking fuel characteristics using iodine and noble gas activity measurement data; 3) A new computational method has been developed for the simulation of leaking fuel rods under transient conditions and the specific parameters for the Paks NPP has been determined; 4) The transient model has been applied to the simulation of shutdown process at the Paks NPP and for the prediction of the time and magnitude of 123 I activity peak; 5) Using Paks NPP data a conservative value has been determined for the upper limit of the 123 I release from failed fuel rods during transients

Control rod for HTGR type reactor

International Nuclear Information System (INIS)

Mogi, Haruyoshi; Saito, Yuji; Fukamichi, Kenjiro.

1990-01-01

Upon dropping control rod elements into the reactor core, impact shocks are applied to wire ropes or spines to possibly deteriorate the integrity of the control rods. In view of the above in the present invention, shock absorbers such as springs or bellows are disposed between a wire rope and a spine in a HTGR type reactor control rod comprising a plurality of control rod elements connected axially by means of a spine that penetrates the central portion thereof, and is suspended at the upper end thereof by a wire rope. Impact shocks of about 5 kg are applied to the wire rope and the spine and, since they can be reduced by the shock absorbers, the control rod integrity can be maintained and the reactor safety can be improved. (T.M.)

Why Rods and Cocci

Indian Academy of Sciences (India)

Bacteria exhibit a wide variety of shapes but the commonly studied species of bacteria are generally either spherical in shape which are called cocci (singular coccus) or have a cylindrical shape and are called rods or bacilli (singular bacillus). In reality rods and cocci are the ends of a continuum. Sonle of the cocci are.

Spacers for fuel rod clusters

International Nuclear Information System (INIS)

Jabsen, F.S.

1978-01-01

The proposition deals with the fixing of nuclear fuel element rods in a grid which consists of a number of crossed Zy-plates which form cells. The rectangular cells have projections which serve as spacers for the fuel rods. According to the invention there are additional butt straps which can be moved in such a way that insertion and extraction of the fuel rods can be done without obstruction and they can be spring-loaded hold in their final position. (UWI) [de

Rod consolidation at the West Valley Demonstration Project

International Nuclear Information System (INIS)

Bailey, W.J.

1986-12-01

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

21 CFR 876.4270 - Colostomy rod.

Science.gov (United States)

2010-04-01

... 21 Food and Drugs 8 2010-04-01 2010-04-01 false Colostomy rod. 876.4270 Section 876.4270 Food and... GASTROENTEROLOGY-UROLOGY DEVICES Surgical Devices § 876.4270 Colostomy rod. (a) Identification. A colostomy rod is a device used during the loop colostomy procedure. A loop of colon is surgically brought out through...

Control rod for FBR type reactor

International Nuclear Information System (INIS)

Nakai, Koichi.

1993-01-01

In a control rod for an LMFBR type reactor, a thermal resistor is disposed between a temperature sensitive cylinder and a cam unit support rod. A thermal expansion difference due to the temperature difference is caused between the temperature sensitive cylinder and the cam unit support rod only upon abrupt temperature change of coolants. A control rod shaft extending mechanism of downwardly depressing an absorbent portion by amplifying the thermal expansion difference by an extension link mechanism and the cam unit is provided. The thermal resistor comprises inconel 625 or like other steel of small heat conductivity. If a certain abnormality should cause to the reactor system to elevate the coolant temperature in the reactor elevates abruptly and the reactor shutdown system does not actuate, since the control rod extension shaft extends to urge the absorbent and lower the reactor core reactivity, so that leading to serious accident can be prevented surely. Further, the control rod extension shaft does not extend upon moderate temperature elevation in the usual startup and causes no unnecessary reactivity change. (N.H.)

Design of active-neutron fuel rod scanner

International Nuclear Information System (INIS)

Griffith, G.W.; Menlove, H.O.

1996-01-01

An active-neutron fuel rod scanner has been designed for the assay of fissile materials in mixed oxide fuel rods. A 252 Cf source is located at the center of the scanner very near the through hole for the fuel rods. Spontaneous fission neutrons from the californium are moderated and induce fissions within the passing fuel rod. The rod continues past a combined gamma-ray and neutron shield where delayed gamma rays above 1 MeV are detected. We used the Monte Carlo code MCNP to design the scanner and review optimum materials and geometries. An inhomogeneous beryllium, graphite, and polyethylene moderator has been designed that uses source neutrons much more efficiently than assay systems using polyethylene moderators. Layers of borated polyethylene and tungsten are used to shield the detectors. Large NaI(Tl) detectors were selected to measure the delayed gamma rays. The enrichment zones of a thermal reactor fuel pin could be measured to within 1% counting statistics for practical rod speeds. Applications of the rod scanner include accountability of fissile material for safeguards applications, quality control of the fissile content in a fuel rod, and the verification of reactivity potential for mixed oxide fuels. (orig.)

Hydraulically centered control rod

International Nuclear Information System (INIS)

Horlacher, W.R.; Sampson, W.T.; Schukei, G.E.

1981-01-01

A control rod suspended to reciprocate in a guide tube of a nuclear fuel assembly has a hydraulic bearing formed at its lower tip. The bearing includes a plurality of discrete pockets on its outer surface into which a flow of liquid is continuously provided. In one embodiment the flow is induced by the pressure head in a downward facing chamber at the end of the bearing. In another embodiment the flow originates outside the guide tube. In both embodiments the flow into the pockets produces pressure differences across the bearing which counteract forces tending to drive the rod against the guide tube wall. Thus contact of the rod against the guide tube is avoided

Spider and burnable poison rod combinations

International Nuclear Information System (INIS)

Edwards, G.T.; Schluderberg, D.C.

1980-01-01

An improved design of burnable poison rods and associated spiders used in fuel assemblies of pressurized water power reactor cores, is described. The rods are joined to the spider arms in a manner which is proof against the reactor core environment and yet allows the removal of the rods from the spider simply, swiftly and delicately. (U.K.)

A two-group study on the gadolinium particle depletion in light water reactor fuel rods

International Nuclear Information System (INIS)

Lee, C.

1989-01-01

The effect of gadolinia particles on the assembly criticality of a light water reactor was investigated using two 2-group models. The particle effect was calculated by comparing the criticalities of two fuel assemblies, each containing one gadolinia-poisoned rod. For purposes of comparison, both rods contained an equal quantity of gadolinia, but the gadolinia fraction in one rod was in particle form. It was assumed that one pseudo-isotope represented Gd-155 and Gd-157, while the other isotopes were not considered. A one-group model developed by Kenneth Hartley(KH), was expanded into a two-group model, using a flat distribution for the fast group neutron flux. Gadolinia density was uniformly reduced by fast neutrons and the gadolinia burnup-rate was increased. The transparency effect of the gadolinia core was also included in the two group-KH model, allowing predictions of smoother changes at the peak of Δk (difference between k of the particle rod assembly and k of the uniform rod assembly). The Oregon State University Collision Probability (OSUCP) two-group model was developed for the investigation of the inter-particle shielding effect. A collision probability method was used to calculate thermal flux, and the flat fast-group flux assumption was used. The results of this study indicated that for small, 10-micron particles, the KH model failed to predict correct Δk behavior for the two assemblies. However, for larger, 100-micron particles both models well in agreement for the Δk profile, and for 500-micron particles both models were in agreement on both the behavior and magnitude of Δk

Essentialism Regarding Human Nature in the Defence of Gender Equality in Education

Science.gov (United States)

Holma, Katariina

2007-01-01

In this article I consider contemporary philosophical conceptions of human nature from the point of view of the ideal of gender equality. My main argument is that an essentialist account of human nature, unlike what I take to be its two main alternatives (the subjectivist account and the cultural account), is able coherently to justify the…

Modeling and simulation performance of sucker rod beam pump

Energy Technology Data Exchange (ETDEWEB)

Aditsania, Annisa, E-mail: annisaaditsania@gmail.com [Department of Computational Sciences, Institut Teknologi Bandung (Indonesia); Rahmawati, Silvy Dewi, E-mail: silvyarahmawati@gmail.com; Sukarno, Pudjo, E-mail: psukarno@gmail.com [Department of Petroleum Engineering, Institut Teknologi Bandung (Indonesia); Soewono, Edy, E-mail: esoewono@math.itb.ac.id [Department of Mathematics, Institut Teknologi Bandung (Indonesia)

2015-09-30

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

Modeling and simulation performance of sucker rod beam pump

International Nuclear Information System (INIS)

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

2015-01-01

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

Detection device for control rod interference

International Nuclear Information System (INIS)

Saito, Noboru.

1984-01-01

Purpose: To enable to detect the mechanical interference or friction between a control rod and a channel box automatically, simply and rapidly. Constitution: A signal from a gate circuit and a signal from a comparison mechanism are inputted into an AND circuit if a control rod has not been displaced by a predetermined distance within a prescribed time Δt after the output of an insertion or withdrawal signal for the control rod, by which a control-rod-interference signal is outputted from the AND circuit. Accordingly, the interference between the control rod and the channel box can be detected automatically, easily and rapidly. Furthermore, by properly adjusting the prescribed time Δt set by the gate circuit, the degree of the interference can also be detected, whereby the safety and the reliability of the reactor can be improved significantly. (Horiuchi, T.)

New studies of the natural convection around a fuel rod of the BME training reactor with PIV/LIF technique

International Nuclear Information System (INIS)

Szijarto, R.; Aszodi, A.; Yamaji, B.

2011-01-01

In this paper the model of a fuel pin of the Training Reactor of Budapest University of Technology and Economics was investigated with Particle Image Velocimetry and Laser Induced Fluorescence measurement methods. An experimental setup was designed, built and optimized to investigate the natural convection around a model of a fuel pin of the Training Reactor. The processes were analysed using an electrically heated rod, which models the geometry of the fuel rods in the Training Reactor. The heated length of the model is the same as the active length of the real fuel rods. The rod is placed in a glass tank with a shape of a square-based prism. An additional cooling system ensures constant flow conditions around the rod. The setup consists of an additional flow channel box, the equivalent diameter of which is equal to the equivalent diameter of the real fuel assembly. Simultaneous measurements of velocity and temperature fields were performed in different vertical positions for both cases of natural convection with and without the flow channel box. The effect of the presence of the channel was analyzed, and a laminarizating influence was observed. The local heat transfer coefficient was calculated for every measurement. The two dimensional measurement techniques gave extensive results, the structure of the hydraulic and thermal boundary layer were fully analyzed. (Authors)

Non-parametric order statistics method applied to uncertainty propagation in fuel rod calculations

International Nuclear Information System (INIS)

Arimescu, V.E.; Heins, L.

2001-01-01

method, which is computationally efficient, is presented for the evaluation of the global statement. It is proved that, r, the expected fraction of fuel rods exceeding a certain limit is equal to the (1-r)-quantile of the overall distribution of all possible values from all fuel rods. In this way, the problem is reduced to that of estimating a certain quantile of the overall distribution, and the same techniques used for a single rod distribution can be applied again. A simplified test case was devised to verify and validate the methodology. The fuel code was replaced by a transfer function dependent on two input parameters. The function was chosen so that analytic results could be obtained for the distribution of the output. This offers a direct validation for the statistical procedure. Also, a sensitivity study has been performed to analyze the effect on the final outcome of the sampling procedure, simple Monte Carlo and Latin Hypercube Sampling. Also, the effect on the accuracy and bias of the statistical results due to the size of the sample was studied and the conclusion was reached that the results of the statistical methodology are typically conservative. In the end, an example of applying these statistical techniques to a PWR reload is presented together with the improvements and new insights the statistical methodology brings to fuel rod design calculations. (author)

Radioactive lightning rods waste treatment

International Nuclear Information System (INIS)

Vicente, Roberto; Dellamano, Jose C.; Hiromoto, Goro

2008-01-01

Full text: In this paper, we present alternative processes that could be adopted for the management of radioactive waste that arises from the replacement of lightning rods with attached Americium-241 sources. Lightning protectors, with Americium-241 sources attached to the air terminals, were manufactured in Brazil until 1989, when the regulatory authority overthrew the license for fabrication, commerce, and installation of radioactive lightning rods. It is estimated that, during the license period, about 75,000 such devices were set up in public, commercial and industrial buildings, including houses and schools. However, the policy of CNEN in regard to the replacement of the installed radioactive rods, has been to leave the decision to municipal governments under local building regulations, requiring only that the replaced rods be sent immediately to one of its research institutes to be treated as radioactive waste. As a consequence, the program of replacement proceeds in a low pace and until now only about twenty thousand rods have reached the waste treatment facilities The process of management that was adopted is based primarily on the assumption that the Am-241 sources will be disposed of as radioactive sealed sources, probably in a deep borehole repository. The process can be described broadly by the following steps: a) Receive and put the lightning rods in initial storage; b) Disassemble the rods and pull out the sources; c) Decontaminate and release the metal parts to metal recycling; d) Store the sources in intermediate storage; e) Package the sources in final disposal packages; and f) Send the sources for final disposal. Up to now, the disassembled devices gave rise to about 90,000 sources which are kept in storage while the design of the final disposal package is in progress. (author)

Snubber assembly for a control rod drive

International Nuclear Information System (INIS)

1976-01-01

A snubber cartridge assembly is described which is mounted to the nozzle of a control rod drive mechanism to insure that it will be located within the liquid filled section of a nuclear reactor vessel whenever the control rod drive is assembled thereto. The snubber assembly includes a piston-mounted proximate to the control rod connecting end of the control rod drive leadscrew to allow the piston to travel within the liquid filled snubber cartridge and controllable exhaust the liquid during a 'scram' condition. The snubber cartridge provides three separate areas of increasing resistance to piston travel to insure a speedy but safe 'scram' of the control rod into the reactor

Automated nuclear fuel rod pattern loading system

International Nuclear Information System (INIS)

Lambert, D.V.; Nyland, T.W.; Byers, J.W.; Haley, D.E. Jr.; Cioffi, J.V.

1990-01-01

This patent describes an apparatus for loading fuel rods in a desired pattern. It comprises: a carousel having a plurality of movable gondolas for stocking thereon fuel rods of known enrichments; an elongated magazine defining a matrix of elongated slots being open at their forward ends for receiving fuel rods; a workstation defining a fuel rod feed path; and a holder and indexing mechanism for movably supporting the magazine and being actuatable for moving the magazine along X-Y axes to successively align one at a time selected ones of the slots with the feed path for loading in the magazine the successive fuel rods in a desired enrichment pattern

Snubber assembly for a control rod drive

International Nuclear Information System (INIS)

Matthews, J.C.

1978-01-01

A snubber cartridge assembly is mounted to the nozzle of a control rod drive mechanism to insure that the snubber assembly will be located within the liquid filled section of a nuclear reactor vessel whenever the control rod drive is assembled thereto. The snubber assembly includes a piston mounted proximate to the control rod connecting end of the control rod drive leadscrew to allow the piston to travel within the liquid filled snubber cartridge and controllably exhaust liquid therefrom during a ''scram'' condition. The snubber cartridge provides three separate areas of increasing resistance to piston travel to insure a speedy but safe ''scram'' of the control rod into the reactor

Method and apparatus for inspection of nuclear fuel rods

International Nuclear Information System (INIS)

Wachter, W.J.

1977-01-01

A method and apparatus are provided for the inspection of nuclear fuel rods to detect defects or failures in such rods. Assemblies of fuel rods are immersed in water and means are provided for causing a change in the relative pressures in the water and within the fuel rod such that fluid is expelled from the rod through any defects that may exist. Means are also provided for thereafter vibrating the rods to cause additional internal fluid or other material that may be trapped in the rod to be expelled. Sensors are provided for detecting the emission of bubbles of fluid or other material from the rod and for locating the position of the defective rod in the assembly. 5 figures

Positioning drive for absorber rods of a nuclear reactor

International Nuclear Information System (INIS)

Acher, H.

1977-01-01

The invention concerns a positioning drive for absorber rods of a nuclear reactor, of a threaded spindle and traveling nut type. In this positioning drive, rollers are provided the nut, which engage with the threads of the spindle and have an axis extending essentially at right angles to the longitudinal axis of the spindle. Three of the rollers are preferably combined in a traveling-nut housing, by means of anti-friction bearing elements. The positioning speed of such mechanical spindle drives can be increased thereby substantially. The invention is of interest particularly for boiling-water reactors. 9 claims, 8 figures

Mechanical properties of bioresorbable self-reinforced posterior cervical rods.

Science.gov (United States)

Savage, Katherine; Sardar, Zeeshan M; Pohjonen, Timo; Sidhu, Gursukhman S; Eachus, Benjamin D; Vaccaro, Alexander

2014-04-01

A biomechanical study. To test the mechanical and physical properties of self-reinforced copolymer bioresorbable posterior cervical rods and compare their mechanical properties to commonly used Irene titanium alloy rods. Bioresorbable instrumentation is becoming increasingly common in surgical spine procedures. Compared with metallic implants, bioresorbable implants are gradually reabsorbed as the bone heals, transferring the load from the instrumentation to bone, eliminating the need for hardware removal. In addition, bioresorbable implants produce less stress shielding due to a more physiological modulus of elasticity. Three types of rods were used: (1) 5.5 mm copolymer rods and (2) 3.5 mm and (3) 5.5 mm titanium alloy rods. Four tests were used on each rod: (1) 3-point bending test, (2) 4-point bending test, (3) shear test, and (4) differential scanning calorimeter test. The outcomes were recorded: Young modulus (E), stiffness, maximum load, deflection at maximum load, load at 1.0% strain of the rod's outer surface, and maximum bending stress. The Young modulus (E) for the copolymer rods (mean range, 6.4-6.8 GPa) was significantly lower than the 3.5 mm titanium rods (106 GPa) and the 5.5 mm titanium rods (95 GPa). The stiffness of the copolymer rods (mean range, 16.6-21.4 N/mm) was also significantly lower than the 3.5 mm titanium alloy rods (43.6 N/mm) and the 5.5 mm titanium alloy rods (239.6 N/mm). The mean maximum shear load of the copolymer rods was 2735 N and they had significantly lower mean maximum loads than the titanium rods. Copolymer rods have adequate shear resistance, but less load resistance and stiffness compared with titanium rods. Their stiffness is closer to that of bone, causing less stress shielding and better gradual dynamic loading. Their use in semirigid posterior stabilization of the cervical spine may be considered.

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

Energy Technology Data Exchange (ETDEWEB)

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

1994-06-01

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

Absorber rod drive for nuclear reactors

International Nuclear Information System (INIS)

Acher, H.

1985-01-01

The invention concerns a further addition to the invention of DE 33 42 830 A1. The free contact of the hollow piston with the nut due to hydraulic pressure is replaced by a hydraulic or spring attachment. The pressure system required to produce the hydraulic pressure is therefore omitted, and the electrical power required for driving the pump or the mass flow is also omitted. The absorber rod slotted along its longitudinal axis is replaced by an absorber rod, in the longitudinal axis of which a hollow piston is connected together with the absorber rod. This makes the absorber rod more stable, and assembly is simplified. (orig./HP) [de

Nuclear reactor fuel rod attachment system

International Nuclear Information System (INIS)

Christiansen, D.W.

1983-01-01

The invention involves a technique to quickly, inexpensively and rigidly attach a nuclear reactor fuel rod to a support member. The invention also allows for the repeated non-destructive removal and replacement of the fuel rod. The proposed fuel rod and support member attachment and removal system consists of a locking cap fastened to the fuel rod and a locking strip fastened to the support member or vice versa. The locking cap has two or more opposing fingers shaped to form a socket. The fingers spring back when moved apart and released. The locking strip has an extension shaped to rigidly attach to the socket's body portion

Temperature escalation in PWR fuel rod simulators due to the zircaloy/steam reaction ESSI-4 ESSI-11

International Nuclear Information System (INIS)

Hagen, S.; Kapulla, H.; Malauscheck, H.; Wallenfels, K.P.; Buescher, B.J.

1985-03-01

The tests had the initial heatup rate as main parameter. The experimental arrangement consisted of a fuel rod simulator (central tungsten heater, UO 2 ring pellets and zircaloy cladding), a zircaloy shroud and the fiber ceramic insulation. A steam flow of ca. 20 g/min was introduced at the lower end of the bundle. A temperature escalation was observed in every test. The maximum cladding surface temperature in the single rod tests never exceeded 2200 0 C. The escalation began in the upper region of the rods and moved down the rods, opposite to the direction of steam flow. For fast initial heatup rates, the runoff of molten zircaloy was a limiting process for the escalation. For slow heatup rates, the formation of a protective oxide layer reduced the reaction rate. The test with less insulation thickness showed a reduction of the escalation. A stronger influence was found for the gap between shroud and insulation. This is caused by convection heat losses to the steam circulating in this gap by natural convection. Removal of the gap between shroud and insulation in essentially the same experimental arrangement produced a faster escalation. The posttest appearance of the fuel rod simulators showed that, at slow heatup rates oxidation of the cladding was complete, and the fuel rod was relatively intact. Conversely, at fast heatup rates, relatively little cladding oxidation with extensive dissolution of the UO 2 pellets and runoff of molten cladding was observed. (orig./HP) [de

Control device for the withdrawal of control rod

International Nuclear Information System (INIS)

Ando, Masaki.

1985-01-01

Purpose: To significantly suppress the maximum value of the control-rod worth upon control rod withdrawal. Constitution: At first, a signal for designating the first class is sent from a class-control section to the group-control section. In the group-control section, the peripheral group among the first class is designated by which the withdrawal of the control rods other than the peripheral group is inhibited and the control-rods in the peripheral group are withdrawn one by one. When all of them have been withdrawn, the group-control section designates the central group of the first class. All the control rods of the central group have been withdrawn, then the group-control section designates the peripheral group of the second class. Thereafter, the central group in the second class is designated. The control rods are thus withdrawn in the same manner hereinafter. The maximum value for the control-rod worth can be decreased by such a withdrawing sequence for the control rods. (Horiuchi, T.)

Control rod drive mechanism

International Nuclear Information System (INIS)

Mizuno, Katsuyuki.

1976-01-01

Object: To restrict the reduction in performance due to stress corrosion cracks by making use of condensate produced in a turbine steam condenser. Structure: Water produced in a turbine steam condenser is forced into a condensed water desalting unit by low pressure condensate pump. The condensate is purified and then forced by a high pressure condensate pump into a feedwater heater for heating before it is returned to the reactor by a feedwater pump. Part of the condensate issuing from the condensate desalting unit is branched from the remaining portion at a point upstream the pump and is withdrawn into a control rod drive water pump after passing through a motordriven bypass valve, an orifice and a condenser water level control valve, is pressurized in the control rod drive water desalting unit and supplied to a control rod drive water pressure system. The control rod is vertically moved by the valve operation of the water pressure system. Since water of high oxygen concentration does not enter during normal operation, it is possible to prevent the stress cracking of the stainless steel apparatus. (Nakamura, S.)

Temperature measurement in cans of fuel rods and fuel rod simulators

International Nuclear Information System (INIS)

Tschoeke, H.; Moeller, R.

1977-01-01

On the sodium-cooled 19-rod cluster model for the SNR 300 the can wall temperature distributions of the non-uniformly cooled rods were measured with thermocouples mounted in outer grooves in the peripheral zone, permitting, in connection with Ni solder, a practically undisturbed measurement. For a more exact determination of the local surface temperature a calibration method, the so-called double-wall method, was developed and applied. The description of this calibration method and the experimental results achieved until now are presented. (orig./RW) [de

Analytic study of transverse shunt resistance and even-odd mode coupling of a rod type RFQ

International Nuclear Information System (INIS)

Koscielniak, S.

1994-06-01

To minimize the ohmic power losses, it is necessary to maximize the transverse shunt resistance, R shunt . The cell of a rod-type RFQ is modelled by a parallel two-rod transmission line supported above a parallel ground conductor by two legs. Due to coupling between neighboring supports, the loading impedance is modified depending on the leg spacing. The shunt resistance is improved by reducing the cell length and increasing the leg spacing, and maximized when the legs are equally spaced. However, this is also the condition for strong excitation of the unwanted 'even-mode' in which a potential difference exists between the ends of the rods mid-plane and the grounding conductor or tank, Once the legs of the support are longitudinally separated, some even-mode excitation of the structure is inevitable because some current must be injected into the ground conductor; the even-mode excitation rises as leg separation increases. Further, when the desired odd-mode voltage is symmetric about the cell centre, the even-mode voltage is anti-symmetric This paper is a very much abridged version of two internal design notes[3], [4]. (author). 4 refs.,1 fig

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

Science.gov (United States)

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.

Fabrication Of Control Rod System Of The RSG-GAS

International Nuclear Information System (INIS)

Sudirdjo, Hari; Setyono; Prasetya, Hendra

2001-01-01

Eight units of control rod mechanical system of RSG-GAS has been fabricated. The control rod mechanical system of RSG-GAS consist of guide tube and lifting rod. Complete construction of the control rod mechanical system of RSG-GAS are guide tube, lifting rod, absorber, and absorber casing. The eight units of the control rod mechanical system of RSG-GAS has been fabricated according to the mechanical engineering design

Microcomputer system for controlling fuel rod length

International Nuclear Information System (INIS)

Meyer, E.R.; Bouldin, D.W.; Bolfing, B.J.

1979-01-01

A system is being developed at the Oak Ridge National Laboratory (ORNL) to automatically measure and control the length of fuel rods for use in a high temperature gas-cooled reactor (HTGR). The system utilizes an LSI-11 microcomputer for monitoring fuel rod length and for adjusting the primary factor affecting length. Preliminary results indicate that the automated system can maintain fuel rod length within the specified limits of 1.940 +- 0.040 in. This system provides quality control documentation and eliminates the dependence of the current fuel rod molding process on manual length control. In addition, the microcomputer system is compatible with planned efforts to extend control to fuel rod fissile and fertile material contents

How to diagnose equal opportunities between women and men in organizations

Directory of Open Access Journals (Sweden)

Amaia Lusa

2009-12-01

Full Text Available Gender equality is now a matter that concerns all European countries. Despite the great effort made by governments and the progress of laws, gender discrimination still exists in family, social, cultural, political and economic spheres. Even today, it is rare to find women at the top level in companies. The path to equal opportunities is long and the only way to improve the situation is by means of appropriate equal opportunities plans. An essential part of the process of designing and implementing such action plans involves diagnosing the situation in a company. In this paper, an innovative indicator model is proposed. A review of numerous international private and public equal opportunities plans led to the design of a model to carry out a diagnose to identify all areas that could be corrected or improved by implementing specific measures, leading the company to deliver real equal opportunities policies. The indicator model is designed for the present historical context, though it is flexible enough to be adapted to each age and situation. The indicator model, which is original, will help companies carrying out one of the most important and essential parts of the process of designing and implementing and equal opportunities plan. Moreover, taking the indicator model companies ensure that all the relevant features are covered.

Control rod for a nuclear reactor

Science.gov (United States)

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.

Control rod for a nuclear reactor

International Nuclear Information System (INIS)

Roman, W.G.; Sutton, H.G. Jr.

1976-01-01

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 resilent members attached to the drive shaft into or out of shouldered engagement, respectively, with an indentation formed in the control rod

Control rod for a nuclear reactor

International Nuclear Information System (INIS)

Roman, W.G.; Sutton, H.G. Jr.

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

ABWR-II Core Design with Spectral Shift Rods for Operation with All Control Rods Withdrawn

International Nuclear Information System (INIS)

Moriwaki, Masanao; Aoyama, Motoo; Anegawa, Takafumi; Okada, Hiroyuki; Sakurada, Koichi; Tanabe, Akira

2004-01-01

An innovative reactor core concept applying spectral shift rods (SSRs) is proposed to improve the plant economy and the operability of the 1700-MW(electric) Advanced Boiling Water Reactor II (ABWR-II). The SSR is a new type of water rod in which a water level is naturally developed during operation and changed according to the coolant flow rate through the channel. By taking advantage of the large size of the ABWR-II bundle, the enhanced spectral shift operation by eight SSRs allows operation of the ABWR-II with all control rods withdrawn. In addition, the uranium-saving factor of 6 to 7% relative to the reference ABWR-II core with conventional water rods can be expected due to the greater effect of spectral shift. The combination of these advantages means the ABWR-II with SSRs should be an attractive alternative for the next-generation nuclear reactor

Process and apparatus for controlling control rods

International Nuclear Information System (INIS)

Gebelin, B.; Couture, R.

1987-01-01

This process and apparatus is characterized by 2 methods, for examination of cluster of nuclear control rods. Foucault current analyzer which examines fraction by fraction all the control rods. This examination is made by rotation of the cluster. Doubtful rods are then analysed by ultrasonic probe [fr

Performance of the NRX shut-off rods

International Nuclear Information System (INIS)

Manson, R.E.

1965-08-01

A new type of shut-off rod of electromechanical design was developed by the American Machine and Foundry Company for use in the NRX reactor following the accident of 1952. The new rods were installed in May, 1956, as part of the control system conversion program which was completed in 1958. Some problems were encountered with limit switch adjustment but minor modifications in design led to much improved operation. he performance of the rods also improved as more experience was gained in the maintenance and adjustment of the various headgear components. Each headgear is now overhauled once a year on a routine basis. The present design of shut-off rod is considered to be very satisfactory. There has only been one occasion when a shut-off rod has failed to come fully down on a trip. Rods have failed to operate correctly on five other occasions but these occurred during shutdown periods or when the reactor was being shutdown manually. (author)

Control rod supporting device in reactor

International Nuclear Information System (INIS)

Seki, Osamu; Itooka, Satoshi; Harada, Kiyoshi; Jodoi, Takashi.

1990-01-01

Since coolants flowing from a reactor core hit against a control rod and a control rod connection pipe, a considerable amount of bending moment for separating an attracting surface between an electromagnet and an armature is formed. Then, a plurality of grooves are formed on a heat sensitive material to dispose a heat collecting fin, and each of upper and lower contact portions of a control rod supporting portion in which the flanged portion of T-like cross section does not slip out is made into a partial spheric surface and a portion between the electromagnet and the attracted member are engaged by the unevenness. With such a constitution, even if a bending moment is applied, the control rod only swings and the bending moment is not transmitted to the attracted member. Further, since the temperature of the heat sensitive material can be rapidly made closer to the peripheral temperature by using the heat collecting fin, the timing for separation is made accurate. Further, since the engaging portion is brought into contact at the spheric surface, the load distribution on the control rod is made uniform, and the positional relationship is made accurate, to support the control rod reliably and the separation depends only on the temperature of the coolants. (N.H.)

Cuisenaire Rods Go to College.

Science.gov (United States)

Chinn, Phyllis; And Others

1992-01-01

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

Control rod driving mechanism

International Nuclear Information System (INIS)

Ooshima, Yoshio.

1983-01-01

Purpose: To perform reliable scram operation, even if abnormality should occur in a system instructing scram operation in FBR type reactors. Constitution: An aluminum alloy member to be melt at a predetermined temperature (about 600sup(o)C) is disposed to a connection part between a control rod and a driving mechanism, whereby the control rod is detached from the driving mechanism and gravitationally fallen to the reactor core. (Ikeda, J.)

Expandable device for a nuclear fuel rod

International Nuclear Information System (INIS)

Gesinski, L.T.

1978-01-01

A nuclear fuel rod and a device for use within the rod cladding to maintain the axial position of the fuel pellets stacked one atop another within the cladding are described. The device is initially of a smaller external cross-section than the fuel rod cladding internal cross-section so as to accommodate loading into the rod at preselected locations. During power operation the device responds to a rise in temperature, so as to permanently maintain its position and restrain any axial motion of the fuel pellets

Effects of ceftazidime and ciprofloxacin on biofilm formation in Proteus mirabilis rods.

Science.gov (United States)

Kwiecińska-Piróg, Joanna; Bogiel, Tomasz; Gospodarek, Eugenia

2013-10-01

Proteus mirabilis rods are one of the most commonly isolated species of the Proteus genus from human infections, mainly those from the urinary tract and wounds. They are often related to biofilm structure formation. The bacterial cells of the biofilm are less susceptible to routinely used antimicrobials, making the treatment more difficult. The aim of this study was to evaluate quantitatively the influence of ceftazidime and ciprofloxacin on biofilm formation on the polyvinyl chloride surface by 42 P. mirabilis strains isolated from urine, purulence, wound swab and bedsore samples. It has been shown that ceftazidime and ciprofloxacin at concentrations equal to 1/4, 1/2 and 1 times their MIC values for particular Proteus spp. strains decrease their ability to form biofilms. Moreover, ciprofloxacin at concentrations equal to 1/4, 1/2 and 1 times their MIC values for particular P. mirabilis strains reduces biofilm formation more efficiently than ceftazidime at the corresponding concentration values.

Method and apparatus for compacting spent nuclear reactor fuel rods

International Nuclear Information System (INIS)

Wachter, W.J.

1988-01-01

In a nuclear reactor system requiring periodic physical manipulation of spent fuel rods, the method of compacting fuel rods from a fuel rod assembly is described comprising the steps of: (1) removing the top end from pulling members having electrodes of weld elements in leading ends thereof in sequence through a fuel rod container and thence through respective consolidating passages in a fuel-rod directing chamber; (3) welding the weld elements of the pulling members to the top end of respective fuel rods corresponding to the respective pulling members; (4) drawing each of the pulling members axially to draw the respective engaged fuel rods in one axial direction through the respective passages in the chamber to thereby consolidate the fuel rods into a compacted configuration of a cross-sectional area smaller than the cross-sectional area occupied thereby within the fuel rod assembly; and (5) drawing all of the engaged fuel rods concurrently and substantially parallel to one another to the one axial direction into the fuel rod container while maintaining the compacting configuration in a fuel rod density which is greater than that of the fuel rod density of the fuel rod assembly

ABC of women workers' rights and gender equality

CERN Document Server

2007-01-01

This second-editioned publication presents important information relevant to women workers in entries on sexual harassment, women in development, the glass ceiling and many more. With an easy-to-follow, this book provides an essential tool raising awareness and legal literacy on gender equality issues.

Shock analysis on hydraulic drive control rod during scram

International Nuclear Information System (INIS)

Song Wei; Qin Benke; Bo Hanliang

2013-01-01

Control rod hydraulic drive mechanism (CRHDM) is a new invention of Institute of Nuclear and New Energy Technology of Tsinghua University. The hydraulic absorber buffers the control rod when it scrams. The control rod fast drop impact experiment was conducted and the key parameters of control rod hydraulic buffering performance were obtained. Based on the test results and according to D'Alembert principle, the maximum inertial impact force on the control rod during the fast drop period was applied as equivalent static load force on the control rod. The deformations and stress distributions on the control rod in this worst case were calculated by using finite element software ABAQUS. Calculation results were compared with the experiment results, and it was verified that nonlinear transient dynamics analysis in this problem can be simplified as static analysis. Damage criterion of the control rod fast drop impact process was also given. And it lays foundation for optimal design of the control rod and hydraulic absorber. (authors)

Acoustic loading effects on oscillating rod bundles

International Nuclear Information System (INIS)

Lin, W.H.

1980-01-01

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

Control Rod Malfunction at the NRAD Reactor

Energy Technology Data Exchange (ETDEWEB)

Thomas L. Maddock

2010-05-01

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

Criticality experiments with low enriched UO2 fuel rods in water containing dissolved gadolinium

International Nuclear Information System (INIS)

Bierman, S.R.; Murphy, E.S.; Clayton, E.D.; Keay, R.T.

1984-02-01

The results obtained in a criticality experiments program performed for British Nuclear Fuels, Ltd. (BNFL) under contract with the United States Department of Energy (USDOE) are presented in this report along with a complete description of the experiments. The experiments involved low enriched UO 2 and PuO 2 -UO 2 fuel rods in water containing dissolved gadolinium, and are in direct support of BNFL plans to use soluble compounds of the neutron poison gadolinium as a primary criticality safeguard in the reprocessing of low enriched nuclear fuels. The experiments were designed primarily to provide data for validating a calculation method being developed for BNFL design and safety assessments, and to obtain data for the use of gadolinium as a neutron poison in nuclear chemical plant operations - particularly fuel dissolution. The experiments program covers a wide range of neutron moderation (near optimum to very under-moderated) and a wide range of gadolinium concentration (zero to about 2.5 g Gd/l). The measurements provide critical and subcritical k/sub eff/ data (1 greater than or equal to k/sub eff/ greater than or equal to 0.87) on fuel-water assemblies of UO 2 rods at two enrichments (2.35 wt % and 4.31 wt % 235 U) and on mixed fuel-water assemblies of UO 2 and PuO 2 -UO 2 rods containing 4.31 wt % 235 U and 2 wt % PuO 2 in natural UO 2 respectively. Critical size of the lattices was determined with water containing no gadolinium and with water containing dissolved gadolinium nitrate. Pulsed neutron source measurements were performed to determine subcritical k/sub eff/ values as additional amounts of gadolinium were successively dissolved in the water of each critical assembly. Fission rate measurements in 235 U using solid state track recorders were made in each of the three unpoisoned critical assemblies, and in the near-optimum moderated and the close-packed poisoned assemblies of this fuel

Control rod driving hydraulic pressure device

International Nuclear Information System (INIS)

Ishida, Kazuo.

1990-01-01

Discharged water after actuating control rod drives in a BWR type reactor is once discharged to a discharging header, then returned to a master control unit and, subsequently, discharged to a reactor by way of a cooling water header. The radioactive level in the discharging header and the master control unit is increased by the reactor water to increase the operator's exposure. In view of the above, a riser is disposed for connecting a hydraulic pressure control unit incorporating a directional control valve and the cooling water head. When a certain control rod is inserted, the pressurized driving water is supplied through a hydraulic pressure control unit to the control rod drives. The discharged water from the control rod drives is entered by way of the hydraulic pressure control unit into the cooling water header and then returned to the reactor by way of other hydraulic pressure control unit and the control rod drives. Thus, the reactor water is no more recycled to the master control unit to reduce the radioactive exposure. (N.H.)

Nuclear reactor with scrammable part length rod

International Nuclear Information System (INIS)

Bevilacqua, F.

1979-01-01

A new part length rod is provided. It may be used to control xenon induced power oscillations but to contribute to shutdown reactivity when a rapid shutdown of the reactor is required. The part length rod consists of a control rod with three regions. The lower control region is a longer weaker active portion separated from an upper stronger shorter poison section by an intermediate section which is a relative non-absorber of neutrons. The combination of the longer weaker control section with the upper high worth poison section permits the part length rod of this to be scrammed into the core when a reactor shutdown is required but also permits the control rod to be used as a tool to control power distribution in both the axial and radial directions during normal operation

Apparatus for loading fuel pellets in fuel rods

International Nuclear Information System (INIS)

Tedesco, R.J.

1976-01-01

An apparatus is disclosed for loading fuel pellets into fuel rods for a nuclear reactor including a base supporting a table having grooves therein for holding a multiplicity of pellets. Multiple fuel rods are placed in alignment with grooves in the pellet table and a guide member channels pellets from the table into the corresponding fuel rods. To effect movement of pellets inside the fuel rods without jamming, a number of electromechanical devices mounted on the base have arms connected to the lower surface of the fuel rod table which cyclically imparts a reciprocating arc motion to the table for moving the fuel pellets longitudinally of and inside the fuel rods. These electromechanical devices include a solenoid having a plunger therein connected to a leaf type spring, the arrangement being such that upon energization of the solenoid coil, the leaf spring moves the fuel rod table rearwardly and downwardly, and upon deenergization of the coil, the spring imparts an upward-forward movement to the table which results in physical displacement of fuel pellets in the fuel rods clamped to the table surface. 8 claims, 6 drawing figures

Control rod housing alignment and repair apparatus

International Nuclear Information System (INIS)

Dixon, R.C.; Deaver, G.A.; Punches, J.R.; Singleton, G.E.; Erbes, J.G.; Offer, H.P.

1991-01-01

This patent describes a welding a repair device for precisely locating and welding the position of the top of a control rod drive housing attached from a stub tube from a corresponding aperture and alignment pin in a core plate within a boiling water nuclear reactor, the welding and repair device. It comprises: a shaft, the shaft extending from the vicinity of the top of the control rod drive housing up to and through the aperture in the core plate; means for registering to the aperture and the alignment pin on the core plate; a fixture attached to the bottom end of the shaft for mating to the top of the control rod drive housing in precise mating relationship; the fixture attached to the bottom end of the shaft whereby the fixture, when mated to the control rod drove housing and the registering means when registered to the alignment pin and aperture on the core plate imparts to the shaft, and angularity between the top of the control rod drive housing and the hole in the core plate; a hollow cylinder, the cylinder mounted for depending and sealed support with respect to the shaft above, about and below the control rod drive housing top; the cylinder depending down below the control rod drive housing to an elevation below the top of the sub tube; a rotating welding apparatus with a welding head for dispensing weldment mounted for rotation with respect to the shaft; the welding head disposed at the juncture between the side of the control rod drive housing and the stub tube; and means for flooding the cylinder with gas whereby the cylinder may be lowered. flooded in a gas environment and effect a weld between the top of the stub tube and the control rod drive housing

Control rod housing alignment and repair method

International Nuclear Information System (INIS)

Dixon, R.C.; Deaver, G.A.; Punches, J.R.; Singleton, G.E.; Erbes, J.G.; Offer, H.P.

1992-01-01

This patent describes a method for underwater welding of a control rod drive housing inserted through a stub tube to maintain requisite alignment and elevation of the top of the control rod drive housing to an overlying and corresponding aperture in a core plate as measured by an alignment device which determines the relative elevation and angularity with respect to the aperture. It comprises providing a welding cylinder dependent from the alignment device such that the elevation of the top of the welding cylinder is in a fixed relationship to the alignment device and is gas-proof; pressurizing the welding cylinder with inert welding gas sufficient to maintain the interior of the welding cylinder dry; lowering the welding cylinder through the aperture in the core plate by depending the cylinder with respect to the alignment device, the lowering including lowering through and adjusting the elevation relationship of the welding cylinder to the alignment device such that when the alignment device is in position to measure the elevation and angularity of the new control rod drive housing, the lower distal end of the welding cylinder extends below the upper periphery of the stub where welding is to occur; inserting a new control rod drive housing through the stub tube and positioning the control rod drive housing to a predetermined relationship to the anticipated final position of the control rod drive housing; providing welding implements transversely rotatably mounted interior of the welding cylinder relative to the alignment device such that the welding implements may be accurately positioned for dispensing weldment around the periphery of the top of the stub tube and at the side of the control rod drive housing; measuring the elevation and angularity of the control rod drive housing; and dispensing weldment along the top of the stub tube and at the side of the control rod drive housing

Processing of poison rods with a view to disposal

International Nuclear Information System (INIS)

Bichet, R.; Charamathieu, A.; Lasseur, C.; Golicheff, I.; Pouteaux, M.

1979-01-01

In the core of the French 900 and 1300 MW reactors, a certain number of rods have to be processed as wastes, particularly the burnable poison rods used during reactor start-up (900 MW: 68 rods; 1300 MW: 96 rods). Several solutions are possible: cutting and conditionning in reactor pool; transfer of the poison rods to a cutting and conditionning facility; transfer of the poison rods and fuel assemblies to a storage area where they are cutted and stored. Each of these solutions are studied, the advantages and disadvantages being presented

Characterization of Emericella nidulans RodA and DewA hydrophobin mutants

DEFF Research Database (Denmark)

Jensen, Britt Guillaume; Nielsen, Jakob Blæsbjerg; Pedersen, Mona Højgaard

hydrophobins RodA and DewA. Individual knock-out mutants rodAΔ, dewAΔ and the double deletion strain rodAΔdewAΔ were constructed. Furthermore, two strains containing a point mutation in the first of the cysteines of RodA (rodA-C57G), where one was coupled to the dewA deletion, were included. The reference...... strain (NID1) and dewAΔ displayed green conidia. However, rodAΔ and rodAΔdewAΔ showed a dark green/brown conidial pigmentation, while rodA-C57G and rodAC57G dewAΔ displayed lighter brown conidia. rodAΔ and rodAΔdewAΔ displayed a higher degree of hülle cells compared to the moderate amount observed...... for NID1 and dewAΔ, while rodA-C57G and rodA-C57G dewAΔ displayed a low number of hülle cells. NID1 and dewAΔ conidia were dispersed as spore chains. rodAΔ, rodAΔdewAΔ, rodA-C57G and rodA-C57G dewAΔ spores were associated in large clumps, where the conidia seemed to adhere to one another. The largest...

Method for depleting BWRs using optimal control rod patterns

International Nuclear Information System (INIS)

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

1991-01-01

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

Failure position detection device for nuclear fuel rod

International Nuclear Information System (INIS)

Ishida, Takeshi; Higuchi, Shin-ichi; Ito, Masaru; Matsuda, Yasuhiko

1987-01-01

Purpose: To easily detect failure position of a nuclear fuel rod by relatively moving an air-tightly shielded detection portion to a fuel rod. Constitution: For detecting the failure position of a leaked fuel assembly, the fuel assembly is dismantled and a portion of withdrawn fuel rod is air-tightly sealed with an inspection portion. The inside of the inspection portion is maintained at a pressure-reduced state. Then, in a case if failed openings are formed at a portion sealed by the inspection portion in the fuel rod, FP gases in the fuel rod are released based on the reduced pressure and the FP gases are detected in the detection portion. Accordingly, by relatively moving the detection portion to the fuel rod, the failure position can be detected. (Yoshino, Y.)

Failure position detection device for nuclear fuel rod

Energy Technology Data Exchange (ETDEWEB)

Ishida, Takeshi; Higuchi, Shin-ichi; Ito, Masaru; Matsuda, Yasuhiko

1987-03-24

Purpose: To easily detect failure position of a nuclear fuel rod by relatively moving an air-tightly shielded detection portion to a fuel rod. Constitution: For detecting the failure position of a leaked fuel assembly, the fuel assembly is dismantled and a portion of withdrawn fuel rod is air-tightly sealed with an inspection portion. The inside of the inspection portion is maintained at a pressure-reduced state. Then, in a case if failed openings are formed at a portion sealed by the inspection portion in the fuel rod, FP gases in the fuel rod are released based on the reduced pressure and the FP gases are detected in the detection portion. Accordingly, by relatively moving the detection portion to the fuel rod, the failure position can be detected. (Yoshino, Y.).

Fuel followed control rod installation at AFRRI

International Nuclear Information System (INIS)

Moore, Mark; Owens, Chris; Forsbacka, Matt

1992-01-01

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

Solitary waves on nonlinear elastic rods. II

DEFF Research Database (Denmark)

Sørensen, Mads Peter; Christiansen, Peter Leth; Lomdahl, P. S.

1987-01-01

In continuation of an earlier study of propagation of solitary waves on nonlinear elastic rods, numerical investigations of blowup, reflection, and fission at continuous and discontinuous variation of the cross section for the rod and reflection at the end of the rod are presented. The results ar...... are compared with predictions of conservation theorems for energy and momentum....

The effect of different screw-rod design on the anti-rotational torque: a biomechanical comparison of three conventional screw-rod constructs.

Science.gov (United States)

Huang, Zifang; Wang, Chongwen; Fan, Hengwei; Sui, Wenyuan; Li, Xueshi; Wang, Qifei; Yang, Junlin

2017-07-28

Screw-rod constructs have been widely used to correct spinal deformities, but the effects of different screw-rod systems on anti-rotational torque have not been determined. This study aimed to analyze the biomechanical effect of different rod-screw constructs on anti-rotational torque. Three conventional spinal screw-rod systems (Legacy, RF-F-10 and USSII) were used to test the anti-rotational torque in the material test machine. ANOVA was performed to evaluate the anti-rotational capacity of different pedicle screws-rod constructs. The anti-rotational torque of Legacy group, RF-F-10 group and USSII group were 12.3 ± 1.9 Nm, 6.8 ± 0.4 Nm, and 3.9 ± 0.8 Nm, with a P value lower than 0.05. This results indicated that the Legacy screws-rod construct could provide a highest anti-rotation capacity, which is 68% and 210% greater than RF-F-10 screw-rod construct and USSII screw-rod respectively. The anti-rotational torque may be mainly affected by screw cap and groove design. Our result showed the anti-rotational torque are: Legacy system > RF-F-10 system > USSII system, suggesting that appropriate rod-screw constructs selection in surgery may be vital for anti-rotational torque improvement and preventing derotation correction loss.

Cutting system for burnable poison rod

International Nuclear Information System (INIS)

Shiina, Atsushi; Toyama, Norihide; Koshino, Yasuo; Fujii, Toshio

1989-01-01

Burnable poison rods attached to spent fuels are contained in a containing box and transported to a receiving pool. The burnable poison rod-containing box is provisionally situated by the operation to a handling device to a provisional setting rack in a cutting pool and attached to a cutting guide of a cutting device upon cutting. The burnable poison rod is cut only in a cutting pool water and tritium generated upon cutting is dissolved into the cutting pool water. Diffusion of tritium is thus restricted. Further, the cutting pool is isolated by a partition device from the receiving pool during cutting of the burnable poison rod. Accordingly, water in which tritium is dissolved is inhibited from moving to the receiving pool and prevail of tritium contamination can be avoided. (T.M.)

Brownian dynamics simulations with stiff finitely extensible nonlinear elastic-Fraenkel springs as approximations to rods in bead-rod models.

Science.gov (United States)

Hsieh, Chih-Chen; Jain, Semant; Larson, Ronald G

2006-01-28

A very stiff finitely extensible nonlinear elastic (FENE)-Fraenkel spring is proposed to replace the rigid rod in the bead-rod model. This allows the adoption of a fast predictor-corrector method so that large time steps can be taken in Brownian dynamics (BD) simulations without over- or understretching the stiff springs. In contrast to the simple bead-rod model, BD simulations with beads and FENE-Fraenkel (FF) springs yield a random-walk configuration at equilibrium. We compare the simulation results of the free-draining bead-FF-spring model with those for the bead-rod model in relaxation, start-up of uniaxial extensional, and simple shear flows, and find that both methods generate nearly identical results. The computational cost per time step for a free-draining BD simulation with the proposed bead-FF-spring model is about twice as high as the traditional bead-rod model with the midpoint algorithm of Liu [J. Chem. Phys. 90, 5826 (1989)]. Nevertheless, computations with the bead-FF-spring model are as efficient as those with the bead-rod model in extensional flow because the former allows larger time steps. Moreover, the Brownian contribution to the stress for the bead-FF-spring model is isotropic and therefore simplifies the calculation of the polymer stresses. In addition, hydrodynamic interaction can more easily be incorporated into the bead-FF-spring model than into the bead-rod model since the metric force arising from the non-Cartesian coordinates used in bead-rod simulations is absent from bead-spring simulations. Finally, with our newly developed bead-FF-spring model, existing computer codes for the bead-spring models can trivially be converted to ones for effective bead-rod simulations merely by replacing the usual FENE or Cohen spring law with a FENE-Fraenkel law, and this convertibility provides a very convenient way to perform multiscale BD simulations.

Analysis of buffering process of control rod hydraulic absorber

International Nuclear Information System (INIS)

Bao Jishi; Qin Benke; Bo Hanliang

2011-01-01

Control Rod Hydraulic Drive Mechanism(CRHDM) is a newly invented build-in control rod drive mechanism. Hydraulic absorber is the key part of this mechanism, and is used to cushion the control rod when the rod scrams. Thus, it prevents the control rod from being deformed and damaged. In this paper dynamics program ANSYS CFX is used to calculate all kinds of flow conditions in hydraulic absorber to obtain its hydraulic characteristics. Based on the flow resistance coefficients obtained from the simulation results, fluid mass and momentum equations were developed to get the trend of pressure change in the hydraulic cylinder and the displacement of the piston rod during the buffering process of the control rod. The results obtained in this paper indicate that the hydraulic absorber meets the design requirement. The work in this paper will be helpful for the design and optimization of the control rod hydraulic absorber. (author)

Control rod experiments in Racine

International Nuclear Information System (INIS)

Stanculescu, A.; Humbert, G.

1981-09-01

A survey of the control-rod experiments planned within the joint CEA/CNEN-DeBeNe critical experiment RACINE is given. The applicability to both heterogeneous and homogeneous large power LMFBR-cores is discussed. Finally, the most significant results of the provisional design calculations performed on behalf of the RACINE control-rod programme are presented

RODMOD: a code for control rod positioning

International Nuclear Information System (INIS)

Vondy, D.R.; Fowler, T.B.

1978-11-01

The report documents a computer code which has been implemented to position control rods according to a prescribed schedule during the calculation of a reactor history. Control rods may be represented explicitly with or without internal black absorber conditions in selected energy groups, or fractional insertion may be done, or both, in a problem. There is provision for control rod follower, movement of materials through a series of zones in a closed loop, and shutdown rod insertion and subsequent removal to allow the reactor history calculation to be continued. This code is incorporated in the system containing the VENTURE diffusion theory neutronics and the BURNER exposure codes for routine use. The implemented automated procedures cause the prescribed control rod insertion schedule to be applied without the access of additional user input data during the calculation of a reactor operating history

French LMFBR's control rods experience and development

International Nuclear Information System (INIS)

Arnaud, G.; Guigon, A.; Verset, L.

1983-06-01

Since the last ten years, the French program has been, first of all, directed to the setting up, and then the development of, at once, the Phenix control rods, and next, the Super-Phenix ones. The vented pin design, with porous plug and sodium bonding, which allows the choices of large diameters, has been taken, since the Rapsodie experience was decisive. The absorber material is sintered, 10 B enriched, boron carbide. The can is made of 316 type stainless steel, stabilised, or not, with titanium. The experience gained in Phenix up to now is important, and deals with about six loads of control rods. Results confirm the validity of the design of the absorber pins. Some difficulties has been encountered for the guiding devices, due to the swelling of the steel. They have required design and material improvements. Such difficulties are discarded by a new design of the bearing, for the Super-Phenix control rods. The other parts of these rods, from the Primary Shut-Down System, are strictly derived from Phenix. The design of the rods from the Secondary Shut-Down System is rather different, but it's not the case for the design of the absorber pins: in many a way, they are derived from Phenix pins and from Rapsodie control rods. Both types of rods irradiation tests are in progress in Phenix [fr

Advanced gray rod control assembly

Science.gov (United States)

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

2013-09-17

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

Refabricated and instrumented fuel rods

International Nuclear Information System (INIS)

Silberstein, K.

2005-01-01

Nuclear Fuel for power reactors capabilities evaluation is strongly based on the intimate knowledge of its behaviour under irradiation. This knowledge can be acquired from refabricated and instrumented fuel rods irradiated at different levels in commercial reactors. This paper presents the development and qualification of a new technique called RECTO related to a double-instrumented rod re-fabrication process developed by CEA/LECA hot laboratory facility at CADARACHE. The technique development includes manufacturing of the properly dimensioned cavity in the fuel pellet stack to house the thermocouple and the use of a newly designed pressure transducer. An analytic irradiation of such a double-instrumented fuel rod will be performed in OSIRIS test reactor starting October 2004. (Author)

Reactor control rod supporting structure

International Nuclear Information System (INIS)

Akimoto, Tokuzo; Miyata, Hiroshi.

1984-01-01

Purpose: To enable stable reactor core control even in extremely great vertical earthquakes, as well as under normal operation conditions in FBR type reactors. Constitution: Since a mechanism for converting the rotational movement of a control rod into vertical movement is placed at the upper portion of the reactor core at high temperature, the mechanism should cause fusion or like other danger after the elapse of a long period of time. In view of the above, the conversion mechanism is disposed to the lower portion of the reactor core at a lower temperature region. Further, the connection between the control rod and the control rod drive can be separated upon great vertical earthquakes. (Seki, T.)

Age-related deterioration of rod vision in mice.

Science.gov (United States)

Kolesnikov, Alexander V; Fan, Jie; Crouch, Rosalie K; Kefalov, Vladimir J

2010-08-18

Even in healthy individuals, aging leads to deterioration in visual acuity, contrast sensitivity, visual field, and dark adaptation. Little is known about the neural mechanisms that drive the age-related changes of the retina and, more specifically, photoreceptors. According to one hypothesis, the age-related deterioration in rod function is due to the limited availability of 11-cis-retinal for rod pigment formation. To determine how aging affects rod photoreceptors and to test the retinoid-deficiency hypothesis, we compared the morphological and functional properties of rods of adult and aged B6D2F1/J mice. We found that the number of rods and the length of their outer segments were significantly reduced in 2.5-year-old mice compared with 4-month-old animals. Aging also resulted in a twofold reduction in the total level of opsin in the retina. Behavioral tests revealed that scotopic visual acuity and contrast sensitivity were decreased by twofold in aged mice, and rod ERG recordings demonstrated reduced amplitudes of both a- and b-waves. Sensitivity of aged rods determined from single-cell recordings was also decreased by 1.5-fold, corresponding to not more than 1% free opsin in these photoreceptors, and kinetic parameters of dim flash response were not altered. Notably, the rate of rod dark adaptation was unaffected by age. Thus, our results argue against age-related deficiency of 11-cis-retinal in the B6D2F1/J mouse rod visual cycle. Surprisingly, the level of cellular dark noise was increased in aged rods, providing an alternative mechanism for their desensitization.

Optimized Control Rods of the BR2 Reactor

Energy Technology Data Exchange (ETDEWEB)

Kalcheva, Silva; Koonen, E.

2007-09-15

At the present time the BR-2 reactor uses control elements with cadmium as neutron absorbing part. The lower section of the control element is a beryllium assembly cooled by light water. Due to the burn up of the lower end of the cadmium section during the reactor operation, the presently used rods for reactivity control of the BR-2 reactor have to be replaced by new ones. Considered are various types Control Rods with full active part of the following materials: cadmium (Cd), hafnium (Hf), europium oxide (Eu2O3) and gadolinium (Gd2O3). Options to decrease the burn up of the control rod material in the hot spot, such as use of stainless steel in the lower active part of the Control Rod are discussed. Comparison with the characteristics of the presently used Control Rods types is performed. The changing of the characteristics of different types Control Rods and the perturbation effects on the reactor neutronics during the BR-2 fuel cycle are investigated. The burn up of the Control Rod absorbing material, total and differential control rods worth, macroscopic and effective microscopic absorption cross sections, fuel and reactivity evolution are evaluated during approximately 30 operating cycles.

Control rod repositioning considerations in core design analysis

International Nuclear Information System (INIS)

Armstrong, B.C.; Buechel, R.J.

1990-01-01

Control rod repositioning is a method for minimizing rod cluster control assembly (RCCA) wear in the upper internals area where the guide cards interface with the rodlets of the RCCAs. A number of utilities have implemented strategies for rod repositioning, which often has no impact on the nuclear analysis for cases where the control rods are never repositioned into the active fuel. Other strategies involve repositioning the control rods several steps into the active fuel. The impact of this type of repositioning on the axial power shape and consequently the total peaking factor F Q T varies, depending on the method in which the repositioning is implemented at the plant. Operating for long periods with all the control and shutdown rods inserted several steps in the active fuel followed by withdrawing them fully from the core results in a shifting of the power distribution toward the top of the core and must be accounted for in the design analysis. On the other hand, an optional plan for control rod repositioning that considers margins available in related design parameters can be devised that minimizes the effects of the repositioning for the reload. This paper summarizes a rod repositioning strategy implemented for a recent reload and some calculated power shape results for this strategy and other scenarios

Optimized Control Rods of the BR2 Reactor

International Nuclear Information System (INIS)

Kalcheva, Silva; Koonen, E.

2007-01-01

At the present time the BR-2 reactor uses control elements with cadmium as neutron absorbing part. The lower section of the control element is a beryllium assembly cooled by light water. Due to the burn up of the lower end of the cadmium section during the reactor operation, the presently used rods for reactivity control of the BR-2 reactor have to be replaced by new ones. Considered are various types Control Rods with full active part of the following materials: cadmium (Cd), hafnium (Hf), europium oxide (Eu2O3) and gadolinium (Gd2O3). Options to decrease the burn up of the control rod material in the hot spot, such as use of stainless steel in the lower active part of the Control Rod are discussed. Comparison with the characteristics of the presently used Control Rods types is performed. The changing of the characteristics of different types Control Rods and the perturbation effects on the reactor neutronics during the BR-2 fuel cycle are investigated. The burn up of the Control Rod absorbing material, total and differential control rods worth, macroscopic and effective microscopic absorption cross sections, fuel and reactivity evolution are evaluated during approximately 30 operating cycles.

Conceptual design report of the SMART fuel rod

Energy Technology Data Exchange (ETDEWEB)

Kim, Dae Ho; Lee, Chan Bock; Bang, Je Gun; Jung, Yeon Ho [Korea Atomic Energy Research Institute, Taejon (Korea)

1999-03-01

The SMART fuel rod is based on 17 x 17 KOFA(Korea Fuel Assembly) fuel rod of the 950MWe pressurize water reactor. The fuel stack length of the KOFA is 3658mm, otherwise SMART fuel rod stack length is 2000mm. The fuel rod contains UO{sub 2} pellets with the enrichment of 4.95%. All the fuel in core will be replaced every 35 months. The average LHGR of the fuel rod is 120 W/cm, commercial PWR is 178 W/cm, SMART LHGR is lower about 31% than commercial PWR. The core inlet and outlet temperature of coolant are respectively 270 deg C and 310 deg C, commercial PWR are respectively 291.6 deg C and 326.8 deg C, SMART inlet and outlet temperature is lower averaged 19.2 deg C than commercial PWR. The coolant use mixed soluble ammonia in high purity water and boron is not in. The general performance of the fuel rod UO{sub 2} pellet has been already verified through the sufficient burnup (60,000 MWd/MTU-rod avg.) experience as the rods of same design in commercial PWR's. But cladding corrosion is required the further verification. (author). 13 refs., 3 figs., 8 tabs.

RodZ and PgsA play intertwined roles in membrane homeostasis of Bacillus subtilis and resistance to weak organic acid stress

Directory of Open Access Journals (Sweden)

Johan Willem Albertus Van Beilen

2016-10-01

Full Text Available Weak organic acids like sorbic and acetic acid are widely used to prevent growth of spoilage organisms such as Bacilli. To identify genes involved in weak acid stress tolerance we screened a transposon mutant library of Bacillus subtilis for sorbic acid sensitivity. Mutants of the rodZ (ymfM gene were found to be hypersensitive to the lipophilic weak organic acid. RodZ is involved in determining the cell’s rod-shape and believed to interact with the bacterial actin-like MreB cytoskeleton. Since rodZ lies upstream in the genome of the essential gene pgsA (phosphatidylglycerol phosphate synthase we hypothesized that expression of the latter might also be affected in rodZ mutants and hence contribute to the phenotype observed. We show that both genes are co-transcribed and that both the rodZ::mini-Tn10 mutant and a conditional pgsA mutant, under conditions of minimal pgsA expression, were sensitive to sorbic and acetic acid. Both strains displayed a severely altered membrane composition. Compared to the wild-type strain, phosphatidylglycerol and cardiolipin levels were lowered and the average acyl chain length was elongated. Induction of rodZ expression from a plasmid in our transposon mutant led to no recovery of weak acid susceptibility comparable to wild-type levels. However, pgsA overexpression in the same mutant partly restored sorbic acid susceptibility and fully restored acetic acid sensitivity. A construct containing both rodZ and pgsA as on the genome led to some restored growth as well. We propose that RodZ and PgsA play intertwined roles in membrane homeostasis and resistance to weak organic acid stress.

Results from In-pile experiments on LWR fuel rod behavior under LOCA conditions with unirradiated rods

International Nuclear Information System (INIS)

Sepold, L.; Karb, E.H.; Pruessmann, M.

1981-06-01

This report summarizes the results of the FR2-in-pile tests at KfK (Kernforschungszentrum Karlsruhe) with unirradiated test rods. The in-pile tests with the objective of investigating the influence of a nuclear environment on the mechanisms of fuel rod failure were being performed with irradiated and unirradiated single rods of a PWR design in the DK loop of the FR2 reactor. The main parameter of the test program was the burnup, ranging from 2.500 to 35.000 MWd/t. The program with unirradiated specimens comprised the series A and B with a total of 14 tests. (orig.) [de

Air-water two-phase flow in a four by four rod bundle with partial length rods

International Nuclear Information System (INIS)

Ohta, Motoki; Kamei, Akihiro; Mizutani, Yoshitaka; Hosokawa, Shigeo; Tomiyama, Akio

2009-01-01

Partial length rods (PLR) are used in fuel bundles of BWR to reduce pressure drops in two-phase regions and to optimize the power distribution. Since little is known about effects of PLR on two-phase flows, air-water two-phase flow around PLRs in a four by four rod bundle is visualized by using a high-speed video camera. The experimental apparatus consists of acrylic channel box and transparent rods. Air and water at atmospheric pressure and room temperature are used for the gas and liquid phases, respectively. The ranges of the gas and liquid volume fluxes, J G and J L , are 0.4 L G L , the flow pattern in the downstream of PLR transits to slug flow, and the flow patterns in the surrounding subchannels transit to bubbly flow due to the redistribution of gas flow. (2) In annular flow, the liquid film on the PLR forms a liquid column above the end cap of PLR. Droplets are generated by column breakup and deposit on liquid films on the neighboring rods. (3) The liquid film thickness on the surface of neighbor rods facing the PLR increases and it reduces that on their opposite surface in the downstream of PLR. (author)

Development of design technology on thermal-hydraulic performance in tight-lattice rod bundle. III - Numerical estimation on rod bowing effect based on X-ray CT data

International Nuclear Information System (INIS)

Misawa, Takeharu; Ohnuki, Akira; Katsuyama, Kozo; Nagamine, Tsuyoshi; Nakamura, Yasuo; Akimoto, Hajime; Mitsutake, Toru; Misawa, Susumu

2007-01-01

Design studies of the Innovative Water Reactor for Flexible Fuel Cycle (FLWR) are being carried out at the Japan Atomic Energy Agency (JAEA) as one candidate for the future reactors. In actual core design, it is precondition to prevent fuel rods contact due to fuel rod bowing. However, the FLWR cores have nonconventional characteristics such as a hexagonal tight lattice arrangement and a high enrichment fuel loading. Therefore, as conservative evaluation, it is important to investigate influence of fuel rod bowing upon the boiling transition. In the JAEA, a 37-rod bundle experiments (base case test section (1.3mm gap width), gap width effect test section (1.0mm gap width), and rod bowing test section) were performed in order to investigate the thermal hydraulic characteristics in the tight lattice bundle. In this paper, the rod bowing effect test is paid attention. It is suspected that the actual fuel rod positions in the rod bowing test section may be different from the design-based positions. Even a slight displacement from the design-based position of fuel rod may occur variation of flow area, and give influence upon the thermal hydraulic characteristics in the rod bundle. Therefore, if the critical power in the rod bundle is evaluated by an analytical approach, the analysis based on more correct input can be performed by using actual fuel rod position data. In this study, the rod positions in the rod bowing test section were measured using the high energy X-ray computer tomography (Xray-CT). Based on the measured rod positions data, the subchannel analysis by the NASCA code was performed, in order to investigate applicability of the NASCA code to BT estimation of the rod bowing test section, and influence of displacement from design-based rod position upon BT estimation by the NASCA code. The predicted critical powers are agreement with those obtained by the experiment. The analysis based on the design-based rod positions is also performed, and the result is

Air velocity profiles near sleeve blockages in an unheated 7 x 7 rod bundle. [PWR

Energy Technology Data Exchange (ETDEWEB)

Creer, J. M.; Bates, J. M.

1979-04-01

Local air velocity measurements were obtained with a laser Doppler anemometer near flow blockages in an unheated 7 x 7 rod bundle. Sleeve blockages were positioned on the center nine rods to create an area reduction of 90% in the center four subchannels of the bundle. Experimental results indicated that severe flow disturbances occurred downstream from the blockage cluster but showed only minor flow disturbances upstream from the blockage. Flow reversals were detected downstream from the blockage and persisted for approximately five subchannel hydraulic diameters. The air velocity profiles were in excellent agreement with water velocity data previously obtained at essentially the same Reynolds number. Subchannel average velocity predictions obtained with the COBRA computer program were in good agreement with subchannel average velocities estimated using the measured local velocity data.

Management of radioactive disused lightning rods

Energy Technology Data Exchange (ETDEWEB)

Santos, Paulo de Oliveira; Silva, Fabio, E-mail: pos@cdtn.br, E-mail: silvaf@cdtn.br [Centro de Desenvolvimento da Energia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

2013-07-01

The manufacture of radioactive lightning rod was allowed from 1970 to 1989. This authorization was based on state-of-the art science of that time that verified that radioactive lightning rods had efficiency superior to the conventional lightning rods, denominated Franklin. However, the experience showed that their efficiency was not superior enough to justify the use of radioactive sources. Consequently, in 1989, the National Commission or Nuclear Energy - CNEN, issued the Resolution 04/89 from 04-19-1989, that forbidden the importation of {sup 241}Am tapes, assembling and commercialization of radioactive lightning-rods. The institutes of CNEN are responsible for receiving these lightning-rods and sending to the users procedures for removing and dispatch to the institutes. Therewith, these devices are kept away from the human being and environment. The Nuclear technology Development Center - CDTN and Institute for Energy and Nuclear Research - IPEN of CNEN, has built laboratories appropriate for dismantling such devices and store the {sup 241}Am tapes safely. Nowadays are being researched methodologies to evaluate the contamination levels of the frame for possible recycling and become better the management of these devices. (author)

Management of radioactive disused lightning rods

International Nuclear Information System (INIS)

Santos, Paulo de Oliveira; Silva, Fabio

2013-01-01

The manufacture of radioactive lightning rod was allowed from 1970 to 1989. This authorization was based on state-of-the art science of that time that verified that radioactive lightning rods had efficiency superior to the conventional lightning rods, denominated Franklin. However, the experience showed that their efficiency was not superior enough to justify the use of radioactive sources. Consequently, in 1989, the National Commission or Nuclear Energy - CNEN, issued the Resolution 04/89 from 04-19-1989, that forbidden the importation of 241 Am tapes, assembling and commercialization of radioactive lightning-rods. The institutes of CNEN are responsible for receiving these lightning-rods and sending to the users procedures for removing and dispatch to the institutes. Therewith, these devices are kept away from the human being and environment. The Nuclear technology Development Center - CDTN and Institute for Energy and Nuclear Research - IPEN of CNEN, has built laboratories appropriate for dismantling such devices and store the 241 Am tapes safely. Nowadays are being researched methodologies to evaluate the contamination levels of the frame for possible recycling and become better the management of these devices. (author)

Stabilizing device for control rod tip

International Nuclear Information System (INIS)

Verdone, G.F.

1982-01-01

A control rod has a spring device on its lower end for eliminating oscillatory contact of the rod against its adjacent guide tube wall. The base of the device is connected to the lower tip of the rod. A plurality of elongated extensions are cantilevered downward from the base. Each extension has a shoulder for contacting the guide tube, and the plurality of shoulders as a group has a transverse dimension that is preset to be larger than the inner diameter of the guide tube such that an interference fit is obtained when the control rod is inserted in the tube. The elongated extensions form an open-ended, substantially hollow member through which most of the liquid coolant flows, and the spaces between adjacent extensions allow the flow to bypass the shoulders without experiencing a significant pressure drop

Detection device for control rod scram

International Nuclear Information System (INIS)

Ishiyama, Satoshi.

1989-01-01

The device of the present invention comprises a control rod dropping separately from a control rod driving mechanism main body, a following tube falling separately accompanying therewith and a guide tube for guiding the dropping of the control rod and the following tube. Further, rare earth permanent magnets are embedded with the pole being axially oriented in the following tube and bobbins each mounted with an inner flange made of high magnetic permeability material are disposed to the guide tube. Coils are wound in the bobbin. In this control rod scram detection device, since magnetic fluxes can effectively be supplied to the coils, it is possible to obtain stable and highly reliable scram detection signals. Further, since the coils and the bobbins can be manufactured separately from the guide tube, their assemblies can be tested independently from the guide tube. (K.M.)

Analysis of control rod behavior based on numerical simulation

International Nuclear Information System (INIS)

Ha, D. G.; Park, J. K.; Park, N. G.; Suh, J. M.; Jeon, K. L.

2010-01-01

The main function of a control rod is to control core reactivity change during operation associated with changes in power, coolant temperature, and dissolved boron concentration by the insertion and withdrawal of control rods from the fuel assemblies. In a scram, the control rod assemblies are released from the CRDMs (Control Rod Drive Mechanisms) and, due to gravity, drop rapidly into the fuel assemblies. The control rod insertion time during a scram must be within the time limits established by the overall core safety analysis. To assure the control rod operational functions, the guide thimbles shall not obstruct the insertion and withdrawal of the control rods or cause any damage to the fuel assembly. When fuel assembly bow occurs, it can affect both the operating performance and the core safety. In this study, the drag forces of the control rod are estimated by a numerical simulation to evaluate the guide tube bow effect on control rod withdrawal. The contact condition effects are also considered. A full scale 3D model is developed for the evaluation, and ANSYS - commercial numerical analysis code - is used for this numerical simulation. (authors)

Nuclear fuel rod end plug weld inspection

International Nuclear Information System (INIS)

Parker, M. A.; Patrick, S. S.; Rice, G. F.

1985-01-01

Apparatus and method for testing TIG (tungsten inert gas) welds of end plugs on a sealed nuclear reactor fuel rod. An X-ray fluorescent spectrograph testing unit detects tungsten inclusion weld defects in the top end plug's seal weld. Separate ultrasonic weld inspection system testing units test the top end plug's seal and girth welds and test the bottom end plug's girth weld for penetration, porosity and wall thinning defects. The nuclear fuel rod is automatically moved into and out from each testing unit and is automatically transported between the testing units by rod handling devices. A controller supervises the operation of the testing units and the rod handling devices

Control rod control device

International Nuclear Information System (INIS)

Seiji, Takehiko; Obara, Kohei; Yanagihashi, Kazumi

1998-01-01

The present invention provides a device suitable for switching of electric motors for driving each of control rods in a nuclear reactor. Namely, in a control rod controlling device, a plurality of previously allotted electric motors connected in parallel as groups, and electric motors of any selected group are driven. In this case, a voltage of not driving predetermined selected electric motors is at first applied. In this state an electric current supplied to the circuit of predetermined electric motors is detected. Whether integration or failure of a power source and the circuit of the predetermined electric motors are normal or not is judged by the detected electric current supplied. After they are judged normal, the electric motors are driven by a regular voltage. With such procedures, whether the selected circuit is normal or not can be accurately confirmed previously. Since the electric motors are not driven just at the selected time, the control rods are not operated erroneously. (I.S.)

Correlation of NTD-silicon rod and slice resistivity

International Nuclear Information System (INIS)

Wolverton, W.M.

1984-01-01

Neutron transmutation doped silicon is an electronic material which presents an opportunity to explore a high level of resistivity characterization. This is due to its excellent uniformity of dopant concentration. Appropriate resistivity measurements on the ingot raw material can be used as a predictor of slice resistivity. Correlation of finished NTD rod (i.e. ingot) resistivity to as-cut slice resistivity (after the sawing process) is addressed in the scope of this paper. Empirical data show that the shift of slice-center resistivity compared to rod-end center resistivity is a function of a new kind of rod radial-resistivity gradient. This function has two domains, and most rods are in domain ''A''. Correlating equations show how to significantly improve the prediction of slice resistivity of rods in domain ''A''. The new rod resistivity specifications have resulted in manufacturing economies in the production of NTD silicon slices

Lifting device for drilling rods

Energy Technology Data Exchange (ETDEWEB)

Radzivilovich, L L; Laptev, A G; Lipkovich, V A

1982-01-01

A lifter is proposed for drilling rods including a spacer stand with rotating bracket, boom with by-pass rollers, spacing and lifting hydrocylinders with rods and flexible tie mechanism. In order to improve labor productivity by improving maneuverability and to increase the maintenance zone, the lifter is equipped with a hydrocylinder of advance and a cross piece which is installed with the possibility of forward and rotational movement on the stand, and in which by means of the hydrocylinder of advance a boom is attached. Within the indicated boom there is a branch of the flexible tie mechanism with end attached with the possibility of regulation over the length on a rotating bracket, while the rod of the lifting hydrocylinder is connected to the cross piece.

Rodding Surgery

Science.gov (United States)

... Physical activity prior to surgery, ï‚· Length of the operation; anesthesia issues, ï‚· Reason for the choice of rod, ï‚· Time in the hospital, ï‚· Length of recovery time at home, ï‚· Pain management including control of muscle spasms, ï‚· The rehabilitation plan. ...

BWR control rod drive scram pilot valve monitoring system

International Nuclear Information System (INIS)

Soden, R.A.; Kelly, V.

1984-01-01

The control rod drive system in a Boiling Water Reactor is the most important safety system in the power plant. All components of the system can be verified except the solenoid operated, scram pilot valves without scramming a rod. The pilot valve mechancial works is the weak link to the control rod drive system. These pilot valves control the hydraulic system which applies pressure to the ''insert'' side of the control rod piston and vents the ''withdraw'' side of the piston causing the rods to insert during a scam. The only verification that the valve is operating properly is to scram the rod. The concern for this portion of the system is demonstrated by the high number of redundant components and complete periodic testing of the electrical circuits. The pilot valve can become hung-up through wear, fracture of internal components, mechanical binding, foreign material or chemicals left in the valve during maintenance, etc. If the valve becomes hung-up the electrical tests performed will not indicate this condition and scramming the rod is in jeopardy. Only an attempt to scram a rod will indicate the hung-up valve. While this condition exists the rod is considered inoperative. This paper describes a system developed at a nuclear power plant that monitors the pilot valves on the control rod drive system. This system utilizes pattern recognition to assure proper internal workings of the scram pilot valves to plant operators. The system is totally automatic such that each time the valve is operated on a ''half scram'', a printout is available to the operator along with light indication that each of the 370 valves (on one unit of a BWR) is operating properly. With this monitoring system installed, all components of the control rod drive system including the solenoid pilot valves can be verified as operational without scramming any rods

BWR control rod drive scram pilot valve monitoring program

International Nuclear Information System (INIS)

Soden, R.A.; Kelly, V.

1986-01-01

The control rod drive system in a Boiling Water Reactor is the most important safety system in the power plant. All components of the system can be verified except the solenoid operated, scram pilot valves without scramming a rod. The pilot valve mechanical works is the weak link to the control rod drive system. These pilot valves control the hydraulic system which applies pressure to the insert side of the control rod piston and vents the withdraw side of the piston causing the rods to insert during a scram. The only verification that the valve is operating properly is to scram the rod. The concern for this portion of the system is demonstrated by the high number of redundant components and complete periodic testing of the electrical circuits. The pilot valve can become hung-up through wear, fracture of internal components, mechanical binding, foreign material or chemicals left in the valve during maintenance, etc. If the valve becomes hung-up the electrical tests performed will not indicate this condition and scramming the rod is in jeopardy. Only an attempt to scram a rod will indicate the hung-up valve. While this condition exists the rod is considered inoperative. This paper describes a system developed at a nuclear power plant that monitors the pilot valves on the control rod drive system. This system utilizes pattern recognition to assure proper internal workings of the scram pilot valves to plant operators. The system is totally automatic such that each time the valve is operated on a half scram, a printout is available to the operator along with light indication that each of the 370 valves (on one unit of a BWR) is operating properly. With this monitoring system installed, all components of the control rod drive system including the solenoid pilot valves can be verified as operational without scramming any rods

International symposium on fuel rod simulators: development and application

Energy Technology Data Exchange (ETDEWEB)

McCulloch, R.W. (comp.)

1981-05-01

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

Apparatus for inspecting a irradiated nuclear fuel rod

International Nuclear Information System (INIS)

Saura, Hideaki; Yonemura, Eizo.

1975-01-01

Object: To increase safety and inspection efficiency by operating irradiated fuel rods, which are accommodated in a water-filled pool after being taken out from the reactor. Structure: When making inspection of irradiated fuel rods, particularly the cladding tube thereof, a fuel box which stores irradiated fuel rods in a water pool is secured to a securement mechanism with slime removal apparatus and inspection apparatus on either side capable of being vertically moved, and it is then stopped at a water depth of about 2 meters. When the lid of the box is opened, irradiated fuel rods are taken out with gripping means and then secured together with the gripping means to an operation base provided on the outside of the pool. Thereafter, the box is lowered by operating pedals on the operation base to completely pull out the irradiated fuel rods from the box, and the irradiated fuel rods are then horizontally moved and then held in a suspended state. Next a slime removal apparatus in raised by operating pedals and an inspection element assembly are progressively raised for inspection of the state of the cladding tube of each fuel rod after removal of slime therefrom. (Nakamura, S.)

Slip-spring model of entangled rod-coil block copolymers

Science.gov (United States)

Wang, Muzhou; Likhtman, Alexei E.; Olsen, Bradley D.

2015-03-01

Understanding the dynamics of rod-coil block copolymers is important for optimal design of functional nanostructured materials for organic electronics and biomaterials. Recently, we proposed a reptation theory of entangled rod-coil block copolymers, predicting the relaxation mechanisms of activated reptation and arm retraction that slow rod-coil dynamics relative to coil and rod homopolymers, respectively. In this work, we introduce a coarse-grained slip-spring model of rod-coil block copolymers to further explore these mechanisms. First, parameters of the coarse-grained model are tuned to match previous molecular dynamics simulation results for coils, rods, and block copolymers. For activated reptation, rod-coil copolymers are shown to disfavor configurations where the rod occupies curved portions of the entanglement tube of randomly varying curvature created by the coil ends. The effect of these barriers on diffusion is quantitatively captured by considering one-dimensional motion along an entanglement tube with a rough free energy potential. Finally, we analyze the crossover between the two mechanisms. The resulting dynamics from both mechanisms acting in combination is faster than from each one individually.

Method of driving control rod in reactor

International Nuclear Information System (INIS)

Osa, Hirotaka.

1986-01-01

Purpose: To improve security and safety of the reactor by reducing reactor output automatically and quickly when circulation of cooling water is stopped. Constitution: When the circulating pump is under operation, fluid pressure in the discharge pipe is transferred to the fluid room of fluid pressure cylinder via the control rod drive pipe and lift up the piston, and then the control rod is drawn out of the reactor core. When the circulating pump is lowered in its functions, discharge pipe fluid pressure decreases, fluid pressure in the fluid room decreases, and with less force of piston movement, the control rod gets lowered by its own weight. At this time, the blocked state of the opening by the piston is released, fluid flows into the room. Lowering of pressure and the control rod is promoted by transferring out fluid below the piston in the fluid room to the upper part of the piston via a small gap when the control rod falls by gravity. (Horiuchi, T.)

Hardfacing and welding rods by P/M

International Nuclear Information System (INIS)

Nayar, H.S.

1977-01-01

Certain hardfacing and welding rods are very hard and non-deformable. They are, as it is well known, generally produced by casting processes. Airco has developed a P/M process for producing these rods. The process is already practiced on a semi-production scale. In this process, the powder is poured into suitably designed and prepared molds, vibrated to pack the powder, and sintered at a temperature between the solidus and the liquidus temperatures of the alloy to produce rods with 85% or more of the theoretical density. The P/M process has some distinct advantages over the conventional casting processes. These advantages are highlighted. The process is suitable for producing Fe-, Ni-, Co-, and Cu-base hardfacing and welding rods with and without second phase hard particles such as WC. Microstructures, dimensional and density controls, weld-evaluations and hardness data are included to present evidence that the rods produced by the P/M process are suitable for various welding and hardfacing applications

Disinfection of vegetable seed by treatment with essential oils, organic acids and plant extract

NARCIS (Netherlands)

Wolf, van der J.M.; Birnbaum, Y.E.; Zouwen, van der P.S.; Groot, S.P.C.

2008-01-01

Various essential oils, organic acids, Biosept, (grapefruit extract), Tillecur and extracts of stinging nettle and golden rod were tested for their antimicrobial properties in order to disinfect vegetable seed. In in vitro assays, thyme oil, oregano oil, cinnamon oil, clove oil and Biosept had the

Control rod driving hydraulic device

International Nuclear Information System (INIS)

Sugano, Hiroshi.

1993-01-01

In a control rod driving hydraulic device for an improved BWR type reactor, a bypass pipeline is disposed being branched from a scram pipeline, and a control orifice and a throttle valve are interposed to the bypass pipeline for restricting pressure. Upon occurrence of scram, about 1/2 of water quantity flowing from an accumulator of a hydraulic control unit to the lower surface of a piston of control rod drives by way of a scram pipeline is controlled by the restricting orifice and the throttle valve, by which the water is discharged to a pump suction pipeline or other pipelines by way of the bypass pipeline. With such procedures, a function capable of simultaneously conducting scram for two control rod drives can be attained by one hydraulic control unit. Further, an excessive peak pressure generated by a water hammer phenomenon in the scram pipeline or the control rod drives upon occurrence of scram can be reduced. Deformation and failure due to the excessive peak pressure can be prevented, as well as vibrations and degradation of performance of relevant portions can be prevented. (N.H.)

Synthesis of homochiral tris-indanyl molecular rods

DEFF Research Database (Denmark)

Kjeldsen, Niels Due; Funder, Erik Daa; Gothelf, Kurt Vesterager

2014-01-01

Homochiral tris-indanyl molecular rods designed for supramolecular surface self-assembly were synthesized. The chiral indanol moiety was constructed via a Ti-mediated alkyne trimerization. Further manipulations resulted in a homochiral indanol monomer. This was employed as the precursor for succe...... for successive Sonogashira and Ohira-Bestman reactions towards the homochiral tris-indanyl molecular rods. The molecular rods will be applied for scanning tunnelling microscopy studies of their surface self-assembly and chirality.......Homochiral tris-indanyl molecular rods designed for supramolecular surface self-assembly were synthesized. The chiral indanol moiety was constructed via a Ti-mediated alkyne trimerization. Further manipulations resulted in a homochiral indanol monomer. This was employed as the precursor...

Apparatus for handling control rod drives

International Nuclear Information System (INIS)

Akimoto, A.; Watanabe, M.; Yoshida, T.; Sugaya, Z.; Saito, T.; Ishii, Y.

1979-01-01

An apparatus for handling control rod drives (CRD's) attached by detachable fixing means to housings mounted in a reactor pressure vessel and each coupled to one of control rods inserted in the reactor pressure vessel is described. The apparatus for handling the CRD's comprise cylindrical housing means, uncoupling means mounted in the housing means for uncoupling each of the control rods from the respective CRD, means mounted on the housing means for effecting attaching and detaching of the fixing means, means for supporting the housing means, and means for moving the support means longitudinally of the CRD

The TRPM1 Channel Is Required for Development of the Rod ON Bipolar Cell-AII Amacrine Cell Pathway in the Retinal Circuit.

Science.gov (United States)

Kozuka, Takashi; Chaya, Taro; Tamalu, Fuminobu; Shimada, Mariko; Fujimaki-Aoba, Kayo; Kuwahara, Ryusuke; Watanabe, Shu-Ichi; Furukawa, Takahisa

2017-10-11

Neurotransmission plays an essential role in neural circuit formation in the central nervous system (CNS). Although neurotransmission has been recently clarified as a key modulator of retinal circuit development, the roles of individual synaptic transmissions are not yet fully understood. In the current study, we investigated the role of neurotransmission from photoreceptor cells to ON bipolar cells in development using mutant mouse lines of both sexes in which this transmission is abrogated. We found that deletion of the ON bipolar cation channel TRPM1 results in the abnormal contraction of rod bipolar terminals and a decreased number of their synaptic connections with amacrine cells. In contrast, these histological alterations were not caused by a disruption of total glutamate transmission due to loss of the ON bipolar glutamate receptor mGluR6 or the photoreceptor glutamate transporter VGluT1. In addition, TRPM1 deficiency led to the reduction of total dendritic length, branch numbers, and cell body size in AII amacrine cells. Activated Goα, known to close the TRPM1 channel, interacted with TRPM1 and induced the contraction of rod bipolar terminals. Furthermore, overexpression of Channelrhodopsin-2 partially rescued rod bipolar cell development in the TRPM1 -/- retina, whereas the rescue effect by a constitutively closed form of TRPM1 was lower than that by the native form. Our results suggest that TRPM1 channel opening is essential for rod bipolar pathway establishment in development. SIGNIFICANCE STATEMENT Neurotransmission has been recognized recently as a key modulator of retinal circuit development in the CNS. However, the roles of individual synaptic transmissions are not yet fully understood. In the current study, we focused on neurotransmission between rod photoreceptor cells and rod bipolar cells in the retina. We used genetically modified mouse models which abrogate each step of neurotransmission: presynaptic glutamate release, postsynaptic glutamate

Material operating behaviour of ABB BWR control rods

International Nuclear Information System (INIS)

Rebensdorff, B.; Bart, G.

2000-01-01

The BWR control rods made by ABB use boron carbide (B 4 C and hafnium as absorber material within a cladding of stainless steel. The general behaviour under operation has proven to be very good. ABB and many of their control rod customers have performed extensive inspection programs of control rod behaviour. However, due to changes in the material properties under fast and thermal neutron irradiation defects may occur in the control rods at high neutron fluences. Examinations of irradiated control rod materials have been performed in hot cell laboratories. The examinations have revealed the defect mechanism Irradiation Assisted Stress Corrosion Cracking (IASCC) to appear in the stainless steel cladding. For IASCC to occur three factors have to act simultaneously. Stress, material sensitization and an oxidising environment. Stress may be obtained from boron carbide swelling due to irradiation. Stainless steel may be sensitized to intergranular stress corrosion cracking under irradiation. Normally the reactor environment in a BWR is oxidising. The presentation focuses on findings from hot cell laboratory work on irradiated ABB BWR control rods and studies of irradiated control rod materials in the hot cells at PSI. Apart from physical, mechanical and microstructural examinations, isotope analyses were performed to describe the local isotopic burnup of boron. Consequences (such as possible B 4 C washout) of a under operation in a ABB BWR, after the occurrence of a crack is discussed based on neutron radiographic examinations of control rods operated with cracks. (author)

Prototypical Rod Consolidation Demonstration Project

International Nuclear Information System (INIS)

1993-05-01

The objective of Phase III of the Prototypical Rod Consolidation Demonstration Project (PRCDP) was to procure, fabricate, assemble, and test the Prototypical Rod Consolidation System as described in the NUS Phase II Final Design Report. This effort required providing the materials, components, and fabricated parts which makes up all of the system equipment. In addition, it included the assembly, installation, and setup of this equipment at the Cold Test Facility. During the Phase III effort the system was tested on a component, subsystem, and system level. Volume IV provides the Operating and Maintenance Manual for the Prototypical Rod Consolidation System that was installed at the Cold Test Facility. This document, Book 1 of Volume IV, discusses: Process overview functional descriptions; Control system descriptions; Support system descriptions; Maintenance system descriptions; and Process equipment descriptions

Prototypical Rod Consolidation Demonstration Project

International Nuclear Information System (INIS)

1993-05-01

The objective of Phase III of the Prototypical Rod Consolidation Demonstration Project (PRCDP) was to procure, fabricate, assemble, and test the Prototypical Rod Consolidation System as described in the NUS Phase II Final Design Report. This effort required providing the materials, components, and fabricated parts which makes up all of the system equipment. In addition, it included the assembly, installation, and setup of this equipment at the Cold Test Facility. During the Phase III effort the system was tested on a component, subsystem, and system level. Volume IV provides the Operating and Maintenance Manual for the Prototypical Rod Consolidation System that was installed at the Cold Test Facility. This document, Book 4 of Volume IV, discusses: Off-normal operating and recovery procedures; Emergency response procedures; Troubleshooting procedures; and Preventive maintenance procedures

Calculation of the power distribution in the fuel rods of the low power research reactor using the MCNP4C code

International Nuclear Information System (INIS)

Dawahra, S.; Khattab, K.

2011-01-01

Highlights: → The MCNP4C code was used to calculate the power distribution in 3-D geometry in the MNSR reactor. → The maximum power of the individual rod was found in the fuel ring number 2 and was found to be 105 W. → The minimum power was found in the fuel ring number 9 and was 79.9 W. → The total power in the total fuel rods was 30.9 kW. - Abstract: The Monte Carlo method, using the MCNP4C code, was used in this paper to calculate the power distribution in 3-D geometry in the fuel rods of the Syrian Miniature Neutron Source Reactor (MNSR). To normalize the MCNP4C result to the steady state nominal thermal power, the appropriate scaling factor was defined to calculate the power distribution precisely. The maximum power of the individual rod was found in the fuel ring number 2 and was found to be 105 W. The minimum power was found in the fuel ring number 9 and was 79.9 W. The total power in the total fuel rods was 30.9 kW. This result agrees very well with nominal power reported in the reactor safety analysis report which equals 30 kW. Finally, the peak power factors, which are defined as the ratios between the maximum to the average and the maximum to the minimum powers were calculated to be 1.18 and 1.31 respectively.

Fuel rod failure due to marked diametral expansion and fuel rod collapse occurred in the HBWR power ramp experiment

International Nuclear Information System (INIS)

Yanagisawa, Kazuaki

1985-12-01

In the power ramp experiment with the BWR type light water loop at the HBWR, the two pre-irradiated fuel rods caused an unexpected pellet-cladding interaction (PCI). One occurred in the fuel rod with small gap of 0.10 mm, which was pre-irradiated up to the burn-up of 14 MWd/kgU. At high power, the diameter of the rod was increased markedly without accompanying significant axial elongation. The other occurred in the rod with a large gap of 0.23 mm, which was pre-irradiated up to the burn-up of 8 MWd/kgU. The diameter of the rod collapsed during a diameter measurement at the maximum power level. The causes of those were investigated in the present study by evaluating in-core data obtained from equipped instruments in the experiment. It was revealed from the investigation that these behaviours were attributed to the local reduction of the coolant flow occurred in the region of a transformer in the ramp rig. The fuel cladding material is seemed to become softened due to temperature increase caused by the local reduction of the coolant flow, and collapsed by the coolant pressure, either locally or wholly depending on the rod diametral gap existed. (author)

Analysis of the equalizing holes resistance in fuel assembly spike for lead-based reactor

International Nuclear Information System (INIS)

Zhang, Guangyu; Jin, Ming; Wang, Jianye; Song, Yong

2017-01-01

Highlights: • A RELAP5 model for a 10 MWth lead-based reactor was built to study the hydrodynamic characteristics between the equalizing holes in the fuel assembly spike. • Different fuel assembly total blockage scenarios and different resistances for different fuel assemblies were examined. • The inherent safety characteristics of the lead-based reactor was improved by optimizing the configuration of equalizing holes in the fuel assembly spike. - Abstract: To avoid the damage of the fuel rod cladding when a fuel assembly (FA) is totally blocked, a special configuration of the fuel assembly spike was designed with some equalizing holes in the center region which can let the coolant to flow during the totally blockage scenarios of FA. To study the hydrodynamic characteristics between the equalizing holes and an appropriate resistance, a RELAP5 model was developed for a 10 MWth lead-based reactor which used lead-bismuth as coolant. Several FA total blockage and partial core blockage scenarios were selected. The simulation results indicated that when all the FA spike equalizing holes had the same hydraulic resistance, only a narrow range of suitable equalizing holes resistances could be chosen when a FA was blocked. However, in the two or more FA blockage scenarios, there were no appropriate resistances to meet the requirement. In addition, with different FA spike equalizing holes with different resistances, a large range of suitable equalizing hole resistances could be chosen. Especially a series of suitable resistances were selected when the small power FA resistance was 1/2, 1/4, 1/8 of the large one. Under these circumstances, one, two or three FA blockages would not damage the core. These demonstrated that selecting a series of suitable hydraulic resistances for the equalizing holes could improve the safety characteristics of the reactor effectively.

Experimental studies of the effect of rod spacing on burnout in a simulated rod bundle

International Nuclear Information System (INIS)

Lee, D.H.; Little, R.B.

1962-08-01

Tests on a dumb-bell shaped flow passage simulating the gap between rods in a fuel element indicated that burnout was not significantly affected by inter-rod gap in the range 0.032'' to 0.22''. Test conditions were: 960 p.s.i.a., 2 x 10 6 1b/ft 2 hr mass velocity, and 10% mean exit quality with vertical upflow of water. (author)

Rebirth of a control rod at the Phenix power plant

International Nuclear Information System (INIS)

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

2007-01-01

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

Means for driving control rod

International Nuclear Information System (INIS)

Sato, Haruo; Sasaki, Masayoshi.

1974-01-01

Object: To enable wire rope to be readily removed from guide pulleys for the inspection or replacement of control rods. Structure: A pair of guide pulleys disposed to oppose each other are provided on their periphery with respective notches which are arranged in a staggered fashion. In this way, the rope is made to be removed from the notches for inspection of the control rod or for other purposes. (Kamimura, M.)

Electro-Optical Parameters Of Hairy Rod Polymer/Dimethylformamide System

International Nuclear Information System (INIS)

Spasevska, Hristina

2003-01-01

Rigid rod polymers are materials with special features, that is the reason why they have large scientific and technological applications like isotropic-nematic and other types of transition. One of the biggest problems that happen while investigation to these polymers, at molecular level, is their poor solubility in most of the common solvents. Solubility gets better if while synthesize junctions flexible side chains, [1]. Remaining polymers belong on one new class - hairy rod polymers and have big potential for their technological applications. Because of their nonlinear optical properties and opportunity for self-organization in super molecular structures (films and fibers), can be used in electronics and telecommunications industry either at display-technologies [2]. That is the main reason why controlling the remaining polymer features and connection of their microscopic and macroscopic characteristics, like an complete characterization (in solutions and solid state) is from essential scientific interest, actually it is an imperative. The polymer POD-2/that is subject of investigation, bellows to hairy rod polymers and it is synthesized for the first time [3] by giving a support to their characterization. Measures on diluted solutions from the system POD-2/dimethylformamide are made by the method electric birefringence, at three temperatures (25, 40 and 55 o C). For different concentrations on system, from obtained electro-optical signals (EOS), dependence on electric birefringence (Δn) against applied electric field (E 2 ), is determined. By investigation of electro-optical effects, especially studding the way of their maintenance and losing, while applying electric square impulses, relaxation time τ o of polymer molecules is calculated. From these parameters, as presented data for EOS rise and decay too, deformational electro-optical effect has been determined. (Author)

Self-contact for rods on cylinders

NARCIS (Netherlands)

G.H.M. van der Heijden; M.A. Peletier (Mark); R. Planqué (Robert)

2004-01-01

textabstractWe study self-contact phenomena in elastic rods that are constrained to lie on a cylinder. By choosing a particular set of variables to describe the rod centerline the variational setting is made particularly simple: the strain energy is a second-order functional of a single scalar

Spider and burnable poison rod combinations

International Nuclear Information System (INIS)

Walton, L.A.

1980-01-01

A description is given of an improved design of burnable poison rods and their associated spiders used in the fuel assemblies of pressurized water power reactor cores which allows the rods to be installed and removed more quickly, simply and gently than in previously described systems. (U.K.)

Control rod homogenization in heterogeneous sodium-cooled fast reactors

International Nuclear Information System (INIS)

Andersson, Mikael

2016-01-01

The sodium-cooled fast reactor is one of the candidates for a sustainable nuclear reactor system. In particular, the French ASTRID project employs an axially heterogeneous design, proposed in the so-called CFV (low sodium effect) core, to enhance the inherent safety features of the reactor. This thesis focuses on the accurate modeling of the control rods, through the homogenization method. The control rods in a sodium-cooled fast reactor are used for reactivity compensation during the cycle, power shaping, and to shutdown the reactor. In previous control rod homogenization procedures, only a radial description of the geometry was implemented, hence the axially heterogeneous features of the CFV core could not be taken into account. This thesis investigates the different axial variations the control rod experiences in a CFV core, to determine the impact that these axial environments have on the control rod modeling. The methodology used in this work is based on previous homogenization procedures, the so-called equivalence procedure. The procedure was newly implemented in the PARIS code system in order to be able to use 3D geometries, and thereby be take axial effects into account. The thesis is divided into three parts. The first part investigates the impact of different neutron spectra on the homogeneous control-rod cross sections. The second part investigates the cases where the traditional radial control-rod homogenization procedure is no longer applicable in the CFV core, which was found to be 5-10 cm away from any material interface. In the third part, based on the results from the second part, a 3D model of the control rod is used to calculate homogenized control-rod cross sections. In a full core model, a study is made to investigate the impact these axial effects have on control rod-related core parameters, such as the control rod worth, the capture rates in the control rod, and the power in the adjacent fuel assemblies. All results were compared to a Monte

Control-rod scram device

International Nuclear Information System (INIS)

Matsui, Yoshiro; Saito, Koji.

1986-01-01

Purpose: To eliminate the requirement for the nitrogen gas system in a scram device and enable safety and reliable shutdown of a water-cooled reactor power plant. Constitution: A piston and a spring are contained within a hydraulic vessel, and the piston is driven by the energy stored in the spring so as to supply hydraulic water to control mechanisms. During usual reactor operation, a scram valve is closed and a high water pressure of about 130 kg/cm 2 is applied to the water filled in the vessel through a check valve. Upon occurrence of abnormal conditions and generation of scram signals, the scram valve is opened to supply the water filled in the vessel through the scram valve to the control rod drive mechanisms. When the water pressure in the vessel is decreased, since the piston is urged upwardly by the energy stored in the spring, the water filled in the vessel is intermitently supplied to the control rod drive mechanisms. Thus, control rods can be inserted into the nuclear reactor to shutdown the same. (Horiuchi, T.)

Study on Sumbawa gold ore liberation using rod mill: effect of rod-number and rotational speed on particle size distribution

Science.gov (United States)

Prasetya, A.; Mawadati, A.; Putri, A. M. R.; Petrus, H. T. B. M.

2018-01-01

Comminution is one of crucial steps in gold ore processing used to liberate the valuable minerals from gaunge mineral. This research is done to find the particle size distribution of gold ore after it has been treated through the comminution process in a rod mill with various number of rod and rotational speed that will results in one optimum milling condition. For the initial step, Sumbawa gold ore was crushed and then sieved to pass the 2.5 mesh and retained on the 5 mesh (this condition was taken to mimic real application in artisanal gold mining). Inserting the prepared sample into the rod mill, the observation on effect of rod-number and rotational speed was then conducted by variating the rod number of 7 and 10 while the rotational speed was varied from 60, 85, and 110 rpm. In order to be able to provide estimation on particle distribution of every condition, the comminution kinetic was applied by taking sample at 15, 30, 60, and 120 minutes for size distribution analysis. The change of particle distribution of top and bottom product as time series was then treated using Rosin-Rammler distribution equation. The result shows that the homogenity of particle size and particle size distribution is affected by rod-number and rotational speed. The particle size distribution is more homogeneous by increasing of milling time, regardless of rod-number and rotational speed. Mean size of particles do not change significantly after 60 minutes milling time. Experimental results showed that the optimum condition was achieved at rotational speed of 85 rpm, using rod-number of 7.

Calculation of the power distribution in the fuel rods of the low power research reactor using the MCNP4C code

International Nuclear Information System (INIS)

Dawahra, S.; Khattab, K.

2012-01-01

The Monte Carlo method, using the MCNP4C code, was used in this paper to calculate the power distribution in 3-D geometry in the fuel rods of the Syrian Miniature Neutron Source Reactor (MNSR). To normalize the MCNP4C result to the steady state nominal thermal power, the appropriate scaling factor was defined to calculate the power distribution precisely. The maximum power of the individual rod was found in the fuel ring number 2 and was found to be 105 W. The minimum power was found in the fuel ring number 9 and was 79.9 W. The total power in the total fuel rods was 30.9 k W. This result agrees very well with nominal power reported in the reactor safety analysis report which equals 30 k W. Finally, the peak power factors, which are defined as the ratios between the maximum to the average and the maximum to the minimum powers were calculated to be 1.18 and 1.31 respectively. (author)

Control rod drives for FBR type reactor

International Nuclear Information System (INIS)

Ikakura, Hiroaki.

1990-01-01

The control rod drives for an FBR type reactor of the present invention eliminate obstacles deposited on attracting surfaces between an electromagnet and an armature which connect control rods to recover their retaining power. That is, a sealed chamber capable of controlling its inner pressure by an operation from the outside of a reactor is disposed in an extension pipe, and a nozzle connected to the sealed chamber and facing at the lower end thereof to the attracting surface is disposed. Liquid sodium sucked by evacuating the sealed chamber is jetted out from the nozzle by pressurizing the chamber to simultaneously eliminate obstacles deposited to the attracting surfaces of the electromagnet and the control rod. Alternatively, a nozzle protruding from and retracting to the lower surface of the electromagnet is disposed opposing to each of the attracting surfaces of the electromagnet and the control rod. Similar effect can also be obtained if gases are jetted out in this state. As a result, control rod drives of high reliability for a FBR type reactor can be obtained. (I.S.)

Overview of Japanese control rods development program

International Nuclear Information System (INIS)

Koyama, M.

1984-01-01

The Japanese control rods development program was established based on the fast breeder reactor program. Therefore, PNC's efforts have been made mainly for the development of analysis, design and fabrication technologies for ''JOYO'' and ''MONJU'' control rods. Laboratory studies were performed to obtain the information for absorber materials. The design and fabrication of the sealed and vented type control rod pins were completed, and water loop tests and in-sodium tests were carried out. Irradiation behavior of enriched B 4 C pellets with low and high density in DFR was examined. Japan's experimental fast reactor, JOYO, has been operated at the rated power of 50MWt and 75MWt since April 1977 when the MK-I core (breeder core) attained initial criticality. Post irradiation examinations on control rod, removed from the reactor, were carried out and their performance behavior were evaluated. In the MK-II core, a control rods monitoring program has been in investigation. Absorber Materials Irradiation Rigs (AMIR) are scheduled to be loaded and irradiated in the JOYO MK-II core from 1984. (author)

Protective guide structure for reactor control rod

International Nuclear Information System (INIS)

Ban, Minoru; Umeda, Kenji; Kubo, Noboru; Ito, Tomohiro.

1996-01-01

The present invention provides an improved protective guide structure for control rods, which does not cause swirling of coolants and resonance even though a slit is formed on a protective tube which surrounds a control rod element in a PWR type reactor. Namely, a reactor control rod is constituted with elongated control elements collectively bundled in the form of a cluster. The protective guide structure protectively guides the collected constituent at the upper portion of a reactor container. The protective structure comprises a plurality of protective tubes each having a C-shaped cross section disposed in parallel for receiving control rod elements individually in which the corners of the opening of the cross section of the protective tube are chamfered to an appropriate configuration. With such a constitution, even if coolant flows in a circumferential direction along the protective tubes surrounding the control rod elements, no shearing stream is caused to the coolants flow since the corners of the cross sectional opening (slit) of the tube are chamfered. Accordingly, occurrence of swirlings can be suppressed. (I.S.)

Essential oils as fumigants for bed bugs (Hemiptera: Cimicidae)

Science.gov (United States)

In Petri dish assays, fumigation of a pyrethroid-susceptible strain of bed bugs Cimex lectularius L. (Hemiptera: Cimicidae) with various essential oils resulted in mortality that approached or equaled 100%, after 5 days. However, when bed bugs were exposed to the same essential oils in sealed, comme...

Installing and detaching apparatus for a control rod drive mechanism

International Nuclear Information System (INIS)

Akimoto, Seiichi; Watanabe, Mitsuhiro; Yoshida, Tomiharu; Sugaya, Jun-ichi; Saito, Takashi.

1976-01-01

Object: To facilitate maintenance and repair of a control rod drive mechanism. Structure: The apparatus comprises a means moving in a moving direction of a control rod within a reactor vessel, said moving means having a housing mounted thereon, a means mounted on the reactor vessel to release a connection between a control rod drive mechanism connected to the control rod and the control rod, and a means for mounting and removing a fixing means which connects the reactor vessel to the control rod drive means. With this arrangement, cooling water of high radioactivity level may not be leaked outside to thereby notably reduce dangerousness of exposure and materially cut time required for mounting and removing the control rod drive mechanism. (Ohara, T.)

Self-contact for rods on cylinders

NARCIS (Netherlands)

Heijden, van der G.H.M.; Peletier, M.A.; Planqué, R.

2006-01-01

We study self-contact phenomena in elastic rods that are constrained to lie on a cylinder. By choosing a particular set of variables to describe the rod centerline the variational setting is made particularly simple: the strain energy is a second-order functional of a single scalar variable, and the

Self-contact for rods on cylinders

NARCIS (Netherlands)

Heijden, van der G.H.M.; Peletier, M.A.; Planqué, R.

2004-01-01

We study self-contact phenomena in elastic rods that are constrained to lie on a cylinder. By choosing a particular set of variables to describe the rod centerline the variational setting is made particularly simple: the strain energy is a second-order functional of a single scalar variable, and the

Effects of different rod spacers (helical types) on coolant crossmixing

International Nuclear Information System (INIS)

Zhukov, A.V.; Sviridenko, E.Ya.; Matyukhin, N.M.; Rymkevich, K.S.; Ushakov, P.A.

1981-11-01

The results of investigations (electromagnetic measuring method) on coolant cross mixing in rod clusters with spiral wire spacers with different winding directions, with alternating unfinned and finned rods (case 'fin to rod'), as well as in rod clusters with much space between the rods, (case 'fin to fin') are reported. The local fluid dynamics parameters (distribution of the transversal and longitudinal velocity component) that define the physical processes of the coolant exchange in the rod clusters with helical spacers are explained. The investigation results for different helical spacer types are compared with each other. (orig.) [de

Ultrasonics aids the identification of failed fuel rods

International Nuclear Information System (INIS)

Anon.

1985-01-01

Over a number of years Brown Boveri Reaktor of West Germany has developed and commercialized an ultrasonic failed fuel rod detection system. Sipping has up to now been the standard technique for failed fuel detection, but sipping can only indicate whether or not an assembly contains defective rods; the BBR system can tell which rod is defective. (author)

Method for installing a control rod driving device in a reactor

International Nuclear Information System (INIS)

Sato, Haruo; Watanabe, Masatoshi.

1975-01-01

Object: To install a device using a wire rope, including individually moving up and down a control rod and a control rod driving device thereby enabling to install them within a low house and to reduce time required for installing operation. Structure: The control rod is temporarily attached to a support structure for the control rod driving device, the control rod driving device is suspended on a crane positioned upwardly of the support structure, a rope connected to the control rod driving device is connected to the control rod, a sagged portion of the rope is then wound about a rotary cylinder, the control rod is disconnected from its temporary attachment, and the wound rope is wound back while the rotary cylinder is rotated to move down the control rod. After the rope has been released from the rotary cylinder, the control rod driving device is moved down by the crane. (Kamimura, M.)

Influence of implant rod curvature on sagittal correction of scoliosis deformity.

Science.gov (United States)

Salmingo, Remel Alingalan; Tadano, Shigeru; Abe, Yuichiro; Ito, Manabu

2014-08-01

Deformation of in vivo-implanted rods could alter the scoliosis sagittal correction. To our knowledge, no previous authors have investigated the influence of implanted-rod deformation on the sagittal deformity correction during scoliosis surgery. To analyze the changes of the implant rod's angle of curvature during surgery and establish its influence on sagittal correction of scoliosis deformity. A retrospective analysis of the preoperative and postoperative implant rod geometry and angle of curvature was conducted. Twenty adolescent idiopathic scoliosis patients underwent surgery. Average age at the time of operation was 14 years. The preoperative and postoperative implant rod angle of curvature expressed in degrees was obtained for each patient. Two implant rods were attached to the concave and convex side of the spinal deformity. The preoperative implant rod geometry was measured before surgical implantation. The postoperative implant rod geometry after surgery was measured by computed tomography. The implant rod angle of curvature at the sagittal plane was obtained from the implant rod geometry. The angle of curvature between the implant rod extreme ends was measured before implantation and after surgery. The sagittal curvature between the corresponding spinal levels of healthy adolescents obtained by previous studies was compared with the implant rod angle of curvature to evaluate the sagittal curve correction. The difference between the postoperative implant rod angle of curvature and normal spine sagittal curvature of the corresponding instrumented level was used to evaluate over or under correction of the sagittal deformity. The implant rods at the concave side of deformity of all patients were significantly deformed after surgery. The average degree of rod deformation Δθ at the concave and convex sides was 15.8° and 1.6°, respectively. The average preoperative and postoperative implant rod angle of curvature at the concave side was 33.6° and 17.8Â

Flow in rod bundles

International Nuclear Information System (INIS)

Hazi, G.; Mayer, G.

2005-01-01

For power upgrading VVER-440 reactors we need to know exactly how the temperature measured by the thermocouples is related to the average outlet temperature of the fuel assemblies. Accordingly, detailed knowledge on mixing process in the rod bundles and in the fuel assembly head have great importance. Here we study the hydrodynamics of rod bundles based on the results of direct numerical and large eddy simulation of flows in subchannels. It is shown that secondary flow and flow pulsation phenomena can be observed using both methodologies. Some consequences of these observations are briefly discussed. (author)

Impact loading of a BWR control rod during braking

International Nuclear Information System (INIS)

Heeschen, U.

1977-01-01

In an emergency case the control rods of a boiling water reactor are shot into the RPV from below against the weight of the rods with drive motors. According to the position of the control rods between the fuel elements the rods can reach in that case velocities up to 4 m/s. The moved masses of the control rods and of the pistons (both of them are connected by a coupling) are braked through a cup spring which transfers its forces to the RPV-bottom sphere. The spring has to be designed that in this case tthe complete kinetic energy of he control rods of about 1000Nm can be taken up. The spring power and the inertia of the moved masses cause extremely high loadings during and shortly after the impact onto the spring. The shock-like loading propagates along the whole rod at the speed of sound, and this is also the reason why the weaker cross-sections have to endure considerable short-term stress peaks. (Auth.)

Rapid and accurate control rod calibration measurement and analysis

International Nuclear Information System (INIS)

Nelson, George W.; Doane, Harry J.

1990-01-01

In order to reduce the time needed to perform control rod calibrations and improve the accuracy of the results, a technique for a measurement, analysis, and tabulation of integral rod worths has been developed. A single series of critical rod positions are determined at constant low power to reduce the waiting time between positive period measurements and still assure true stable reactor period data. Reactivity values from positive period measurements and control rod drop measurements are used as input data for a non-linear fit to the expected control rod integral worth shape. With this method, two control rods can be calibrated in about two hours, and integral and differential calibration tables for operator use are printed almost immediately. Listings of the BASIC computer programs for the non-linear fitting and calibration table preparation are provided. (author)

Drilling Experiments of Dummy Fuel Rods Using a Mock-up Drilling Device and Detail Design of Device for Drilling of Irradiated Nuclear Fuel Rods

Energy Technology Data Exchange (ETDEWEB)

Kim, Jae Yong; Lee, H. K.; Chun, Y. B.; Park, S. J.; Kim, B. G

2007-07-15

KAERI are developing the safety evaluation method and the analysis technology for high burn-up nuclear fuel rod that is the project, re-irradiation for re-instrumented fuel rod. That project includes insertion of a thermocouple in the center hole of PWR nuclear fuel rod with standard burn-up, 3,500{approx}4,000MWD/tU and then inspection of the nuclear fuel rod's heat performance during re-irradiation. To re-fabricate fuel rod, two devices are needed such as a drilling machine and a welding machine. The drilling machine performs grinding a center hole, 2.5 mm in diameter and 50 mm in depth, for inserting a thermocouple. And the welding machine is used to fasten a end plug on a fuel rod. Because these two equipment handle irradiated fuel rods, they are operated in hot cell blocked radioactive rays. Before inserting any device into hot cell, many tests with that machine have to be conducted. This report shows preliminary experiments for drilling a center hole on dummy of fuel rods and optimized drilling parameters to lessen operation time and damage of diamond dills. And the design method of a drilling machine for irradiated nuclear fuel rods and detail design drawings are attached.

Verification test of control rod system for HTR-10

International Nuclear Information System (INIS)

Zhou Huizhong; Diao Xingzhong; Huang Zhiyong; Cao Li; Yang Nianzu

2002-01-01

There are 10 sets of control rods and driving devices in 10 MW High Temperature Gas-cooled Test Reactor (HTR-10). The control rod system is the controlling and shutdown system of HTR-10, which is designed for reactor criticality, operation, and shutdown. In order to guarantee technical feasibility, a series of verification tests were performed, including room temperature test, thermal test, test after control rod system installed in HTR-10, and test of control rod system before HTR-10 first criticality. All the tests data showed that driving devices working well, control rods running smoothly up and down, random position settling well, and exactly position indicating

Development of a large scale structure in the rod gap region for turbulent in-line flow through closely spaced rod arrays

International Nuclear Information System (INIS)

Hooper, J.D.

1984-01-01

Experimental studies of developed axial single-phase flow through closely spaced rod arrays have shown, with reducing p/d ratio, the development of high axial and azimuthal turbulence intensities in the rod gap region. Associated with this is the existence of very high levels of the azimuthal Reynolds shear stress component either side of the rod gap centre. Spatial correlation analysis of the three turbulent velocity components has shown a large scale coherent and almost periodic structure in the rod gap region. The structure is markedly different to the currently accepted secondary flow model. 14 references

Low fluid level in pulse rod shock absorber

Energy Technology Data Exchange (ETDEWEB)

Aderhold, H. C.

1974-07-01

On various occasions during pulse mode operation the shim and regulating control rods would drop when the pulse rod was withdrawn. Subsequent investigation traced the problem to the pulse rod shock absorber which was found to be low in hydraulic fluid. The results of the investigation, the corrective action taken, and a method for measuring the shock absorber fluid level are presented. (author)

Summary of Skoda JS rod drop measurements analysis

International Nuclear Information System (INIS)

Svarny, J.; Krysl, V.

1999-01-01

A summary is presented of the Skoda JS rod drop reactivity measurements analysis provided during last two years based on control rod worth measurements by the outer ion chambers. Standard analysis based on comparisons of dynamics macrocode MOBY-DICK-SK and experimental data is extended to the 8-th group delayed neutron structure and new features of rod drop process are investigated. (author)

Low fluid level in pulse rod shock absorber

International Nuclear Information System (INIS)

Aderhold, H.C.

1974-01-01

On various occasions during pulse mode operation the shim and regulating control rods would drop when the pulse rod was withdrawn. Subsequent investigation traced the problem to the pulse rod shock absorber which was found to be low in hydraulic fluid. The results of the investigation, the corrective action taken, and a method for measuring the shock absorber fluid level are presented. (author)

Experiment studies of fuel rod vibration in coolant flow for substantiation of vibration stability of fuel rods with no fretting-wear

International Nuclear Information System (INIS)

Egorov, Yu. V.; Afanasiev, A. V.; Makarov, V. V.; Matvienko, I. V.

2013-01-01

For substantiation of vibration stability it is necessary to determine the ultimate permissible vibration levels which do not cause fretting, to compare them with the level of fuel rod vibration caused by coolant flow. Another approach is feasible if there is experience of successful operation of FA-prototypes. In this case in order to justify vibration stability it may be sufficient to demonstrate that the new element does not cause increased vibration of the fuel rod. It can be done by comparing the levels of hydro-dynamic fuel rod vibration and FA new designs. Program of vibration tests of TVS-2M model included studies of forced oscillations of 12 fuel rods in the coolant flow in the spans containing intensifiers, in the reference span without intensifiers, in the lower spans with assembled ADF and after its disassembly. The experimental results for TVS-2M show that in the spans with intensifier «Sector run» the level of movements is 6% higher on the average than in the span without intensifiers, in the spans with intensifier «Eddy» it is 2% higher. The level of fuel rod vibration movements in the spans with set ADF is 2 % higher on the average than without ADF. During the studies of TVS-KVADRAT fuel rod vibration, the following tasks were solved: determination of acceleration of the middle of fuel rod spans at vibration excited due to hydrodynamics; determination of influence of coolant thermal- hydraulic parameters (temperature, flowrate, dynamic pressure) on fuel rod vibration response; determination of influence of span lengths on the vibration level. Conclusions: 1) The vibration tests of the full-scale model of TVS-2M in the coolant flow showed that the new elements of TVS-2M design (intensifiers of heat exchange and ADF) are not the source of fuel rod increased vibration. Considering successful operation of similar fuel rod spans in the existing TVS-2M design, vibration stability of TVS-2M fuel rods with new elements is ensured on the mechanism of

On the Wave Stresses in the Rods of Anvil Hammers

Directory of Open Access Journals (Sweden)

V. M. Sinitskiy

2014-01-01

Full Text Available With operating anvil hammers, there are rigid impacts of die tools, and as a result, almost instantaneous impact stops of the falling parts of hammer. Such operating conditions lead to the accelerated breakdowns of rods because of significant wave stresses arising in them. Common differential and integral methods to estimate wave stresses are widespread in engineering practice. However, to use them a researcher has to possess certain skills and special software. We consider the method for estimating the wave stresses in the rods of anvil hammers based on Laplace transforms (LT of wave equation. The article shows a procedure to set up and solve differential wave equations by operator method. These equations describe the wave propagation process of strains and stresses in the rods of anvil hammers with rigid impact and taking into account a damping rod connection with the head of hammer. The method takes into consideration an influence of both piston and rod weights and of mechanical and geometrical characteristics of rod on the stress value in the placement of rod in hammer head. Results analysis shows that a sufficiently efficient method for practical improving the durability of rods is the method of damping impact load on the rod through setting the damping devices in the form either of elastic "pad" of one or another design or of hydraulic shock absorbers in the placement of its connection with the hammer head. In this case there is a change of the wave front, it becomes flatter. It is shown that the stresses in the rod are proportional to the amount of wave stresses because of the own impact of rod and piston, which make a total weight of the system. Effect of piston weight on the stresses value at the rod during impact is directly proportional to the ratio of its weight to the rod weight. The geometric parameters of rod and the speed of the falling parts before the impact also influence on the value of stresses in the rod.The represented

Tipping Time of a Quantum Rod

Science.gov (United States)

Parrikar, Onkar

2010-01-01

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

Temperature actuated automatic safety rod release

International Nuclear Information System (INIS)

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

1987-01-01

This patent describes a nuclear reactor having a core, a safety rod for downward insertion into and upward withdrawal from the core, a drive shaft for supporting and operating the safety rod, and drive means connected to the drive shaft for operating the shaft. An apparatus is described for releasably supporting the safety rod, the apparatus comprising an upper adapter adapted to be affixed to the upper end of the safety rod, the upper adapter having a retention means, a lower portion on the drive shaft and having a hollow interior for housing the upper adapter, a bimetallic means supported within the hollow interior of the lower portion and having at least one ledge which engages the retention means to support the upper adapter, the bimetallic means being a substantially cylindrical bimetallic member for receiving the upper adapter in a generally coaxial relation, the substantially cylindrical bimetallic member comprising an inner layer and an outer layer, and the inner layer having a greater coefficient of thermal expansion than the outer layer

Automatic operation device for control rods

International Nuclear Information System (INIS)

Sekimizu, Koichi

1984-01-01

Purpose: To enable automatic operation of control rods based on the reactor operation planning, and particularly, to decrease the operator's load upon start up and shutdown of the reactor. Constitution: Operation plannings, demand for the automatic operation, break point setting value, power and reactor core flow rate change, demand for operation interrupt, demand for restart, demand for forecasting and the like are inputted to an input device, and an overall judging device performs a long-term forecast as far as the break point by a long-term forecasting device based on the operation plannings. The automatic reactor operation or the like is carried out based on the long-term forecasting and the short time forecasting is performed by the change in the reactor core status due to the control rod operation sequence based on the control rod pattern and the operation planning. Then, it is judged if the operation for the intended control rod is possible or not based on the result of the short time forecasting. (Aizawa, K.)

Chitin Fiber and Chitosan 3D Composite Rods

International Nuclear Information System (INIS)

Wang, Z.; Hu, Q.; Cai, L.

2010-01-01

Chitin fiber (CHF) and chitosan (CS) 3D composite rods with layer-by-layer structure were constructed by in situ precipitation method. CHF could not be dissolved in acetic acid aqueous solution, but CS could be dissolved due to the different deacetylation degree (D.D) between CHF and CS. CHF with undulate surfaces could be observed using SEM to demonstrate that the sufficiently rough surfaces and edges of the fiber could enhance the mechanical combining stress between fiber and matrix. XRD indicated that the crystallinity of CHF/CS composites decreased and CS crystal plane d-spacing of CHF/CS composites became larger than that of pure CS rod. TG analysis showed that mixing a little amount of CHF could enhance thermal stability of CS rod, but when the content of CHF was higher than the optimum amount, its thermal stability decreased. When 0.5% CHF was added into CS matrix, the bending strength and bending modulus of the composite rods arrived at 114.2 MPa and 5.2 GPa, respectively, increased by 23.6% and 26.8% compared with pure CS rods, indicating that CHF/CS composite rods could be a better candidate for bone fracture internal fixation.

Commissioning of a passive rod scanner at INB

Energy Technology Data Exchange (ETDEWEB)

Junqueira, Fabio da Silva; Oliveira, Carlos A.; Palheiros, Franklin, E-mail: carlossilva@inb.gov.br, E-mail: franklin@inb.gov.br [Industrias Nucleares do Brasil (INB), Resende, RJ (Brazil). Superintendencia de Engenharia do Combustivel; Fernandez, Pablo Jesus Piñer, E-mail: pineiro@tecnatom.es [Tecnatom, San Sebastian de los Reyes, Madrid (Spain)

2015-07-01

For the 21st reload for Angra 1, a shift from Standard to Advanced fuel design will be introduced, where the fuel assemblies under the new design will contain fuel rods with axial blanket, in line with ELETRONUCLEAR's requirement for a higher energy efficient reactor fuel. Additionally, fuel rods for Angra 2 and 3, using gadolinium type burnable poison, have to be submitted to inspections due to the demand for the same type of inspection, which cannot be certified at INB currently. In keeping with CNEN regulations, every fuel-assembly component must be inspected and certified by a qualified method. Nevertheless, INB lacks the means to perform the certification-required inspection aimed at determining the uranium enrichment and presence of gadolinium pellets inside the closed rods. Hence, the use is necessary of a scanner capable of inspecting differently enriched fuel rods and/or gadolinium pellets (axial blanket). This work aims to present the recent Passive Rod Scanner installed at INB with most advance technology in the area, making possible to completely fulfill Angra 1, 2 and 3 rods inspection at INB Resende site. (author)

Influence of miscut on crystal truncation rod scattering

International Nuclear Information System (INIS)

Munkholm, A.; Brennan, S.

1999-01-01

X-rays can be used to measure the roughness of a surface by the study of crystal truncation rod scattering. It is shown that for a simple cubic lattice the presence of a miscut surface with a regular step array has no effect on the scattered intensity of a single rod and that a distribution of terrace widths on the surface is shown to have the same effect as adding roughness to the surface. For a perfect crystal without miscut, the scattered intensity is the sum of the intensity from all the rods with the same in-plane momentum transfer. For all real crystals, the scattered intensity is better described as that from a single rod. It is shown that data-collection strategies must correctly account for the sample miscut or there is a potential for improperly measuring the rod intensity. This can result in an asymmetry in the rod intensity above and below the Bragg peak, which can be misinterpreted as being due to a relaxation of the surface. The calculations presented here are compared with data for silicon (001) wafers with 0.1 and 4 miscuts. (orig.)

Burnable poison rod for a nuclear reactor fuel assembly

International Nuclear Information System (INIS)

Funk, C.E.; Oneufer, A.S.

1984-01-01

A burnable poison rod for use in a nuclear reactor fuel assembly which includes concentrically disposed rods having an annular space therebetween which extends the full length of the rods. The inner rod is hollow to permit circulation of coolant therethrough. Annular burnable poison pellets are positioned in the annular space which is closed at both ends by plugs. A spring clip is located in the plenum space above the pellet stack in the rods. The spring clip is of cylindrical configuration having a gap in the material which provides two ends adapted to be squeezed toward each other. A cross section of the clip shows that its ends contain alternating flat and round edges, the round edges conforming to the outer rod inner surface to provide a retentive force which is releasably applied to the pellet stack as it grows during operation in a reactor

Substitute safety rods: Physics design and NTG calibration

International Nuclear Information System (INIS)

Baumann, N.P.

1991-07-01

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

Seismic analysis of hydraulic control rod driving system

International Nuclear Information System (INIS)

Zheng, Yanhua; Bo, Hanliang; Dong, Duo

2002-01-01

A simplified mathematical model was developed for the Hydraulic Control Rod Driving System (HCRDS) of a 200 MW nuclear heating reactor, which incorporated the design of its chamfer-hole step cylinder, to analyze its seismic response characteristics. The control rod motion was analyzed for different sine-wave vibration loadings on platform vibrator. The vibration frequency domain and the minimum acceleration amplitude of the control rod needed to cause the control rod to step to its next setting were compared with the design acceleration amplitude spectrum. The system design was found to be safety within the calculated limits. The safety margin increased with increasing frequency. (author)

Prototypical Rod Consolidation Demonstration Project

International Nuclear Information System (INIS)

1993-05-01

The objective of Phase 3 of the Prototypical Rod consolidation Demonstration Project (PRCDP) was to procure, fabricate, assemble, and test the Prototypical Rod consolidation System as described in the NUS Phase 2 Final Design Report. This effort required providing the materials, components, and fabricated parts which makes up all of the system equipment. In addition, it included the assembly, installation, and setup of this equipment at the Cold Test Facility. During the Phase 3 effort the system was tested on a component, subsystem, and system level. This volume 1, discusses the PRCDP Phase 3 Test Program that was conducted by the HALLIBURTON NUS Environmental Corporation under contract AC07-86ID12651 with the United States Department of Energy. This document, Volume 1, Book 2 discusses the following topics: Fuel Rod Extraction System Test Results and Analysis Reports and Clamping Table Test Results and Analysis Reports

Control rod drives for HTGR type reactor

International Nuclear Information System (INIS)

Nishiguchi, Isoharu; Katagiri, Shigeo.

1991-01-01

The device of the present invention has a feature of having stable braking characteristics upon scram operation of control rods. That is, control rod drives are moved upon and down by a dram which rotates the control rod suspended from to a wire rope, and the dram is disconnected from the driving mechanism by a crutch mechanism upon scram, to rapidly insert the control rod in the reactor by its own weight. An electric generator is used as a braking mechanism for controlling the scram speed of the control rod. A plurality of resistors disposed outside of the reactor coolants boundary are connected in parallel between input/output terminals of the electric generator. With such a constitution, braking characteristics are determined by the intensity of the permanent magnet, number of the coil windings and values of the resistors constituting the power generator. Accordingly, the braking characteristics are less changed relative to the working circumstantial conditions, the history of use and the state of mounting. As a result, stable braking characteristics can always be obtained. Further, braking characteristics can easily be controlled by varying the resistance value. (I.S.)

Clad buffer rod sensors for liquid metals

International Nuclear Information System (INIS)

Jen, C.-K.; Ihara, I.

1999-01-01

Clad buffer rods, consisting of a core and a cladding, have been developed for ultrasonic monitoring of liquid metal processing. The cores of these rods are made of low ultrasonic-loss materials and the claddings are fabricated by thermal spray techniques. The clad geometry ensures proper ultrasonic guidance. The lengths of these rods ranges from tens of centimeters to 1m. On-line ultrasonic level measurements in liquid metals such as magnesium at 700 deg C and aluminum at 960 deg C are presented to demonstrate their operation at high temperature and their high ultrasonic performance. A spherical concave lens is machined at the rod end for improving the spatial resolution. High quality ultrasonic images have been obtained in the liquid zinc at 600 deg C. High spatial resolution is needed for the detection of inclusions in liquid metals during processing. We also show that the elastic properties such as density, longitudinal and shear wave velocities of liquid metals can be measured using a transducer which generates and receives both longitudinal and shear waves and is mounted at the end of a clad buffer rod. (author)

Thermal hydraulic performance of naturally aspirated control rod housing assemblies

International Nuclear Information System (INIS)

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

1992-01-01

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

Intrasacral rod fixation for pediatric lumbopelvic fusion.

Science.gov (United States)

Ilharreborde, Brice; Mazda, Keyvan

2014-07-01

This paper reports the authors' 19 years experience with pediatric intrasacral rod fixation. After insertion of two cannulated screws in S1 with and an original template guiding them into the anterior third of the endplate, two short fusion rods were inserted into the sacrum according to Jackson's technique distally to S3. In neuromuscular scoliosis, pelvic obliquity was reduced by connecting the proximal and distal constructs, distraction or compression, and in situ rod bending. In children with high-grade spondylolisthesis, lumbosacral kyphosis was reduced by rotation of the sacrum and in situ bending. There were no direct neurological or vascular injuries. The main complication was infection (7%). No pseudarthrosis or significant loss of correction at the lumbosacral junction was observed during follow-up. Intrasacral rod fixation appears to be safe and reliable for lumbopelvic fusion in pediatric patients.

Depletion calculations of adjuster rods in Darlington

Energy Technology Data Exchange (ETDEWEB)

Arsenault, B.; Tsang, K., E-mail: benoit.arsenault@amecfw.com, E-mail: kwok.tsang@amecfw.com [AMEC Foster Wheeler, Toronto, ON (Canada)

2015-07-01

This paper describes the simulation methodology and reactivity worth calculated for aged adjuster rods in the Darlington core. ORIGEN-S IST was applied to simulate the isotope transmutation process of the stainless steel and titanium adjusters. The compositions were used in DRAGON-IST to calculate the change in incremental properties of aged adjusters. Pre-simulations of the reactivity worth of the stainless steel and titanium adjusters in Darlington were performed using RFSP-IST and the results showed that the titanium adjuster rods exhibit faster reactivity-worth drop than that of stainless steel rods. (author)

Cost targets for at-reactor spent fuel rod consolidation

International Nuclear Information System (INIS)

Macnabb, W.V.

1985-01-01

The high-level nuclear waste management system in the US currently envisions the disposal of spent fuel rods that have been removed from their assemblies and reconfigured into closely packed arrays. The process of fuel rod removal and packaging, referred to as rod consolidation, can occur either at reactors or at an integrated packaging facility, monitored retrievable storage (MRS). Rod consolidation at reactors results in cost savings down stream of reactors by reducing needs for additional storage, reducing the number of shipments, and reducing (eliminating, in the extreme) the amount of fuel handling and consolidation at the MRS. These savings accrue to the nuclear waste fund. Although private industry is expected to pay for at-reactor activities, including rod consolidation, it is of interest to estimate cost savings to the waste system if all fuel were consolidated at reactors. If there are savings, the US Department of Energy (DOE) may find it advantageous to pay for at-reactor rod consolidation from the nuclear waste fund. This paper assesses and compares the costs of rod consolidation at reactors and at the MRS in order to determine at what levels the former could be cost competitive with the latter

Evaluation of rod insertion issue for NPP Krsko

International Nuclear Information System (INIS)

Gunstek, A.; Kurincic, B.

1998-01-01

The last couple of years incident with control rods sticking in lower part of the fuel assemblies have been reported of several reactor operators and fuel vendors throughout of the world. Several activities were initiated immediately to determine the root cause of incomplete rod insertion. The purpose of this activities were to collect plants trip history data and testing results, review of available worldwide experience, review of plant operation and fuel management, detailed review of manufacturing and material property and to maintain detailed mechanical model. In this paper, we will present activities in Nuclear Power Plant Krsko which have been performed after NRC initiated the Root Cause Process (NRC Bulletin 96-01). NPP Krsko has not experienced rod insertion anomaly yet but anyway the additional tests were carried out. Rod drop time measurements that were performed normally at beginning of cycle at nominal temperature and pressure (HSB mode) have been extended also to end of cycle. Rod drop time, velocity of dropped rods and magnitudes of the initial recoil bounces vs. burnup were also analyzed. Also RCCA drag test with upper internals in place and drive shafts attached to RCCAs has been performed since then. At last two outages (1997 and 1998) drag test were carried out with digital scale meter to gather additional information. In addition to that, the reload core design has been performed with new constrains on rodded fuel assembly burnup as proposed by the industry.(author)

Correlated and uncorrelated invisible temporal white noise alters mesopic rod signaling.

Science.gov (United States)

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

2016-03-01

We determined how rod signaling at mesopic light levels is altered by extrinsic temporal white noise that is correlated or uncorrelated with the activity of one (magnocellular, parvocellular, or koniocellular) postreceptoral pathway. Rod and cone photoreceptor excitations were independently controlled using a four-primary photostimulator. Psychometric (Weibull) functions were measured for incremental rod pulses (50 to 250 ms) in the presence (or absence; control) of perceptually invisible subthreshold extrinsic noise. Uncorrelated (rod) noise facilitates rod detection. Correlated postreceptoral pathway noise produces differential changes in rod detection thresholds and decreases the slope of the psychometric functions. We demonstrate that invisible extrinsic noise changes rod-signaling characteristics within the three retinogeniculate pathways at mesopic illumination depending on the temporal profile of the rod stimulus and the extrinsic noise type.

Fuel rods

International Nuclear Information System (INIS)

Adachi, Hajime; Ueda, Makoto

1985-01-01

Purpose: To provide a structure capable of measuring, in a non-destructive manner, the releasing amount of nuclear gaseous fission products from spent fuels easily and at a high accuracy. Constitution: In order to confirm the integrity and the design feasibility of a nuclear fuel rod, it is important to accurately determine the amount of gaseous nuclear fission products released from nuclear pellets. In a structure where a plurality of fuel pellets are charged in a fuel cladding tube and retained by an inconel spring, a hollow and no-sealed type spacer tube made of zirconium or the alloy thereof, for example, not containing iron, cobalt, nickel or manganese is formed between the spring and the upper end plug. In the fuel rod of such a structure, by disposing a gamma ray collimator and a gamma ray detector on the extension of the spacer pipe, the gamma rays from the gaseous nuclear fission products accumulated in the spacer pipe can be detected while avoiding the interference with the induction radioactivity from inconel. (Kamimura, M.)

Analysis of addition of the safety rods at RSG-GAS core

International Nuclear Information System (INIS)

S, Tukiran; S, Tagor Malem; K, Iman

2002-01-01

The silicide fuel loading of the RSG-GAS core is planned to increase from 250 gU to 300 gU. Increasing of fuel loading will prolong the operation cycle length from 25 days to 32,5 days, but ability of reactivity compensation by control rods system decreased because the reactivity shut-down margin is available only 1,03 %, expectation is 2.2 %. One of solutions is added two safety control rods in B-3 and G-10 positions the aim of installing two safety rods (BKP) in RSG-GAS core is to increase core safety margin. So before using the safety control rods in the RSG-GAS core, it is necessary to know its performance, one of the tests showing its performance is to measure the reactivity of the safety control rods. Measurement of safety control rods were done to know each reactivity worth of safety control rods at middle cycle so that the safety rod be used in the RSG-GAS core. Measurement done by using calibration control rods with couple compensation method which always using in the RSG-GAS core to measure the existing control rods. The results of measurement showed that two safety rods (BKP01 and BKP02) have reactivity worth of 93.5 cent and 87.5 cent, respectively. the total reactivity worth of safety control rods is 1.38%. So the two safety rods can be used to increase safety margin of the RSG-GAS core if the fuel is exchanged to 300 gU of loading

Relation of fuel rod service parameters and design requirements to produced fuel rod and their components

International Nuclear Information System (INIS)

Bibilashvili, Yu.K.

1999-01-01

Based on the presented material it is possible to state that there is a very close link between the fuel operational parameters and the requirements for its design and production process. The required performance and life-time of a fuel rod can be only assured by the correctly selected design and process solutions. The economical aspect of this problem is significantly depend on the commercial feasibility of a particular selected solution with the provision of an automated and comparative by inexpensive production of a fuel rod and its components. The operational conditions are also important for the life time of the fuel rods. If there are no special measures for the mitigation of the certain operation conditions the leakage of fuel elements can take place before the planned time. (authors)

Physics calculations for the RIA 1-3 irradiated rod test

International Nuclear Information System (INIS)

Young, T.E.

1981-06-01

The RIA 1-3 test would employ a square array of four pre-irradiated BWR rods to provide information on fuel failure modes and consequences of postulated Reactivity Initiated Accidents in power reactors. Calculations were done to: (1) predict R-O power distributions in the test rods for thermal-hydraulic and fuel-failure analysis; and (2) predict the steady-state and transient ratios of test fuel energy deposition to core energy deposition (Figures of Merit). Fission distributions for the test were computed with the RAFFL Monte Carlo code using an external neutron current source from a complete-reactor radial calculation with the SCAMP S/sub n/ code. Energies per fission for the rods were computed using the SINBAD buildup and depletion code, the GAMSOR gamma ray source code, and the QAD-BSA point-kernel shielding code. The calculated rod average-to-test average energy deposition ratios are 0.99, 0.99, and 0.97 for the rods irradiated to approximately 12 CWd/tu, and 1.04 for the rod irradiated to 4.8 GWd/tu. The maximum deviation of the power density of 1/12-rod azimuthal segments from the rod average is 4%. For an estimated control rod position of 0.591 m withdrawn the predicted radial average energy deposition at the axial peak in an average test rod is 1.71 (kW/m)/MW during preconditioning, and 1.84 (kJ/kg UO 2 ) MW.S during the burst. 16 figures, 7 tables

Rod internal pressure quantification and distribution analysis using Frapcon

Energy Technology Data Exchange (ETDEWEB)

Jessee, Matthew Anderson [ORNL; Wieselquist, William A [ORNL; Ivanov, Kostadin [Pennsylvania State University, University Park

2015-09-01

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

Segmented fuel and moderator rod

International Nuclear Information System (INIS)

Doshi, P.K.

1987-01-01

This patent describes a continuous segmented fuel and moderator rod for use with a water cooled and moderated nuclear fuel assembly. The rod comprises: a lower fuel region containing a column of nuclear fuel; a moderator region, disposed axially above the fuel region. The moderator region has means for admitting and passing the water moderator therethrough for moderating an upper portion of the nuclear fuel assembly. The moderator region is separated from the fuel region by a water tight separator

Measurement and analysis of CEFR safety and shim rod worth

International Nuclear Information System (INIS)

Chen Yiyu; Yang Yong; Gang Zhi; Xu Li; Yang Xiaoyan; Zhou Keyuan; Hu Dingsheng

2013-01-01

The reactivity worth of safety rods and shim rods in critical phase and operating phase was calculated respectively using Monte Carlo program in this paper. In addition, the reactivity worth of safety rods and shim rods was measured by the rod drop-off method and period method. The experimental results are in good agreement with the calculated values with less than 5% error. It illustrates the high calculation precision of Monte Carlo program, which provides a practical reference for subsequent application of Monte Carlo program in future demonstration fast reactors. (authors)

Study on flow-induced vibration of the fuel rod in HTTR

International Nuclear Information System (INIS)

Takase, Kazuyuki

1988-03-01

This study was performed in order to investigate flow-induced vibration characteristics of a fuel rod in HTTR (High Temperature engineering Test Reactor) from both an experiment and a numerical simulation. Two kinds of fuel rods were used in this experiment: one was a graphite rod which simulated a specification of the HTTR's fuel rod and the other was an aluminum rod whose weight was a half of the graphite one. The experiment was carried out up to Re = 31000 using air at room temperature and pressure. Air flowed downstream in an annular passage which consisted of the fuel rod and the graphite channel. Numerical simulations by fluid and frequency equations were also carried out. Numerical and experimental results were then compared. The following conclusions were drived: (1) The fuel rod amplitudes increase with the flow rate and with a decrease of the fuel rod weight. (2) The fuel rod amplitudes are obtained by δ/De = 2.22 x 10 -10 Re 1.43 , 9000 ≤ Re ≤ 31000, where δ is a vibration amplitude, De is a hydraulic diameter and Reis Reynolds number. (3) The fuel rod frequencies shift from lower natural frequency to higher as the flow rate increases. (4) The flow-induced vibration behavior of the fuel rod can simulate well by simultaneous equations which used the turbulence model for fluid and the mass model for vibration of the fuel rod. (author)

Review of control rod calibration methods for irradiated AGRs

Energy Technology Data Exchange (ETDEWEB)

Telford, A. R.R.

1975-10-15

Methods of calibrating control rods with particular reference to irradiated CAGR are surveyed. Some systematic spatial effects are found and an estimate of their magnitude made. It is concluded that control rod oscillation provides a promising method of calibrating rods at power which is as yet untried on CAGR. Also the rod drop using inverse kinetics provides a rod calibration but spatial effects may be large and these would be difficult to correct theoretically. The pulsed neutron technique provides a calibration route with small errors due to spatial effects provided a suitable K-tube can be developed. The xenon transient method is shown to have spatial effects which have not needed consideration in earlier reactors but which in CAGR would need very careful evaluation.

Pressurized water reactor fuel rod design methodology

International Nuclear Information System (INIS)

Silva, A.T.; Esteves, A.M.

1988-08-01

The fuel performance program FRAPCON-1 and the structural finite element program SAP-IV are applied in a pressurized water reactor fuel rod design methodology. The applied calculation procedure allows to dimension the fuel rod components and characterize its internal pressure. (author) [pt

LOFT advanced fuel rod instrumentation development

International Nuclear Information System (INIS)

Billeter, T.R.; Brown, R.L.; Chan, A.I.Y.; Day, C.K.; Meyers, S.C.; Sheen, E.M.; Stringer, J.L.

1978-01-01

Advanced fuel rod instrumentation for the Loss of Fluid Test (LOFT) reactor is being developed by the Hanford Engineering Development Laboratory for the Nuclear Regulatory Commission. This effort calls for development of sensors to measure fuel rod axial motion, fuel centerline temperature (to 2200 0 C), fuel rod plenum gas pressure (to 2500 psig), and plenum gas temperature (to 1500 0 F). A parallel test and evaluation of several modified commercial sensors was undertaken and will result in commercial availability of the final qualified sensors. Necessary test facilities were prepared for the development and evaluation effort. Tests to date indicate a three coil Linear Variable Differential Transformer (LVDT), operated from temperature compensating signal source and processing electronics, will meet the desired requirements

Modelling of pellet-cladding interaction for PWRs reactors fuel rods

International Nuclear Information System (INIS)

Esteves, A.M.

1991-01-01

The pellet-cladding interaction that can occur in a PWR fuel rod design is modelled with the computer codes FRAPCON-1 and ANSYS. The fuel performance code FRAPCON-1 analyzes the fuel rod irradiation behavior and generates the initial conditions for the localized fuel rod thermal and mechanical modelling in two and three-dimensional finite elements with ANSYS. In the mechanical modelling, a pellet fragment is placed in the fuel rod gap. Two types of fuel rod cladding materials are considered: Zircaloy and austenitic stainless steel. Linear and non-linear material behaviors are allowed. Elastic, plastic and creep behaviors are considered for the cladding materials. The modelling is applied to Angra-II fuel rod design. The results are analyzed and compared. (author)

Fuel rod puncturing and fission gas monitoring system examination techniques

International Nuclear Information System (INIS)

Song, Woong Sup

1999-02-01

Fission gas products accumulated in irradiated fuel rod is 1-2 cm 3 in CANDU and 40-50 cm 3 in PWR fuel rod. Fuel rod puncturing and fission gas monitoring system can be used for both CANDU and PWR fuel rod. This system comprises puncturing device located at in cell part and monitoring device located at out cell part. The system has computerized 9 modes and can calculate both void volume and mass volume only single puncturing. This report describes techniques and procedure for operating fuel rod puncturing and gas monitoring system which can be play an important role in successful operation of the devices. Results obtained from the analysis can give more influence over design for fuel rods. (Author). 6 refs., 9 figs

Control rod position detector for nuclear reactor

International Nuclear Information System (INIS)

Kudo, Mitsuru; Fujiwara, Hiroshi.

1981-01-01

Purpose: To improve the reliability of a control rod position detector by detecting a reactive code with a combination of control rod position change signals produced from vertical and horizontal axis decoders, generation an error signal and thus simultaneously detecting the operation of more than two lead switches. Constitution: Horizontal and vertical axis position signals responsive to changes in the control rod position are applied from lead switches connected in a predetermined matrix connection corresponding to the notches of the positions of respective position detecting probes, the reactive output from the decoder is detected by a reactive code detecting circuit, which in turn generates a fault signal, and the control rod position code converted in a notch number generating circuit is converted to a predetermined value indicating invalidity. Accordingly, a fault caused by the simultaneous operation of a plurality of failed lead switches can be effectively detected. (Yoshino, Y.)

Rod bundle burnout data and correlation comparisons

International Nuclear Information System (INIS)

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

1985-01-01

Rod bundle burnout data from 30 steady-state and 3 transient tests were obtained from experiments performed in the Thermal Hydraulic Test Facility at the Oak Ridge National Laboratory. The tests covered a parameter range relevant to intact core reactor accidents ranging from large break to small break loss-ofcoolant conditions. Instrumentation within the 64-rod test section indicated that burnout occurred over an axial range within the bundle. The distance from the point where the first dry rod was detected to the point where all rods were dry was up to 60 cm in some of the tests. The burnout data should prove useful in developing new correlations for use in reactor thermalhydraulic codes. Evaluation of several existing critical heat flux correlations using the data show that three correlations, the Barnett, Bowring, and Katto correlations, perform similarly and correlate the data better than the Biasi correlation

Analytical prediction of turbulent friction factor for a rod bundle

International Nuclear Information System (INIS)

Bae, Jun Ho; Park, Joo Hwan

2011-01-01

An analytical calculation has been performed to predict the turbulent friction factor in a rod bundle. For each subchannel constituting a rod bundle, the geometry parameters are analytically derived by integrating the law of the wall over each subchannel with the consideration of a local shear stress distribution. The correlation equations for a local shear stress distribution are supplied from a numerical simulation for each subchannel. The explicit effect of a subchannel shape on the geometry parameter and the friction factor is reported. The friction factor of a corner subchannel converges to a constant value, while the friction factor of a central subchannel steadily increases with a rod distance ratio. The analysis for a rod bundle shows that the friction factor of a rod bundle is largely affected by the characteristics of each subchannel constituting a rod bundle. The present analytic calculations well predict the experimental results from the literature with rod bundles in circular, hexagonal, and square channels.

Chitin Fiber and Chitosan 3D Composite Rods

Directory of Open Access Journals (Sweden)

Zhengke Wang

2010-01-01

Full Text Available Chitin fiber (CHF and chitosan (CS 3D composite rods with layer-by-layer structure were constructed by in situ precipitation method. CHF could not be dissolved in acetic acid aqueous solution, but CS could be dissolved due to the different deacetylation degree (D.D between CHF and CS. CHF with undulate surfaces could be observed using SEM to demonstrate that the sufficiently rough surfaces and edges of the fiber could enhance the mechanical combining stress between fiber and matrix. XRD indicated that the crystallinity of CHF/CS composites decreased and CS crystal plane d-spacing of CHF/CS composites became larger than that of pure CS rod. TG analysis showed that mixing a little amount of CHF could enhance thermal stability of CS rod, but when the content of CHF was higher than the optimum amount, its thermal stability decreased. When 0.5% CHF was added into CS matrix, the bending strength and bending modulus of the composite rods arrived at 114.2 MPa and 5.2 GPa, respectively, increased by 23.6% and 26.8% compared with pure CS rods, indicating that CHF/CS composite rods could be a better candidate for bone fracture internal fixation.

Ion selectivity of the cation transport system of isolated intact cattle rod outer segments: evidence for a direct communication between the rod plasma membrane and the rod disk membranes.

Science.gov (United States)

Schnetkamp, P P

1980-05-08

The ion selectivity of cation transport through the plasma membrane of isolated intact cattle rod outer segments (rods) is investigated by means of 45Ca-exchange experiments and light-scattering experiments. These techniques appear to provide complementary information: the 45Ca experiments (45Ca fluxes in rods) describe electroneutral antiport, whereas the light-scattering experiments (shrinkage and swelling of rods upon hypertonic shocks with various electrolytes) reveal electrogenic uniport. Electroneutral symport of ions (salt transport) does not take place without addition of external ionophores and application of salts of weak acids. 1. Intact rods recover from a hypertonic shock in the presence of FCCP when lithium, sodium and potassium acetate are applied, but not when ammonium chloride, calcium and magnesium acetate are used. This indicates that the plasma membrane of isolated intact cattle rods is relatively permeable to net transport of Na+, Li+ and K+, and relatively impermeable to net transport of Cl-, Mg2+ and Ca2+ under conditions that do not give rise to diffusion potentials. 2. Rapid (t1/2 exchange diffusion of internal 45Ca with external Na+, Ca2+, Sr2+ and Ba2+, respectively. 3. All tested cations lower the rate of 45Ca uptake. The latter can be described by a single rate constant indicating a homogeneous rod preparation and a homogeneous endogenous Ca2+ pool. However, only those cations which stimulate 45Ca efflux from preloaded rods lower the final equilibrium of 45Ca uptake. Except for the effects of K+, Rb+ and Cs+ the reduction of the rate of 45Ca uptake by external cations appears to arise from competition for a common site on the plasms membrane. The observed affinities for this site do not correlate with actual transport (as indicated by the ability to stimulate 45Ca efflux). 4. K+ increases the affinity of the exchange diffusion system to Ca2+ from 1 microM to 0.15 microM and changes the relative affinities with respect to Ca2+ for the

The development and validation of control rod calculation methods

International Nuclear Information System (INIS)

Rowlands, J.L.; Sweet, D.W.; Franklin, B.M.

1979-01-01

Fission rate distributions have been measured in the zero power critical facility, ZEBRA, for a series of eight different arrays of boron carbide control rods. Diffusion theory calculations have been compared with these measurements. The normalised fission rates differ by up to about 30% in some regions, between the different arrays, and these differences are well predicted by the calculations. A development has been made to a method used to produce homogenised cross sections for lattice regions containing control rods. Calculations show that the method also reproduces the reaction rate within the rod and the fission rate dip at the surface of the rod in satisfactory agreement with the more accurate calculations which represent the fine structure of the rod. A comparison between diffusion theory and transport theory calculations of control rod reactivity worths in the CDFR shows that for the standard design method the finite mesh approximation and the difference between diffusion theory and transport theory (the transport correction) tend to cancel and result in corrections to be applied to the standard mesh diffusion theory calculations of about +- 2% or less. This result applies for mesh centred finite difference diffusion theory codes and for the arrays of natural boron carbide control rods for which the calculations were made. Improvements have also been made to the effective diffusion coefficients used in diffusion theory calculations for control rod followers and these give satisfactory agreement with transport theory calculations. (U.K.)

Simulation of fuel rod irradiation capsules in water loops by electric heater rods

International Nuclear Information System (INIS)

Lopez, J.; Montes, M.; Serrano, J.; Haefner, H.E.

1984-01-01

The out of pile simulation of irradiation devices was carried out by J.E.N. in the frame of the KfK-JEN joint experiment for irradiation of fast reactor fuel rods (IVO-FR2-Vg7). A typical single-wall-Nak (22% Na, 78% K) electrical heated capsule was fabricated and hydraulical tests were done. The capsule was instrumented with 10 thermocouples in order to obtain the radial temperature profile into the capsule in function of the electrical rod power (max. 215 w/cm), flow rate (max. 2,4 m 3 /h) and coolant temperature (max. 60degC). The experimental values are compared to the Tecap-Code results. (author)

Control rod studies in small and medium sized fast reactors

International Nuclear Information System (INIS)

John, T.M.; Mohanakrishnan, P.; Mahalakshmi, B.; Singh, R.S.

1988-01-01

Control rods are the primary safety mechanism in the operation of fast reactors. Neutronic parameters associated with the control rods have to be evaluated precisely for studying the behaviour of the reactor under various operating conditions. Control rods are strong neutron absorbers discretely distributed in the reactor core. Accurate estimation of control rod parameters demand, in principle transport theory solutions in exact geometry. But computer codes for such evaluations usually consume exorbitantly large computer time and memory for even a single parameter evaluation. During the design of reactors, evaluation of these parameters will be required for many configurations of control rods. In this paper, the method used at Indira Gandhi Centre for Atomic Research for estimating the parameters associated with control rods is presented. Diffusion theory solutions were used for computations. A scheme using three dimensional geometry represented by triangular meshes and diffusion theory solutions in few energy groups for control rod parameter evaluation is presented. This scheme was employed in estimating the control rod parameters in a 500 Mw(e) fast reactor. Error due to group collapsing is estimated by comparing with 25 group calculations in three dimensions for typical cases. (author). 5 refs, 4 figs, 3 tabs

Adequacy of the analysis of mock-up control rod experiment with FCA

International Nuclear Information System (INIS)

Mizoo, Nobutatsu; Nakano, Masafumi

1977-07-01

A method of numerical analysis has been investigated for the mock-up control rod experiment of FCA VII-1 assembly constructed as the engineering mock-up of prototype fast breeder reactor MONJU. The results of criticality and B 4 C mock-up control rod worths analysis for the assembly are described in comparison with the experimental ones. The tendency of the C/E value with 10 B enrichment and the interaction effect of the multiple rods array was also examined. Reactivities and the mock-up rods worths were obtained with the X-Y geometry six groups diffusion theory. Twelve kinds of the mock-up rods with different 10 B contents and/or enrichments were used in the experiment; effective cross-sections are provided for each rod by calculation using the collision probability method. Criticality of VII-1 90Z assembly is underestimated for 3 reference critical configurations, ranging from -0.65%Δk/k to -0.77%Δk/k. The C/E values at core center for 12 kinds of B 4 C mock-up rods range from 1.03 to 1.09. The overestimate of the rod worth increases with macroscopic absorption cross-section of the rod region. The C/E values for 24 different arrays of the mock-up rods ranging from single rod to five rods lie between 1.04 and 1.08. The C/E value tends to decrease with increase in the number of rods inserted, the values for five rods arrays being about 4% lower than those for single rod arrays. The calculated interaction effects of the multiple rods arrays are slightly more negative than the experimental ones. (auth.)

SEFLEX - fuel rod simulator effects in flooding experiments. Pt. 2

International Nuclear Information System (INIS)

Ihle, P.; Rust, K.

1986-03-01

This report presents typical data and a limited heat transfer analysis from unblocked bundle reflood tests of an experimental thermal-hydraulic program. Full-length bundles of 5 x 5 fuel rod simulators having a gas-filled gap between the Zy cladding and the alumina pellets were tested in the test rig designed for the earlier Flooding Experiments with Blocked Arrays (FEBA-program). The 5 x 5 FEBA rod bundle tests were performed with gapless heater rods. These rods have a close thermal contact between the stainless steel cladding and the electric insulation material. A comparison of the SEFLEX data with the reference data of FEBA obtained under identical initial and reflood conditions shows the influence of different fuel rod simulators on the thermal-hydraulic behavior during forced feed bottom reflooding of unblocked and blocked arrays. Compared to bundles of gapless rods, bundles of rods with Zy claddings and a gas filled gap between claddings and pellets, which more closely represent the features that exist in an actual fuel rod geometry, produced higher quench front velocities, enhanced removal of stored heat in the rods, reduced peak cladding temperatures, increased grid spacer effects and absolutely unproblematic coolability of 90 percent blockages with bypass. The data offer the opportunity for further validation of computer codes to make realistic predictions of safety margins during a LOCA in a PWR. (orig./HP) [de

SEFLEX fuel rod simulator effects in flooding experiments. Pt. 3

International Nuclear Information System (INIS)

Ihle, P.; Rust, K.

1986-03-01

This report presents typical data and a limited heat transfer analysis from blocked bundle reflood tests of an experimental thermal-hydraulic program. Full-length bundles of 5x5 fuel rod simulators having a gas-filled gap between the Zy cladding and the alumina pellets were tested in the test rig designed for the earlier Flooding Experiments with Blocked Arrays (FEBA-program). The 5x5 FEBA rod bundle tests were performed with gapless heater rods. These rods have a close thermal contact between the stainless steel cladding and the electric insulation material. A comparison of the SEFLEX data with the reference data of FEBA obtained under identical initial and reflood conditions shows the influence of different fuel rod simulators on the thermal-hydraulic behavior during forced feed bottom reflooding of unblocked and blocked arrays. Compared to bundles of gapless rods, bundles of rods with Zy claddings and a gas filled gap between claddings and pellets, which more closely represent the features that exist in an actual fuel rod geometry, produced higher quench front velocities, enhanced removal of stored heat in the rods, reduced peak cladding temperatures, increased grid spacer effects and absolutely unproblematic coolability of 90 percent blockages with bypass. The data offer the opportunity for further validation of computer codes to make realistic predictions of safety margins during a LOCA in a PWR. (orig./HP) [de

Fuel rod pressure in nuclear power reactors: Statistical evaluation of the fuel rod internal pressure in LWRs with application to lift-off probability

Energy Technology Data Exchange (ETDEWEB)

Jelinek, Tomas

2001-02-01

In this thesis, a methodology for quantifying the risk of exceeding the Lift-off limit in nuclear light water power reactors is outlined. Due to fission gas release, the pressure in the gap between the fuel pellets and the cladding increases with burnup of the fuel. An increase in the fuel-clad gap due to clad creep would be expected to result in positive feedback, in the form of higher fuel temperatures, leading to more fission gas release, higher rod pressure, etc, until the cladding breaks. An increase in the fuel-clad gap that leads to this positive feedback is a phenomenon called Lift-off and is a limitation that must be considered in the fuel core management. Lift-off is a consequence of very high internal fuel rod pressure. The internal fuel rod pressure is therefore used as a Lift-off indicator. The internal fuel rod pressure is closely connected to the fission gas release into the fuel rod plenum and is thus used to increase the database. It is concluded that the dominating error source in the prediction of the pressure in Boiling Water Reactors (BWR), is the power history. There is a bias in the fuel pressure prediction that is dependent on the fuel rod position in the fuel assembly for BWRs. A methodology to quantify the risk of the fuel rod internal pressure exceeding a certain limit is developed; the risk is dependent of the pressure prediction and the fuel rod position. The methodology is based on statistical treatment of the discrepancies between predicted and measured fuel rod internal pressures. Finally, a methodology to estimate the Lift-off probability of the whole core is outlined.

A novel active equalization method for lithium-ion batteries in electric vehicles

International Nuclear Information System (INIS)

Wang, Yujie; Zhang, Chenbin; Chen, Zonghai; Xie, Jing; Zhang, Xu

2015-01-01

Highlights: • Build an active equalization method for lithium-ion batteries. • A bidirectional transformer topology is introduced for active equalization. • The PF method is used for cell SOC estimation to eliminate drift noise of current. • The SOC based equalization algorithm is analyzed with different SOC bounds. - Abstract: Cell inconsistency is inevitable due to manufacturing constraint. Therefore, cell equalization is essentially required. In this paper, we propose a novel active equalization method based on the remaining capacity of cells which is feasible for lithium-ion battery packs in electric vehicles (EVs). The cell models are established based on a combined electrochemical model of lithium-ion batteries. The remaining capacity and state-of-charge (SOC) of cells are observed at the beginning of equalization. The particle filter (PF) method is employed to estimate the cell SOCs during equalization in order to eliminate the drift noise of the current sensor. The first high-SOC cell discharge (FHCD) and first low-SOC cell charge (FLCC) equalization algorithms are proposed and compared with 1% and 3% SOC bounds, respectively. The validation experiment results have shown that the proposed algorithm is suitable for equalization of lithium-ion batteries in EVs

Process and device for exchanging neutron absorber rods

International Nuclear Information System (INIS)

Baero, G.; Kraus, W.; Stindt, W.

1987-01-01

The control element repair device contains lifting equipment for inserting the control element in the accommodation device. Due to the case position assigned to each absorber rod of a control element, after removing the carrier with the absorber rods fixed to it, the defective rods can be replaced by new ones. The accommodation device has a support to support the carrier. Turning the control element for the PWR through 180 0 is prevented. (DG) [de

Lumped-parameter fuel rod model for rapid thermal transients

International Nuclear Information System (INIS)

Perkins, K.R.; Ramshaw, J.D.

1975-07-01

The thermal behavior of fuel rods during simulated accident conditions is extremely sensitive to the heat transfer coefficient which is, in turn, very sensitive to the cladding surface temperature and the fluid conditions. The development of a semianalytical, lumped-parameter fuel rod model which is intended to provide accurate calculations, in a minimum amount of computer time, of the thermal response of fuel rods during a simulated loss-of-coolant accident is described. The results show good agreement with calculations from a comprehensive fuel-rod code (FRAP-T) currently in use at Aerojet Nuclear Company

Theoretical and computational studies of entangled rod-coil block copolymer diffusion

Science.gov (United States)

Wang, Muzhou; Alexander-Katz, Alfredo; Olsen, B. D.

2012-02-01

Despite continued interest in the thermodynamics of rod-coil block copolymers for functional nanostructured materials in organic electronics and biomaterials, relatively few studies have investigated the dynamics of these systems which are important for understanding diffusion, mechanics, and self-assembly kinetics. Here, the diffusion of coil-rod-coil block copolymers through entangled melts is simulated using the Kremer-Grest molecular dynamics model, demonstrating that the mismatch between the curvature of the rod and coil blocks results in dramatically slower reptation through the entanglement tube. For rod lengths near the tube diameter, this hindered diffusion is explained by a local curvature-dependent free energy penalty produced by the curvature mismatch, resulting in a rough energy surface as the rod moves along the tube contour. Compared to coil homopolymers which reptate freely along the tube, rod-coil block copolymers undergo an activated diffusion process which is considerably slower as the rod length increases. For large rods, diffusion of the rod through the tube only occurs when the coil blocks occupy straight entanglement tubes, which requires ``arm retraction'' as the dominant relaxation mechanism.

Gamma-ray spectroscopy on irradiated fuel rods

International Nuclear Information System (INIS)

Terremoto, Luis Antonio Albiac

2009-01-01

The recording of gamma-ray spectra along an irradiated fuel rod allows the fission products to be qualitatively and quantitatively examined. Among all nondestructive examinations performed on irradiated fuel rods by gamma-ray spectroscopy, the most comprehensive one is the average burnup measurement, which is quantitative. Moreover, burnup measurements by means of gamma-ray spectroscopy are less time-consuming and waste-generating than burnup measurements by radiochemical, destructive methods. This work presents the theoretical foundations and experimental techniques necessary to measure, using nondestructive gamma-ray spectroscopy, the average burnup of irradiated fuel rods in a laboratory equipped with hot cells. (author)

Oligo(naphthylene–ethynylene) Molecular Rods

DEFF Research Database (Denmark)

Cramer, Jacob Roland; Ning, Yanxiao; Shen, Cai

2013-01-01

of palladium-catalyzed Sonogashira reactions between naphthyl halides and acetylenes. The triazene functionality was used as a protected iodine precursor to allow linear extension of the molecular rods during the synthe-ses. The carboxylic acid groups in the target molecules were protected as esters during......Molecular rods designed for surface chirality studies have been synthesized in high yields. The molecules are composed of oligo(naphthylene–ethynylene) skeletons and functionalized at their two termini with carboxylic acids and hydrophobic groups. The molecular skeletons were constructed by means...

Investigation of axial power gradients near a control rod tip

Energy Technology Data Exchange (ETDEWEB)

Loberg, John, E-mail: John.Loberg@fysast.uu.se [Uppsala University, Department of Physics and Astronomy, Division of Applied Nuclear Physics, Box 525, SE-75120 Uppsala (Sweden); Osterlund, Michael, E-mail: Michael.Osterlund@fysast.uu.se [Uppsala University, Department of Physics and Astronomy, Division of Applied Nuclear Physics, Box 525, SE-75120 Uppsala (Sweden); Bejmer, Klaes-Hakan, E-mail: Klaes-Hakan.Bejmer@vattenfall.com [Vattenfall Nuclear Fuel AB, Jaemtlandsgatan 99, 162 60 Vaellingby, Stockholm (Sweden); Blomgren, Jan, E-mail: Jan.Blomgren@vattenfall.com [Vattenfall Nuclear Fuel AB, Jaemtlandsgatan 99, 162 60 Vaellingby, Stockholm (Sweden); Kierkegaard, Jesper, E-mail: Jesper.Kierkegaar@vattenfall.com [Vattenfall Nuclear Fuel AB, Jaemtlandsgatan 99, 162 60 Vaellingby, Stockholm (Sweden)

2011-07-15

Highlights: > Pin power gradients near BWR control rod tips have been investigated. > A control rod tip is modeled in MCNP and compared to simplified 2D/3D geometry. > Small nodes increases pin power gradients; standard nodes underestimates gradients. > The MCNP results are validated against axial gamma scan of a controlled fuel pin. - Abstract: Control rod withdrawal in BWRs induces large power steps in the adjacent fuel assemblies. This paper investigates how well a 2D/3D method, e.g., CASMO5/SIMULATE5 computes axial pin power gradients adjacent to an asymmetrical control-rod tip in a BWR. The ability to predict pin power gradients accurately is important for safety considerations whereas large powers steps induced by control rod withdrawal can cause Pellet Cladding Interaction. The computation of axial pin power gradients axially around a control rod tip is a challenging task for any nodal code. On top of that, asymmetrical control rod handles are present in some BWR designs. The lattice code CASMO requires diagonal symmetry of all control rod parts. This introduces an error in computed pin power gradients that has been evaluated by Monte Carlo calculations. The results show that CASMO5/SIMULATE5, despite the asymmetrical control rod handle, is able to predict the axial pin power gradient within 1%/cm for axial nodal sizes of 15-3.68 cm. However, a nodal size of 3.68 cm still causes underestimations of pin power gradients compared with 1 cm nodes. Furthermore, if conventional node sizes are used, {approx}15 cm, pin power gradients can be underestimated by over 50% compared with 1 cm nodes. The detailed axial pin power profiles from MCNP are corroborated by measured gamma scan data on fuel rods irradiated adjacent to control rods.

Investigation of axial power gradients near a control rod tip

International Nuclear Information System (INIS)

Loberg, John; Osterlund, Michael; Bejmer, Klaes-Hakan; Blomgren, Jan; Kierkegaard, Jesper

2011-01-01

Highlights: → Pin power gradients near BWR control rod tips have been investigated. → A control rod tip is modeled in MCNP and compared to simplified 2D/3D geometry. → Small nodes increases pin power gradients; standard nodes underestimates gradients. → The MCNP results are validated against axial gamma scan of a controlled fuel pin. - Abstract: Control rod withdrawal in BWRs induces large power steps in the adjacent fuel assemblies. This paper investigates how well a 2D/3D method, e.g., CASMO5/SIMULATE5 computes axial pin power gradients adjacent to an asymmetrical control-rod tip in a BWR. The ability to predict pin power gradients accurately is important for safety considerations whereas large powers steps induced by control rod withdrawal can cause Pellet Cladding Interaction. The computation of axial pin power gradients axially around a control rod tip is a challenging task for any nodal code. On top of that, asymmetrical control rod handles are present in some BWR designs. The lattice code CASMO requires diagonal symmetry of all control rod parts. This introduces an error in computed pin power gradients that has been evaluated by Monte Carlo calculations. The results show that CASMO5/SIMULATE5, despite the asymmetrical control rod handle, is able to predict the axial pin power gradient within 1%/cm for axial nodal sizes of 15-3.68 cm. However, a nodal size of 3.68 cm still causes underestimations of pin power gradients compared with 1 cm nodes. Furthermore, if conventional node sizes are used, ∼15 cm, pin power gradients can be underestimated by over 50% compared with 1 cm nodes. The detailed axial pin power profiles from MCNP are corroborated by measured gamma scan data on fuel rods irradiated adjacent to control rods.

Vibration characteristics of a long flexible rod supported with multiple gaps

International Nuclear Information System (INIS)

Umeda, Kenji; Ban, Minoru; Ito, Tomohiro; Nakamura, Tomoichi; Fujita, Katuhisa.

1991-01-01

Control rods are long flexible rods supported with multiple gaps and forced to vibrate by hydraulic forces of reactor coolant flow. In order to find methods, to extend control rod life time, flow-induced vibration and wear mechanism of control rod should be identified. As a basic approach for this objective a vibration test in air using a single control rod and nonlinear vibration analyses were conducted to study characteristic of vibration and wear at support points of the control rod. Several test and analytical cases were performed with several initial support conditions, exciting points and exciting force level. With these test results, some information on the vibration and wear mechanism of control rods that explain wear features in actual plants was obtained. (author)

Long-term effects of retinopathy of prematurity (ROP) on rod and rod-driven function.

Science.gov (United States)

Harris, Maureen E; Moskowitz, Anne; Fulton, Anne B; Hansen, Ronald M

2011-02-01

The purpose of this study was to determine whether recovery of scotopic sensitivity occurs in human ROP, as it does in the rat models of ROP. Following a cross-sectional design, scotopic electroretinographic (ERG) responses to full-field stimuli were recorded from 85 subjects with a history of preterm birth. In 39 of these subjects, dark adapted visual threshold was also measured. Subjects were tested post-term as infants (median age 2.5 months) or at older ages (median age 10.5 years) and stratified by severity of ROP: severe, mild, or none. Rod photoreceptor sensitivity, S (ROD), was derived from the a-wave, and post-receptor sensitivity, log σ, was calculated from the b-wave stimulus-response function. Dark adapted visual threshold was measured using a forced-choice preferential procedure. For S (ROD), the deficit from normal for age varied significantly with ROP severity but not with age group. For log σ, in mild ROP, the deficit was smaller in older subjects than in infants, while in severe ROP, the deficit was quite large in both age groups. In subjects who never had ROP, S (ROD) and log σ in both age groups were similar to those in term born controls. Deficits in dark adapted threshold and log σ were correlated in mild but not in severe ROP. The data are evidence that sensitivity of the post-receptor retina improves in those with a history of mild ROP. We speculate that beneficial reorganization of the post-receptor neural circuitry occurs in mild but not in severe ROP.

Inlet for fuel assembly having finger control rods

International Nuclear Information System (INIS)

Berglund, A.; Suvanto, A.; Tornblom, L.

1975-01-01

A nuclear reactor with vertically arranged fuel assemblies positioned on supporting members and with control rods displaceably arranged in guide tubes between the fuel rods inside the fuel assemblies is described. The supporting plate is provided with a transverse end piece with throttling means for the liquid flow which passes from below up through the supporting member and past the fuel rods in the fuel assembly. The inlets for the guide tubes for the control rods are located below the end piece and the throttling means. In this way a higher pressure prevails at the inlet to the guide tubes than above the end piece, so that a stronger flow of coolant is produced through guide tubes than through the fuel assembly. (U.S.)

Device for detecting defective nuclear reactor fuel rods

International Nuclear Information System (INIS)

Steven, J.

1976-01-01

A moisture sensor is provided for a nuclear fuel rod for water-cooled nuclear reactors wherein moisture can be present. The fuel rod has an end cap and a charge of nuclear fuel. The moisture sensor is disposed between the end cap and the charge and serves to detect a leak in the fuel rod. The moisture sensor includes a capsule-like housing having an inner space and having openings through which moisture can pass into the inner space in the event of a leak in the fuel rod. Ferromagnetic material is disposed in the inner space of the housing together with a moisture detector responsive to moisture for altering the diposition of the ferromagnetic material in the inner space. 5 claims, 6 drawing figures

Method of operating control rods for BWR type reactors

International Nuclear Information System (INIS)

Shirakawa, Toshihisa.

1979-01-01

Purpose: To eliminate the danger such as fuel element failures due to rapid power increase and form a control rod pattern for obtaining a required power level in a relatively short time. Method: Control rods are disposed so as to vertically enter into and retract from the central region of each four fuel assemblies adjacent to each other respectively. Upon operation of the control rods, every other control rods in the lateral and longitudinal directions among the entire control rods that are inserted completely are extracted completely at the lower flow limit of coolants. Then, the control rods completely inserted are divided into groups inserted deeply and groups inserted less deeply. The less deeply inserted groups are extracted just before the excess of thermal limit value successively in the lower flow limit of the coolants and then the deeply inserted groups are extracted successively till a predetermined power level in the same manner. Therefore, the coolant flow to the reactor core is increased and the power level is raised. (Furukawa, Y.)

Calculation method for control rod dropping time in reactor

International Nuclear Information System (INIS)

Nogami, Takeki; Kato, Yoshifumi; Ishino, Jun-ichi; Doi, Isamu.

1996-01-01

If a control rod starts dropping, the dropping speed is rapidly increased, then settled substantially constant, rapidly decreased when it reaches a dash pot. A second detection signal generated by removing an AC component from a first detection signal is differentiated twice. The time when the maximum value among the twice differentiated values is generated is determined as a time when the control rods starts dropping. The time when minimum value among the twice differentiated values is generated is determined as a time when the control rod reaches the dash pot of the reactor. The measuring time within a range from the time when the control rod starts dropping to the time when the control rod reaches the dash pot of the reactor is determined. As a result, processing for the calculation of the dropping start time and dash pot reaching time of the control rod can be automatized. Further, it is suffice to conduct differentiation twice till the reaching time, which can facilitate the processing thereby enabling to determine a reliable time range. (N.H.)

CONTROL ROD

Science.gov (United States)

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)

Regulatory perspective on incomplete control rod insertions

International Nuclear Information System (INIS)

Chatterton, M.

1997-01-01

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

Cone rod dystrophies

Science.gov (United States)

Hamel, Christian P

2007-01-01

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

Cone rod dystrophies

Directory of Open Access Journals (Sweden)

Hamel Christian P

2007-02-01

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

Method of changing the control rod pattern in BWR type reactors

International Nuclear Information System (INIS)

Yoshida, Kenji.

1984-01-01

Purpose: To enable to change the control rod pattern in a short time with ease, as well as improve the availability factor of the reactor. Method: Control rods other than those being inserted into the reactor core are inserted into the reactor core to reduce the power by the reduction in the reactor core flow rate. Then, the control rod to be operated is operated partially for the change of the control rod pattern to restrict the linear heat rating of the fuels to less than 0.1 kW/ft per one hour to change the control pattern to the aimed control rod pattern. Then, the reactor core flow rate is increased after the pattern exchange for the control rod to increase the power. Since only the control rod operation is performed without adjusting the reactor core flow rate upon change of the control rod pattern, procedures can be made simply in a short time to thereby improve the availability factor of the reactor. (Moriyama, K.)

The turbulent flow in rod bundles

International Nuclear Information System (INIS)

Moeller, S.V.

1989-01-01

Experimental studies have shown that the axial and azimuthal turbulence intensities in the gap regions of rod bundles increase strongly with decreasing rod spacing; the fluctuating velocities in the axial and azimuthal directions have a quasi-periodic behaviour. To determine the origin of this phenomenon, an its characteristics as a function of the geometry and the Reynolds number, an experimental investigation was performed on the turbulent in several rod bundles with different aspect ratios (P/D, W/D). Hot-wires and microsphones were used for the measurements of velocity and wall pressure fluctuations. The data were evaluated to obtain spectra as well as auto and cross correlations. Based on the results, a phenomenological model is presented to explain this phenomenon. By means of the model, the mass exchange between neighbouring subchannels is explained [pt

Device and method of cooling control rod drives

International Nuclear Information System (INIS)

Togashi, Hidetoshi; Mase, Noriaki; Matsumura, Yuichi.

1985-01-01

Purpose: To prevent the generation of local temperature rise depending on the reactor core position of the control rod drives and control the temperature to an averaged state in BWR type reactors. Method: Control rod drives having a large charging length of the housing in the pressure vessel involve such a factor that the temperature of the control rod drives is increased by the synergistic effect due to the radiation heat from the reactor core and to the unevenness of the cooling water flow rate, which renders an appropriate temperature control difficult for the reactor core position. A cooling water flow rate controlling device having a restriction mechanism is disposed on the cooling water feed path for each of the hydraulic control units of the control rod drives, so that flow rate to the control rod drives is increased at the center of the reactor core and decreased at the periphery thereof. As a result, average temperature state can be set, temperature increase due to cloggings can be prevented and the thermal effect can be eliminated to thereby improve the reliability. (Moriyama, K.)

Plasmonic-cavity model for radiating nano-rod antennas

DEFF Research Database (Denmark)

Peng, Liang; Mortensen, N. Asger

2014-01-01

In this paper, we propose the analytical solution of nano-rod antennas utilizing a cylindrical harmonics expansion. By treating the metallic nano-rods as plasmonic cavities, we derive closed-form expressions for both the internal and the radiated fields, as well as the resonant condition and the ......In this paper, we propose the analytical solution of nano-rod antennas utilizing a cylindrical harmonics expansion. By treating the metallic nano-rods as plasmonic cavities, we derive closed-form expressions for both the internal and the radiated fields, as well as the resonant condition...... and the radiation efficiency. With our theoretical model, we show that besides the plasmonic resonances, efficient radiation takes advantage of (a) rendering a large value of the rods' radius and (b) a central-fed profile, through which the radiation efficiency can reach up to 70% and even higher in a wide...... frequency band. Our theoretical expressions and conclusions are general and pave the way for engineering and further optimization of optical antenna systems and their radiation patterns....

Cadmium safety rod thermal tests

International Nuclear Information System (INIS)

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

1992-01-01

Thermal testing of cadmium safety rods was conducted as part of a program to define the response of Savannah River Site (SRS) production reactor core components to a hypothetical LOCA leading to a drained reactor tank. The safety rods are present in the reactor core only during shutdown and are not used as a control mechanism during operation; thus, their response to the conditions predicted for the LOCA is only of interest to the extent that it could impact the progression of the accident. This document provides a description of this testing

Self-cleaning threaded rod spinneret for high-efficiency needleless electrospinning

Science.gov (United States)

Zheng, Gaofeng; Jiang, Jiaxin; Wang, Xiang; Li, Wenwang; Zhong, Weizheng; Guo, Shumin

2018-07-01

High-efficiency production of nanofibers is the key to the application of electrospinning technology. This work focuses on multi-jet electrospinning, in which a threaded rod electrode is utilized as the needless spinneret to achieve high-efficiency production of nanofibers. A slipper block, which fits into and moves through the threaded rod, is designed to transfer polymer solution evenly to the surface of the rod spinneret. The relative motion between the slipper block and the threaded rod electrode promotes the instable fluctuation of the solution surface, thus the rotation of threaded rod electrode decreases the critical voltage for the initial multi-jet ejection and the diameter of nanofibers. The residual solution on the surface of threaded rod is cleaned up by the moving slipper block, showing a great self-cleaning ability, which ensures the stable multi-jet ejection and increases the productivity of nanofibers. Each thread of the threaded rod electrode serves as an independent spinneret, which enhances the electric field strength and constrains the position of the Taylor cone, resulting in high productivity of uniform nanofibers. The diameter of nanofibers decreases with the increase of threaded rod rotation speed, and the productivity increases with the solution flow rate. The rotation of electrode provides an excess force for the ejection of charged jets, which also contributes to the high-efficiency production of nanofibers. The maximum productivity of nanofibers from the threaded rod spinneret is 5-6 g/h, about 250-300 times as high as that from the single-needle spinneret. The self-cleaning threaded rod spinneret is an effective way to realize continuous multi-jet electrospinning, which promotes industrial applications of uniform nanofibrous membrane.

Dynamics of Longitudinal Impact in the Variable Cross-Section Rods

Science.gov (United States)

Stepanov, R.; Romenskyi, D.; Tsarenko, S.

2018-03-01

Dynamics of longitudinal impact in rods of variable cross-section is considered. Rods of various configurations are used as elements of power pulse systems. There is no single method to the construction of a mathematical model of longitudinal impact on rods. The creation of a general method for constructing a mathematical model of longitudinal impact for rods of variable cross-section is the goal of the article. An elastic rod is considered with a cross-sectional area varying in powers of law from the longitudinal coordinate. The solution of the wave equation is obtained using the Fourier method. Special functions are introduced on the basis of recurrence relations for Bessel functions for solving boundary value problems. The expression for the square of the norm is obtained taking into account the orthogonality property of the eigen functions with weight. For example, the impact of an inelastic mass along the wide end of a conical rod is considered. The expressions for the displacements, forces and stresses of the rod sections are obtained for the cases of sudden velocity communication and the application of force. The proposed mathematical model makes it possible to carry out investigations of the stress-strain state in rods of variable and constant cross-section for various conditions of dynamic effects.

Simple measuring rod method for the coaxiality of serial holes

Science.gov (United States)

Wang, Lei; Yang, Tongyu; Wang, Zhong; Ji, Yuchen; Liu, Changjie; Fu, Luhua

2017-11-01

Aiming at the rapid coaxiality measurement of serial hole part with a small diameter, a coaxiality measuring rod for each layer hole with a single LDS (laser displacement sensor) is proposed. This method does not require the rotation angle information of the rod, and the coaxiality of serial holes can be calculated from the measured values of LDSs after randomly rotating the measuring rod several times. With the mathematical model of the coaxiality measuring rod, each factor affecting the accuracy of coaxiality measurement is analyzed by simulation, and the installation accuracy requirements of the measuring rod and LDSs are presented. In the tolerance of a certain installation error of the measuring rod, the relative center of the hole is calculated by setting the over-determined nonlinear equations of the fitting circles of the multi-layer holes. In experiment, coaxiality measurement accuracy is realized by a 16 μm precision LDS, and the validity of the measurement method is verified. The manufacture and measurement requirements of the coaxiality measuring rod are low, by changing the position of LDSs in the measuring rod, the serial holes with different sizes and numbers can be measured. The rapid coaxiality measurement of parts can be easily implemented in industrial sites.

Apparatus for inspecting the quality of nuclear fuel rod ends

International Nuclear Information System (INIS)

Brashier, R.W.; Pfau, E.D.

1990-01-01

This patent describes an apparatus for inspecting the quality of both ends of nuclear fuel rods. It comprises: a housing including a pair of longitudinally separated slots for receiving X-ray downwardly therethrough from an external source and so as to define first and second longitudinally spaced apart operating positions, means for serially guiding nuclear fuel rods longitudinally through the housing and to a first rod position wherein the forward ends of the rods are aligned below the first operating position and to a second rod position wherein the rear ends of the rods are aligned below the second operating position, belt conveyor assembly means for serially advancing X-ray film cartridges longitudinally through the housing and below the rods, and so that each cartridge may be selectively aligned below the first and second operating positions; and table means supported by the conveyor frame for selectively lifting the film cartridges supported by the belts and so that the conveyor belts may be advanced while the film cartridges are held stationary

Method for wrapping a wire round a nuclear fuel rod

International Nuclear Information System (INIS)

Nakayasu, Fumio.

1974-01-01

Object: To provide a method for winding a wire round a nuclear fuel rod with accurate pitches without imparting any local strain or torsion to the wire. Structure: A wire is fixed on one end of the fuel rod, and the other end of the wire is secured to a universal joint leaving a winding allowance to the fuel rod. The wire is linearly stretched by a predetermined tension through the universal joint so as to provide an angle of development theta corresponding to the desired winding pitch, and then, the fuel rod may be rotated so that the end of the wire on the side of the universal joint is moved towards the fuel rod so as to render the angle of development theta constant in proportion to said rotation of the fuel rod. (Kamimura, M.)

Process development and fabrication for sphere-pac fuel rods

International Nuclear Information System (INIS)

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

1981-06-01

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

System for manipulating radioactive fuel rods within a nuclear fuel assembly

International Nuclear Information System (INIS)

Tolino, R.W.; King, W.E.; Blickenderfer, J.L.; Roth, C.H. Jr.

1987-01-01

A tool is described for manipulating the peripherally located fuel rods of a fuel assembly so that the rods can be visually inspected. The fuel assembly includes top and bottom nozzles, each of which is connected to a support skeleton, as well as grids, and wherein the rods are retained within the grids and confined between the top and bottom nozzles thereof. It consists of: (a) a fixture that is detachably connectable to one of the nozzles of the fuel assembly. The fixture having holes therein, (b) rotating means pivotally mountable within the holes of the fixture for selectively gripping and rotating the rod, and (c) a displacing means mounted on the fixture for reciprocably displacing the rods within the fuel assembly, including a lifting assembly and a push-down assembly for lifting and pushing down a selected one of the rods, respectively, whereby the rods can be selectively rotated, lifted, and pushed down in order to expose portions of the rods which are normally hidden to visual inspection while the nozzles stay connected to the support skeleton and the rods stay confined between the top and bottom nozzles of the fuel assembly

Dynamic insertion analysis of control rods of BWR under seismic excitation

International Nuclear Information System (INIS)

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

2007-01-01

The dynamic insertion characteristics of the control rods for the boiling water reactors under the seismic excitation are investigated using non-linear analytical models. The control rod insertion capability is one of the most important items for the safety of nuclear power plants under the seismic events. Predicting the control rod insertion behavior during the earthquake is important in the course of the control rod seismic design. We developed the analytical models using the finite element method (FEM). The effect of the interaction force between the control rod and the fuel assemblies is considered in the non-linear analysis. This interaction force courses the resistance force to the control rod during its insertion behavior. The validity of analytical methods was confirmed by comparing the analytical results with the experimental ones. Using the analytical models, the effects of input seismic motion and structural parameters of the control rods and the fuel assemblies, such as the thickness of the channel box, on the insertion time are investigated. These analytical methods can predict insertion time of the control rod, and are useful for the seismic design of the control rod assemblies. (author)

Behavior of defective LWR-type fuel rods irradiated under postulated accident conditions

International Nuclear Information System (INIS)

Hobbins, R.R.; Croucher, D.W.; Seiffert, S.L.; Cook, B.A.; Kerwin, D.K.; Mehner, A.S.; Ploger, S.A.

1979-05-01

The irradiation experiments reported here have been conducted by the Thermal Fuels Behavior Program of EG and G Idaho, Inc., for the United States Nuclear Regulatory Commission in the Power Burst Facility (PBF) at the Idaho National Engineering Laboratory. Five of the rods were irradiated in PCM tests and one in a LOC test. During these tests, the six rods lost cladding integrity prior to or during the transient phase of the test due to either manufacturing defects or intentional rod design and operation. Of the five defective rods tested under PCM conditions, one (Rod IE-008, Test IE-1) had a hydride rupture below the region of the rod, which was in film boiling during the transient; two (Rod A-0021, Test PCM-3 and Rod IE-019, Test IE-5) contained defects (a pin hole and a small axial crack, respectively) within the film boiling zone; and two (Rod 201-1, Test PCM-1 and Rod 205-8, Test PCM-5) failed by cladding embrittlement within the film boiling zone. Rod 312-3 was waterlogged before being subjected to LOC conditions in Test LLR-3

Study on the quantitative rod internal pressure design criterion

International Nuclear Information System (INIS)

Kim, Kyu Tae; Kim, Oh Hwan; Han, Hee Tak

1991-01-01

The current rod internal pressure criterion permits fuel rods to operate with internal pressures in excess of system pressure only if internal overpressure does not cause the diametral gap enlargement. In this study, the generic allowable internal gas pressure not violating this criterion is estimated as a function of rod power. The results show that the generic allowable internal gas pressure decreases linearly with the increase of rod power. Application of the generic allowable internal gas pressure for the rod internal pressure design criterion will result in the simplication of the current design procedure for checking the diametral gap enlargement caused by internal overpressure because according to the current design procedure the cladding creepout rate should be compared with the fuel swelling rate at each axial node at each time step whenever internal pressure exceeds the system pressure. (Author)

SFAK, Unscattered Gamma Self-Absorption from Regular Fuel Rod Assemblies

International Nuclear Information System (INIS)

Wand, H.

1982-01-01

1 - Description of problem or function: Calculation of the self- absorption of unscattered (gamma-) radiation from fuel assemblies which contain a regular arrangement of identical fuel rods. 2 - Method of solution: The point-kernel is integrated over the radiation sources, i.e. the fuel rods. A uniform mesh of integration points is used for each of the fuel rods. 3 - Restrictions on the complexity of the problem: Number of fuel rods is dynamically allocated

FLECHT-SEASET 21-rod bundle flow blockage heat transfer during reflood

International Nuclear Information System (INIS)

Loftus, M.; Hochreiter, L.; Lee, N.

1983-01-01

The effect of various flow blockage shapes and distributions during a PWR reflood was investigated using six 21-rod bundles with full length, internally heated, cosine power-shaped electrical rods. The flow blockage shapes, simulating the fuel rod clad ballooning, were made of thin-wall stainless steel tubes hydroformed into a short, concentric shape and along, nonconcentric shape. The blockage sleeves were distributed both coplanar, with all sleeves located at the same elevation, and non-coplanar. The initial and boundary conditions were varied to include parametric effects of pressure, inlet water temperature, and primarily, flooding rate. The initial mid-plane rod temperature was 871 0 C (1600 0 F) in all tests. Rod and vapor temperature measurements were made throughout the rod bundle with emphasis on the blockage region. The rod heat transfer downstream of the blockage was found to be greater for rods in a blocked bundle than for similar rods in an unblocked bundle. The heat transfer improvement decreases both with time after flood initiation and as the distance increased downstream of the blockage. The improvement in the heat transfer is attributed primarily to the breakup of the water droplets entrained in the steam flow. The smaller droplets subsequently evaporate and desuperheat the steam, which then improves the heat transfer between the rods and the steam in and downstream of the blockage zone

Design requirement on KALIMER control rod assembly duct

International Nuclear Information System (INIS)

Hwang, W.; Kang, H. Y.; Nam, C.; Kim, J. O.; Kim, Y. J.

1998-03-01

This document establishes the design guidelines which are needs for designing the control rod assembly duct of the KALIMER as design requirements. it describes control rod assembly duct of the KALIMER and its requirements that includes functional requirements, performance requirements, interfacing systems, design limits and strength requirements, seismic requirements, structural requirements, environmental requirements, reliability and safety requirements, standard and codes, QA programs, and other requirements. The control rod system consists of three parts, which are drive mechanism, drive-line, and absorber bundle. This report deals with the absorber bundle and its outer duct only because the others are beyond the scope of fuel system design. The guidelines for design requirements intend to be used for an improved design of the control rod assembly duct of the KALIMER. (author). 19 refs

Design requirement on KALIMER control rod assembly duct

Energy Technology Data Exchange (ETDEWEB)

Hwang, W.; Kang, H. Y.; Nam, C.; Kim, J. O.; Kim, Y. J

1998-03-01

This document establishes the design guidelines which are needs for designing the control rod assembly duct of the KALIMER as design requirements. it describes control rod assembly duct of the KALIMER and its requirements that includes functional requirements, performance requirements, interfacing systems, design limits and strength requirements, seismic requirements, structural requirements, environmental requirements, reliability and safety requirements, standard and codes, QA programs, and other requirements. The control rod system consists of three parts, which are drive mechanism, drive-line, and absorber bundle. This report deals with the absorber bundle and its outer duct only because the others are beyond the scope of fuel system design. The guidelines for design requirements intend to be used for an improved design of the control rod assembly duct of the KALIMER. (author). 19 refs.

Vibration Analysis for Monitoring of Ancient Tie-Rods

Directory of Open Access Journals (Sweden)

L. Collini

2017-01-01

Full Text Available This paper presents an application of vibration analysis to the monitoring of tie-rods. An algorithm for the axial load estimation based on experimentally measured natural frequencies is introduced and its application to a case study is reported. The proposed model of a tie-rod incorporates elastic bed-type boundary conditions that represent the contact between stonework and the tie-rod. The weighed differences between experimentally and numerically determined frequencies are minimized with respect to the parameters of the model, the main being the axial load and the stiffness at the tie-rod/wall interface. Thus, the multidimensional optimization problem is solved. Results are analysed in comparison to a model with simple fixed-end boundary conditions. In addition, the analytical formulation of the problem is delivered.

Modelling of pellet cladding interaction during power ramps in PWR rods by means of Transuranus fuel rod analysis code

International Nuclear Information System (INIS)

Di Marcello, V.; Luzzi, L.

2008-01-01

Pellet-cladding interaction (PCI) in PWR type rods subjected to power ramps was analysed by means of TRANSURANUS (TU) fuel rod performance code. PCI phenomena depend on the fuel power history - i.e. by several irradiation and thermal induced phenomena occurring in the fuel rod and mutually interacting during its life in reactor - and may become critical for cladding integrity under accidental conditions. Ten test fuel rods, whose power histories and post irradiation experiment (PIE) data were available from the OECD/NEA-IAEA International Fuel Performance Experiment (UTE) database through the Studsvik SUPER-RAMP Project, were simulated by TRANSURANUS. During a power ramp pellet gaseous swelling can be inhibited by cladding pressure and can be over-predicted by a normal operation swelling model. This phenomenon was simulated by a new formulation of a fuel swelling model already available in the code, in order to consider hot pressing of inter-granular -fuel porosity due to the high hydrostatic stress resulting from PCI: it was found that TRANSURANUS, as a result of the proposed swelling formulation as well as of the accurate modelling of the other phenomena occurring during irradiation, gives correct predictions on PCI induced fuel rod failures. In addition, PCI failure threshold identified by TRANSURANUS was compared with the technological limits known in literature: the possibility of relaxing these limits for low burn-up values and the preponderance of the European fuel rod design in front of PCI emerged from TU analyses. Finally, a good agreement was found between TU evaluations and PIE data, with regard to fission gas release, fuel grain growth, and creep, corrosion and elongation of the cladding. (authors)

Broadband Vibration Attenuation Using Hybrid Periodic Rods

Directory of Open Access Journals (Sweden)

S. Asiri

2008-12-01

Full Text Available This paper presents both theoretically and experimentally a new kind of a broadband vibration isolator. It is a table-like system formed by four parallel hybrid periodic rods connected between two plates. The rods consist of an assembly of periodic cells, each cell being composed of a short rod and piezoelectric inserts. By actively controlling the piezoelectric elements, it is shown that the periodic rods can efficiently attenuate the propagation of vibration from the upper plate to the lower one within critical frequency bands and consequently minimize the effects of transmission of undesirable vibration and sound radiation. In such a system, longitudinal waves can propagate from the vibration source in the upper plate to the lower one along the rods only within specific frequency bands called the "Pass Bands" and wave propagation is efficiently attenuated within other frequency bands called the "Stop Bands". The spectral width of these bands can be tuned according to the nature of the external excitation. The theory governing the operation of this class of vibration isolator is presented and their tunable filtering characteristics are demonstrated experimentally as functions of their design parameters. This concept can be employed in many applications to control the wave propagation and the force transmission of longitudinal vibrations both in the spectral and spatial domains in an attempt to stop/attenuate the propagation of undesirable disturbances.

Multiple rod–cone and cone–rod photoreceptor transmutations in snakes: Evidence from visual opsin gene expression

Science.gov (United States)

Simoe, Bruno F; Sampaio, Filipa L.; Loew, Ellis R.; Sanders, Kate L.; Fisher, Robert N.; Hart, Nathan S.; Hunt, David M.; Partridge, Julian C.; Gower, David J.

2016-01-01

In 1934, Gordon Walls forwarded his radical theory of retinal photoreceptor ‘transmutation’. This proposed that rods and cones used for scotopic and photopic vision, respectively, were not fixed but could evolve into each other via a series of morphologically distinguishable intermediates. Walls' prime evidence came from series of diurnal and nocturnal geckos and snakes that appeared to have pure-cone or pure-rod retinas (in forms that Walls believed evolved from ancestors with the reverse complement) or which possessed intermediate photoreceptor cells. Walls was limited in testing his theory because the precise identity of visual pigments present in photoreceptors was then unknown. Subsequent molecular research has hitherto neglected this topic but presents new opportunities. We identify three visual opsin genes, rh1, sws1 and lws, in retinal mRNA of an ecologically and taxonomically diverse sample of snakes central to Walls' theory. We conclude that photoreceptors with superficially rod- or cone-like morphology are not limited to containing scotopic or photopic opsins, respectively. Walls' theory is essentially correct, and more research is needed to identify the patterns, processes and functional implications of transmutation. Future research will help to clarify the fundamental properties and physiology of photoreceptors adapted to function in different light levels.

In-pile experiments on fuel rod behavior during a LOCA

International Nuclear Information System (INIS)

Karb, E.; Pruessmann, M.; Sepold, L.

1980-05-01

This report describes the results of the Test Series F, Tests F 1 through F 5, in the in-pile experimental program with single rods in the DK loop of the FR2 reactor at the Kernforschungszentrum Karlsruhe (KfK). The research is part of the Nuclear Safety Project's (PNS) fuel behavior program. The main objective of the FR2-LOCA tests is to provide information about the effects of a nuclear environment on the mechanisms of fuel rod failure in the second heatup phase of a LOCA. The test rods have a heated length of 50 cm, and their radial dimensions are identical with those of a commercial German PWR. The main parameter of the FR2-LOCA test program is the burnup. The F tests were perfomed from Oct. 25, 1977 to Nov. 22, 1977. They were the first tests in this program to use pre-irradiated fuel rods. The nominal burnup of the test rods was 20 000 MWd/t. During the transient test, the test rods were subjected to rod powers between 36 and 41 W/cm and were pressurized with He to hot internal pressures between 46 and 83 bar. The test rods during the heatup phase at pressures of 56, 53, 42, 72 and 60 bar, respectively. The burst temperatures were determined to be 890, 893, 932, 835 and 880 0 C for test F 1 through F 5. The maximum total circumferential elongations amount to 59, 38, 27, 34 and 41%, respectively. The F tests revealed a fragmentation of the fuel after the irradiation (prior to the tests) and a disintegration of the fuel pellet column after the transient tests due to cladding ballooning. The post-test results indicated a significant reduction of the pellet stack length for all five test rods. The burst data of the F tests did not reveal any difference between tests with unirradiated fuel rods and the irradiated fuel rods of this test series. (orig./HP) [de

Prototypical Rod Consolidation Demonstration Project

International Nuclear Information System (INIS)

1993-05-01

The objective of Phase 3 of the Prototypical Rod consolidation Demonstration Project (PRCDP) was to procure, fabricate, assemble, and test the Prototypical Rod consolidation System as described in the NUS Phase 2 Final Design Report. This effort required providing the materials, components, and fabricated parts which makes up all of the system equipment. In addition, it included the assembly, installation, and setup of this equipment at the Cold Test Facility. During the Phase 3 effort the system was tested on a component, subsystem, and system level. This volume 1, discusses the PRCDP Phase 3 Test Program that was conducted by the HALLIBURTON NUS Environmental Corporation under contract AC07-86ID12651 with the United States Department of Energy. This document, Volume 1, Book 4 discusses the following topics: Rod Compaction/Loading System Test Results and Analysis Report; Waste Collection System Test Results and Analysis Report; Waste Container Transfer Fixture Test Results and Analysis Report; Staging and Cutting Table Test Results and Analysis Report; and Upper Cutting System Test Results and Analysis Report

Dry Rod Consolidation Technology Project results

International Nuclear Information System (INIS)

Mullen, C.K.; Feldman, E.M; Vinjamuri, K.; Griebenow, B.L.; Lynch, R.J.; Arave, A.E.; Hill, R.C.

1988-01-01

The Dry Rod Consolidation Technology (DRCT) Project conducted at the Idaho National Engineering Laboratory (INEL), in 1987 demonstrated the technical feasibility of a dry horizontal fuel rod consolidation process. Fuel rods from Westinghouse 15 /times/ 15 pressurized water reactor (PWR) spent fuel assemblies were consolidated into canisters to achieve a 2:1 volume reduction ratio. The consolidation equipment was operated at an existing hot cell complex at the INEL. The equipment was specifically designed to interface with the existing facility fuel handling and operational capabilities and was instrumented to provide data collection for process technology research. During the operational phase, data were collected from observation of the consolidation process, fuel assembly handling, and fuel rod behavior and characteristics. Equipment performance was recorded and data measurements were compiled on crud and contamination generated and spread. Fuel assembly skeletons [non-fuel bearing components (NFBC)] were gamma scanned and analyzed for isotopic content and profile. The above data collection was enhanced by extensive photograph and video documentation. The loaded consolidation fuel canisters were utilized for a test of the Transnuclear, Inc. TN-24P dry storage cask with consolidated fuel. The NFBC material was stored for a future volume reduction demonstration project. 14 figs., 4 tabs

Oxide nano-rod array structure via a simple metallurgical process

International Nuclear Information System (INIS)

Nanko, M; Do, D T M

2011-01-01

A simple method for fabricating oxide nano-rod array structure via metallurgical process is reported. Some dilute alloys such as Ni(Al) solid solution shows internal oxidation with rod-like oxide precipices during high-temperature oxidation with low oxygen partial pressure. By removing a metal part in internal oxidation zone, oxide nano-rod array structure can be developed on the surface of metallic components. In this report, Al 2 O 3 or NiAl 2 O 4 nano-rod array structures were prepared by using Ni(Al) solid solution. Effects of Cr addition into Ni(Al) solid solution on internal oxidation were also reported. Pack cementation process for aluminizing of Ni surface was applied to prepare nano-rod array components with desired shape. Near-net shape Ni components with oxide nano-rod array structure on their surface can be prepared by using the pack cementation process and internal oxidation,

Performance analysis of LMFBR control rods

International Nuclear Information System (INIS)

Pitner, A.L.; Birney, K.R.

1975-01-01

Control rods in the FFTF and LMFBR's will consist of pin bundles of stainless steel-clad boron carbide pellets. In the FFTF reference design, sixty-one pins of 0.474-inch diameter each containing a 36-inch stack of 0.362-inch diameter boron carbide pellets comprise a control rod. Reactivity control is provided by the 10 B (n,α) 7 Li reaction in the boron carbide. This reaction is accompanied by an energy release of 2.8 MeV, and heating from this reaction typically approaches 100 watts/cm 3 for natural boron carbide pellets in an LMFBR flux. Performance analysis of LMFBR control rods must include an assessment of the thermal performance of control pins. In addition, irradiation performance with regard to helium release, pellet swelling, and reactivity worth depletion as a function of service time must be evaluated

Inferred residual gaps for IFA-527 rods, compared to PIE measurements

International Nuclear Information System (INIS)

Lanning, D.D.

1983-05-01

The NRC-sponsored assembly IFA-527 contained six xenon-filled rods, each instrumented with fuel thermocouples at each end. Five of the six rods contained stable UO 2 fuel pellets with a fabricated diametral gap size of 230 microns. The rods were unique among instrumented test rods in that their lifetime peak powers were quite low (less than 20 kW/m). The assembly was removed at low exposure (about 2 MWd/kgM) due to pressure seal failures. The prefailure measured fuel temperatures in the five identical rods were quite similar, and indicated an effective fuel relocation of about 30 to 35 percent of the as-fabricated gap. Examination of rod cross sections in PIE, however, reveals virtually no reduction in the as-fabricated gap. In an attempt to reconcile these two observations, it is demonstrated that pellet eccentricity could account for the inferred effective relocation for these xenon-filled rods. That analysis is extended to an estimate of the proper effective relocation value for low-powered He-filled rods

Suppressing turbulence of self-propelling rods by strongly coupled passive particles.

Science.gov (United States)

Su, Yen-Shuo; Wang, Hao-Chen; I, Lin

2015-03-01

The strong turbulence suppression, mainly for large-scale modes, of two-dimensional self-propelling rods, by increasing the long-range coupling strength Γ of low-concentration passive particles, is numerically demonstrated. It is found that large-scale collective rod motion in forms of swirls or jets is mainly contributed from well-aligned dense patches, which can push small poorly aligned rod patches and uncoupled passive particles. The more efficient momentum transfer and dissipation through increasing passive particle coupling leads to the formation of a more ordered and slowed down network of passive particles, which competes with coherent dense active rod clusters. The frustration of active rod alignment ordering and coherent motion by the passive particle network, which interrupt the inverse cascading of forming large-scale swirls, is the key for suppressing collective rod motion with scales beyond the interpassive distance, even in the liquid phase of passive particles. The loosely packed active rods are weakly affected by increasing passive particle coupling due to the weak rod-particle interaction. They mainly contribute to the small-scale modes and high-speed motion.

Analysis of the rod drop accident for Angra-1

International Nuclear Information System (INIS)

Veloso, M.A.; Atayde, P.A.

1989-01-01

The aim of this work is to present a rod drop accident analysis for the third cycle of the Angra-1 nuclear power plant operating in the automatic control mode. In this analysis all possible configurations for dropped rods caused by a single failure in the controller circuits have been considered. The dropped rod worths, power distributions and excore detector tilts were determined by using the Siemens/KWU neutronic code system, in particular the MEDIUM2, PINPOW and DETILT codes. The transient behaviour of the plant during the rod drop event was simulated with the SACI2/MOD0 code, developed at CDTN. Determinations related to the DNBR design limit were conducted by utilizing the CDTN PANTERA-1P subchannel code. The transient analysis indicated that for dropped rod worths greater than about 425 pcm reactor trip from negative neutron flux rate will take place independently of core conditions. In the range from 0 to 425 pcm large power overshoots may occur as a consequence of the automatic control system action. The magnitude of the maximum power peaking during the event increases with the dropped rod worth, as far as the control bank is able to compensate the initial reactivity decrease. Thermal-hydraulic evaluations carried out with the PANTERA-1P code show that for all the relevant dropped rod worths the minimum DNBR will remain above a limit value of 1.365. Even if this conservative limit is met, the calculated nuclear power peaking factors, F N AH , will be at least 6% higher than the allowable F N AH -values. Therefore, the DNBR design margin will be preserved at the event of rod drop. (author)

ZnO Nano-Rod Devices for Intradermal Delivery and Immunization.

Science.gov (United States)

Nayak, Tapas R; Wang, Hao; Pant, Aakansha; Zheng, Minrui; Junginger, Hans; Goh, Wei Jiang; Lee, Choon Keong; Zou, Shui; Alonso, Sylvie; Czarny, Bertrand; Storm, Gert; Sow, Chorng Haur; Lee, Chengkuo; Pastorin, Giorgia

2017-06-15

Intradermal delivery of antigens for vaccination is a very attractive approach since the skin provides a rich network of antigen presenting cells, which aid in stimulating an immune response. Numerous intradermal techniques have been developed to enhance penetration across the skin. However, these methods are invasive and/or affect the skin integrity. Hence, our group has devised zinc oxide (ZnO) nano-rods for non-destructive drug delivery. Chemical vapour deposition was used to fabricate aligned nano-rods on ZnO pre-coated silicon chips. The nano-rods' length and diameter were found to depend on the temperature, time, quality of sputtered silicon chips, etc. Vertically aligned ZnO nano-rods with lengths of 30-35 µm and diameters of 200-300 nm were selected for in vitro human skin permeation studies using Franz cells with Albumin-fluorescein isothiocyanate (FITC) absorbed on the nano-rods. Fluorescence and confocal studies on the skin samples showed FITC penetration through the skin along the channels formed by the nano-rods. Bradford protein assay on the collected fluid samples indicated a significant quantity of Albumin-FITC in the first 12 h. Low antibody titres were observed with immunisation on Balb/c mice with ovalbumin (OVA) antigen coated on the nano-rod chips. Nonetheless, due to the reduced dimensions of the nano-rods, our device offers the additional advantage of excluding the simultaneous entrance of microbial pathogens. Taken together, these results showed that ZnO nano-rods hold the potential for a safe, non-invasive, and painless intradermal drug delivery.

A comparison of thermal algorithms of fuel rod performance code systems

International Nuclear Information System (INIS)

Park, C. J.; Park, J. H.; Kang, K. H.; Ryu, H. J.; Moon, J. S.; Jeong, I. H.; Lee, C. Y.; Song, K. C.

2003-11-01

The goal of the fuel rod performance is to identify the robustness of a fuel rod with cladding material. Computer simulation of the fuel rod performance becomes one of important parts to designed and evaluate new nuclear fuels and claddings. To construct a computing code system for the fuel rod performance, several algorithms of the existing fuel rod performance code systems are compared and are summarized as a preliminary work. Among several code systems, FRAPCON, and FEMAXI for LWR, ELESTRES for CANDU reactor, and LIFE for fast reactor are reviewed. Thermal algorithms of the above codes are investigated including methodologies and subroutines. This work will be utilized to construct a computing code system for dry process fuel rod performance

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

International Nuclear Information System (INIS)

Kohli, R.; Pasupathi, V.

1986-09-01

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

A comparison of thermal algorithms of fuel rod performance code systems

Energy Technology Data Exchange (ETDEWEB)

Park, C. J.; Park, J. H.; Kang, K. H.; Ryu, H. J.; Moon, J. S.; Jeong, I. H.; Lee, C. Y.; Song, K. C

2003-11-01

The goal of the fuel rod performance is to identify the robustness of a fuel rod with cladding material. Computer simulation of the fuel rod performance becomes one of important parts to designed and evaluate new nuclear fuels and claddings. To construct a computing code system for the fuel rod performance, several algorithms of the existing fuel rod performance code systems are compared and are summarized as a preliminary work. Among several code systems, FRAPCON, and FEMAXI for LWR, ELESTRES for CANDU reactor, and LIFE for fast reactor are reviewed. Thermal algorithms of the above codes are investigated including methodologies and subroutines. This work will be utilized to construct a computing code system for dry process fuel rod performance.

Fuel rod attachment system

International Nuclear Information System (INIS)

Christiansen, D.W.

1982-01-01

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

Physics analysis of the gang partial rod drive event

International Nuclear Information System (INIS)

Boman, C.; Frost, R.L.

1992-08-01

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

Spatially confined synthesis of SiOx nano-rod with size-controlled Si quantum dots in nano-porous anodic aluminum oxide membrane.

Science.gov (United States)

Pai, Yi-Hao; Lin, Gong-Ru

2011-01-17

By depositing Si-rich SiOx nano-rod in nano-porous anodic aluminum oxide (AAO) membrane using PECVD, the spatially confined synthesis of Si quantum-dots (Si-QDs) with ultra-bright photoluminescence spectra are demonstrated after low-temperature annealing. Spatially confined SiOx nano-rod in nano-porous AAO membrane greatly increases the density of nucleated positions for Si-QD precursors, which essentially impedes the route of thermally diffused Si atoms and confines the degree of atomic self-aggregation. The diffusion controlled growth mechanism is employed to determine the activation energy of 6.284 kJ mole(-1) and diffusion length of 2.84 nm for SiO1.5 nano-rod in nano-porous AAO membrane. HRTEM results verify that the reduced geometric dimension of the SiOx host matrix effectively constrain the buried Si-QD size at even lower annealing temperature. The spatially confined synthesis of Si-QD essentially contributes the intense PL with its spectral linewidth shrinking from 210 to 140 nm and its peak intensity enhancing by two orders of magnitude, corresponding to the reduction on both the average Si-QD size and its standard deviation from 2.6 to 2.0 nm and from 25% to 12.5%, respectively. The red-shifted PL wavelength of the Si-QD reveals an inverse exponential trend with increasing temperature of annealing, which is in good agree with the Si-QD size simulation via the atomic diffusion theory.

Control rod drives

International Nuclear Information System (INIS)

Yamanaka, Toshikatsu.

1979-01-01

Purpose: To protect bellows against failures due to negative pressure to prevent the loss of pressure balance caused by the expansion of the bellows upon scram. Constitution: An expansion pipe connected to the control rod drive is driven along a guide pipe to insert a control rod into the reactor core. Expansible bellows are provided at the step between the expansion pipe and the guide pipe. Further, a plurality of bore holes or slits are formed on the side wall of the guide pipe corresponding to the expansion portion of the bellows. In such an arrangement, when the expansion pipe falls rapidly and the bellows are expanded upon scram, the volume between each of the pipes of the bellows and the guide pipe is increased to produce a negative pressure, but the effect of the negative pressure on the bellows can be eliminated by the flowing-in of coolants corresponding to that pressure through the bore holes or the slits. (Furukawa, Y.)

Fuel Rod Vibration Measurement Method using a Flap and its Verification

Energy Technology Data Exchange (ETDEWEB)

Yoo, Joo Young; Park, Nam Gyu; Suh, Jung Min; Jeon, Kyeong Lak [KEPCO NF Co., Daejeon (Korea, Republic of)

2011-10-15

Flow-induced vibration is a critical factor for the mechanical integrity of a fuel rod. This vibration can cause leaked fuel through the mechanism, such as grid to rod fretting. To minimize the failures caused by flow-induced vibration, a robust design is needed which takes into account vibrational characteristics. That is, the spacer grid design should be developed to avoid any excessive vibration. On the one hand, if fuel rod vibration can be measured, an estimation of the excitation forces, which are a critical cause of rod failure, should be possible. Therefore, by applying an external force, flow-induced vibration can be roughly estimated when the fuel rod vibration model is used. KEPCO Nuclear Fuel developed the test loop to research flow-induced vibration as shown in Fig.1. The investigation flow-induced vibration (INFINIT) - the test facility - can measure the grid strap vibration and pressure drop of a 5x5 small scale fuel bundle. Basically, using a Laser Doppler Vibrometer (LDV), the vibration of a structure immersed in high speed fluid can be measured. Grid strap vibration is easily measured using an LDV. However, it is quite difficult to measure fuel rod vibration because of the round surface shape of the rods. In addition, measuring current method using the LDV, it was only possible to directly measure fuel rod vibration at the first row of the bundle as the rods behind the first row are obscured. To solve this problem, a thin flap, as shown in Fig. 2(a) can be used as a reflecting target, gaining access to rods within the bundle. The flap is attached to the fuel rod, as in Fig. 2(b). As a result, most of the inner rod vibration can be measured. Before using a flap to measure fuel rod vibration, a verification process was needed to show whether the LDV signal from the flap vibration provided equivalent and reliable signals. Therefore, impact testing was carried out on the fuel rod using a flap. The LDV signals were then compared with accelerometer

Nuclear Fuel Test Rod Fabrication for Data Acquisition Test

International Nuclear Information System (INIS)

Joung, Chang-Young; Hong, Jin-Tae; Kim, Ka-Hye; Huh, Sung-Ho

2014-01-01

A nuclear fuel test rod must be fabricated with precise welding and assembly technologies, and confirmed for their soundness. Recently, we have developed various kinds of processing systems such as an orbital TIG welding system, a fiber laser welding system, an automated drilling system and a helium leak analyzer, which are able to fabricate the nuclear fuel test rods and rigs, and keep inspection systems to confirm the soundness of the nuclear fuel test rods and rids. The orbital TIG welding system can be used with two kinds of welding methods. One can perform the round welding for end-caps of a nuclear fuel test rod by an orbital head mounted in a low-pressure chamber. The other can do spot welding for a pin-hole of a nuclear fuel test rod in a high-pressure chamber to fill up helium gas of high pressure. The fiber laser welding system can weld cylindrical and 3 axis samples such as parts of a nuclear fuel test rod and instrumentation sensors which is moved by an index chuck and a 3 axis (X, Y, Z) servo stage controlled by the CNC program. To measure the real-time temperature change at the center of the nuclear fuel during the irradiation test, a thermocouple should be instrumented at that position. Therefore, a hole needs to be made at the center of fuel pellet to instrument the thermocouple. An automated drilling system can drill a fine hole into a fuel pellet without changing tools or breaking the work-piece. The helium leak analyzer (ASM-380 model of DEIXEN Co.) can check the leak of the nuclear fuel test rod filled with helium gas. This paper describes not only the assembly and fabrication methods used by the process systems, but also the results of the data acquisition test for the nuclear fuel test rod. A nuclear fuel test rod for the data acquisition test was fabricated using the welding and assembling echnologies acquired from previous tests

Nuclear Fuel Test Rod Fabrication for Data Acquisition Test

Energy Technology Data Exchange (ETDEWEB)

Joung, Chang-Young; Hong, Jin-Tae; Kim, Ka-Hye; Huh, Sung-Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-10-15

A nuclear fuel test rod must be fabricated with precise welding and assembly technologies, and confirmed for their soundness. Recently, we have developed various kinds of processing systems such as an orbital TIG welding system, a fiber laser welding system, an automated drilling system and a helium leak analyzer, which are able to fabricate the nuclear fuel test rods and rigs, and keep inspection systems to confirm the soundness of the nuclear fuel test rods and rids. The orbital TIG welding system can be used with two kinds of welding methods. One can perform the round welding for end-caps of a nuclear fuel test rod by an orbital head mounted in a low-pressure chamber. The other can do spot welding for a pin-hole of a nuclear fuel test rod in a high-pressure chamber to fill up helium gas of high pressure. The fiber laser welding system can weld cylindrical and 3 axis samples such as parts of a nuclear fuel test rod and instrumentation sensors which is moved by an index chuck and a 3 axis (X, Y, Z) servo stage controlled by the CNC program. To measure the real-time temperature change at the center of the nuclear fuel during the irradiation test, a thermocouple should be instrumented at that position. Therefore, a hole needs to be made at the center of fuel pellet to instrument the thermocouple. An automated drilling system can drill a fine hole into a fuel pellet without changing tools or breaking the work-piece. The helium leak analyzer (ASM-380 model of DEIXEN Co.) can check the leak of the nuclear fuel test rod filled with helium gas. This paper describes not only the assembly and fabrication methods used by the process systems, but also the results of the data acquisition test for the nuclear fuel test rod. A nuclear fuel test rod for the data acquisition test was fabricated using the welding and assembling echnologies acquired from previous tests.

Substitute safety rods: Physics of operation and irradiation

International Nuclear Information System (INIS)

Baumann, N.P.

1991-01-01

Under certain assumed accidents, an SRS reactor may lose most of its bulk moderator while maintaining flow to fuel assemblies. If this occurs immediately after operation at power, components normally dependent on convective heat transfer to the moderator will heat up with the possibility of melting that component. One component at risk is the currently used cadmium safety rod. A substitute safety rod consisting solely of sintered B 4 C and stainless steel has been designed which is capable of withstanding much higher temperatures. This memorandum provides the physics basis for the adequacy of the rod for reactor shutdown and provides a set of criteria for acceptance in the NTG tests. This memorandum provides physics data for other aspects of operation. These include: Heat production and helium production, along with related phenomena, resulting from inadvertent irradiation at power. Gamma heat input under drained tank conditions. An equivalent rod design suitable for charge design and safety analyses. Degradation under normal operation. Thermal flux ripple in adjacent fuel due to axial striping of alternate B 4 C and steel pellets. Possible effect on safety analyses. Safety rod withdrawal during reactor startup

Neural signal processing for identifying failed fuel rods in nuclear reactors

International Nuclear Information System (INIS)

Seixas, Jose M. de; Soares Filho, William; Pereira, Wagner C.A.; Teles, Claudio C.B.

2002-01-01

Ultrasonic pulses were used for automatic detection of failed nuclear fuel rods. For experimental tests of the proposed method, an assembly prototype of 16 x 16 rods was built by using genuine rods but without fuel inside (just air). Some rods were partially filled with water to simulate cracked rods. Using neural signal processing on the received echoes of the emitted ultrasonic pulses, a detection efficiency of 97% was obtained. Neural detection is shown to outperform other classical discriminating methods and can also reveal important features of the signal structure of the received echoes. (author)

Failed fuel rod detection method by ultrasonic wave

International Nuclear Information System (INIS)

Takamatsu, Masatoshi; Muraoka, Shoichi; Ono, Yukio; Yasojima, Yujiro.

1990-01-01

Ultrasonic wave signals sent from an ultrasonic receiving element are supplied to an evaluation circuit by way of a gate. A table for gate opening and closing timings at the detecting position in each of the fuel rods in a fuel assembly is stored in a memory. A fuel rod is placed between an ultrasonic transmitting element and the receiving element to determine the positions of the transmitting element and the receiving element by positional sensors. The opening and closing timings at the positions corresponding to the result of the detection are read out from the table, and the gates are opened and closed by the timing. This can introduce the ultrasonic wave signals transmitted through a control rod always to the evaluation circuit passing through the gate. Accordingly, the state of failure of the fuel rod can be detected accurately. (I.N.)

Release of fission products by a rod with a manufacturing defect (BREAK). The BOUFFON 08 J 4 experiment

International Nuclear Information System (INIS)

Chenebault, P.; Kurka, G.

The BOUFFON 08 experiment consisted of the circuit irradiation in the Siloe reactor of the thermosiphon and the 15 x 15 type rod with a seal defect simulating a faulty solder (weld). The conditions of power, temperature, and pressure were typical of PWR conditions. The fission products released by the faulty rod and the presence of a water cooling circuit have been analyzed using samples. The γ-activity was essentially due to the volatile fission products; it progressively increased over the course of 25 days of irradiation by the rod and reaches 30 to 35 curies per kW of nuclear waste. Over the course of the test, the activities radiated by the tube circuits were one to two orders higher than those that uniquely correspond to the nuclides that were present in the water: the iodine and noble gases were deposited on the walls and this deposit was continuously resupplied by the circulating water. At the end of irradiation, secondary ruptures were apparent at the level of the fuel; these were revealed by a gross increase in the concentrations of solid fission products and by the abnormal distribution of radioactive nuclides in the fuel