WorldWideScience

Sample records for rod describing solutions

  1. Investigation of Different Colloidal Porous Silicon Solutions and Their Composite Solid Matrix Rods by Optical Techniques

    Science.gov (United States)

    Khan, M. Naziruddin; Aldalbahi, Ali; Almohammedi, Abdullah

    2018-03-01

    Colloidal porous silicon (PSi) in different solvents was synthesized by simple chemical etching. Colloidal solutions were then prepared using different quantities of silicon wafer pieces (Pcs) and chloroplatinic (Pt) acid in catalyst solution. The effect on the properties of the colloidal solutions and composite rods were investigated using various optical characterization techniques. Absorption and photoluminescence (PL) intensity of the colloidal PSi solutions are observed to depend on the quantity of wafer Pcs, the Pt-solution, and the porosity formation on the wafer surface. The morphological structure of the PSi in a solvent and the solid-rod environments were studied using field-emission scanning electron microscopy (FE-SEM) and were observed to have different structures. A mono-oriented structure of PSi exists in tetrahydrofuran, which has stereo orientation in dioxane and dimethylsulfoxide (approximately 5-8 nm as confirmed using high resolution transmission electron microscopy). Subsequently, some colloidal PSi solutions were directly embedded in three types of sol-gel-based matrices, silica, ormosils (or organically modified silica) and polymer, which easily generated solid rods. Spontaneous emission (SE) of the PSi solutions and their composite rods were examined using a high power picosecond 355 nm laser source. The emitted PL and SE signals of the colloidal PSi solutions were dependent on the Pt volume, nature of the solvent, quantity of Si wafer piece, and pumping energy. The response of SE signals from the PSi composites rods is an interesting phenomenon, and such nanocomposites may be used for future research on light amplification.

  2. Two-group Current-equivalent Parameters for Control Rod Cells. Autocode Programme CRCC

    Energy Technology Data Exchange (ETDEWEB)

    Norinder, O; Nyman, K

    1962-06-15

    In two-group neutron diffusion calculations there is mostly necessary to describe the influence of control rods by equivalent homogeneous two-group parameters in regions about the control rods. The problem is solved for a control rod in a medium characterized by two-group parameters. The property of fast and thermal neutr. on current equivalence is selected to obtain equivalent two-group parameters for a homogeneous cell with the same radius as the control rod cell. For the parameters determined one obtains the same fast and thermal neutron current into the rod cell and the equivalent cell independent of the fast and thermal flux amplitudes on the cell boundaries. The equivalent parameters are obtained as a solution of a system of transcendental equations. A Ferranti Mercury Autocode performing the solution is described. Calculated equivalent parameters for control rods in a heavy water lattice are given for some representative cases.

  3. Numerical solution of the elastic non-axial contact between pellet and cladding of fuel rod in PWR

    International Nuclear Information System (INIS)

    Zymak, J.

    1987-08-01

    Elastic non-axial contacts between the pellet and the cladding of a fuel rod in a pressurized water reactor were calculated. The existence and the uniqueness of the solution were proved. The problem was approximated by the finite element method and quadratic programming was used for the solution. The results will be used in the solution of the probabilistic model of a fuel rod with non-axial pellets in a PWR. (author). 10 figs., 4 tabs., 10 refs

  4. Investigation on Minimum Film Boiling Point of Highly Heated Vertical Metal Rod in Aqueous Surfactant Solution

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chi Young; Kim, Jae Han [Pukyong Nat’l Univ., Busan (Korea, Republic of)

    2017-09-15

    In this study, experiments were conducted on the MFB(minimum film boiling) point of highly heated vertical metal rod quenched in aqueous surfactant solution at various temperature conditions. The aqueous Triton X-100 solution(100 wppm) and pure water were used as the liquid pool. Their temperatures ranged from 77 °C to 100 °C. A stainless steel vertical rod of initial center temperature of 500 °C was used as a test specimen. In both liquid pools, as the liquid temperature decreased, the time to reach the MFB point decreased with a parallel increase in the temperature and heat flux of the MFB point. However, over the whole present temperature range, in the aqueous Triton X-100 solution, the time to reach the MFB point was longer, while the temperature and heat flux of the MFB point were reduced when compared with pure water. Based on the present experimental data, this study proposed the empirical correlations to predict the MFB temperature of a high temperature vertical metal rod in pure water and in aqueous Triton X-100 solution.

  5. Finite-element solutions of the AER-2 rod ejection benchmark by CRONOS

    International Nuclear Information System (INIS)

    Kolev, N.P.; Lenain, R.; Fedon-Magnaud, C.

    2001-01-01

    The finite-element option in CRONOS was used to analyse the AER-2 rod-ejection benchmark for WWER-440. The objective is to obtain spatially converged solutions by means of node subdivision and approximation refinement. This paper presents the first phase of analysis dealing with the initial and just-ejected states used for calculation of the initial reactivity. Fine-mesh and extrapolated to zero mesh size solutions were obtained and verified by comparison to MAG code solutions. These differences provide potential for large deviations in the transient results and deserve further attention in reactor safety analysis (Authors)

  6. New soliton solution to the longitudinal wave equation in a magneto-electro-elastic circular rod

    Directory of Open Access Journals (Sweden)

    Aly R. Seadawy

    2018-03-01

    Full Text Available This paper examines the effectiveness of an integration scheme which called the extended trial equation method (ETEM in exactly solving a well-known nonlinear equation of partial differential equations (PDEs. In this respect, the longitudinal wave equation (LWE that arises in mathematical physics with dispersion caused by the transverse Poisson’s effect in a magneto-electro-elastic (MEE circular rod, which a series of exact traveling wave solutions for the aforementioned equation is formally extracted. Explicit new exact solutions are derived in different form such as dark solitons, bright solitons, solitary wave, periodic solitary wave, rational function, and elliptic function solutions of the longitudinal wave equation. The movements of obtained solutions are shown graphically, which helps to understand the physical phenomena of this longitudinal wave equation. Many other such types of nonlinear equations arising in non-destructive evaluation of structures made of the advanced MEE material can also be solved by this method. Keywords: Extended trial equation method, Longitudinal wave equation in a MEE circular rod, Dark solitons, Bright solitons, Solitary wave, Periodic solitary wave

  7. Exact solutions and numerical simulation of longitudinal vibration of the Rayleigh-Love rods with variable cross-sections

    CSIR Research Space (South Africa)

    Shatalov, M

    2012-09-01

    Full Text Available Exact solutions of equations of longitudinal vibration of conical and exponential rod are analyzed for the Rayleigh-Love model. These solutions are used as reference results for checking accuracy of the method of lines. It is shown that the method...

  8. 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.

  9. 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.

  10. Stability of twisted rods, helices and buckling solutions in three dimensions

    KAUST Repository

    Majumdar, Apala; Raisch, Alexander

    2014-01-01

    © 2014 IOP Publishing Ltd & London Mathematical Society. We study stability problems for equilibria of a naturally straight, inextensible, unshearable Kirchhoff rod allowed to deform in three dimensions (3D), subject to terminal loads. We investigate the stability of the twisted, straight state in 3D for three different boundary-value problems, cast in terms of Dirichlet and Neumann boundary conditions for the Euler angles, with and without isoperimetric constraints. In all cases, we obtain explicit stability estimates in terms of the twist, external load and elastic constants and in the Dirichlet case, we compute bifurcation diagrams for the Euler angles as a function of the external load. In the same vein, we obtain explicit stability estimates for a family of prototypical helical equilibria in 3D and demonstrate that they are stable for a range of tensile and compressive forces. We propose a numerical L2-gradient flow model to study the stability and dynamical evolution (in viscous model situations) of Kirchhoff rod equilibria. In Nizette and Goriely 1999 J. Math. Phys. 40 2830-66, the authors construct a family of localized buckling solutions. We apply our L2-gradient flow model to these localized buckling solutions, demonstrate that they are unstable, study their evolution and the simulations demonstrate rich spatio oral patterns that strongly depend on the boundary conditions and imposed isoperimetric constraints.

  11. Stability of twisted rods, helices and buckling solutions in three dimensions

    KAUST Repository

    Majumdar, Apala

    2014-11-03

    © 2014 IOP Publishing Ltd & London Mathematical Society. We study stability problems for equilibria of a naturally straight, inextensible, unshearable Kirchhoff rod allowed to deform in three dimensions (3D), subject to terminal loads. We investigate the stability of the twisted, straight state in 3D for three different boundary-value problems, cast in terms of Dirichlet and Neumann boundary conditions for the Euler angles, with and without isoperimetric constraints. In all cases, we obtain explicit stability estimates in terms of the twist, external load and elastic constants and in the Dirichlet case, we compute bifurcation diagrams for the Euler angles as a function of the external load. In the same vein, we obtain explicit stability estimates for a family of prototypical helical equilibria in 3D and demonstrate that they are stable for a range of tensile and compressive forces. We propose a numerical L2-gradient flow model to study the stability and dynamical evolution (in viscous model situations) of Kirchhoff rod equilibria. In Nizette and Goriely 1999 J. Math. Phys. 40 2830-66, the authors construct a family of localized buckling solutions. We apply our L2-gradient flow model to these localized buckling solutions, demonstrate that they are unstable, study their evolution and the simulations demonstrate rich spatio oral patterns that strongly depend on the boundary conditions and imposed isoperimetric constraints.

  12. 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...

  13. PIN99W, Modelling of VVER and PWR Fuel Rod Thermomechanical Behaviour

    International Nuclear Information System (INIS)

    Valach, M.; Strizhov, P.; Svoboda, R.

    2000-01-01

    1 - Description of program or function: The Code is developed to describe fuel rod thermomechanical behaviour in operational conditions. The main goal of this code is to calculate fuel temperature, gap conductivity, fission gas release and inner gas pressure. 2 - Methods: - fuel rod temperature response is solved by using one-dimensional finite element method combined with weighted residuals method; - the code involves models describing physical phenomena typical for the fuel irradiated in Light Water Power Reactors (densification, restructuring, fission gas release, swelling and relocation) ; - this code is updated and improves PIN-micro code. 3 - Restrictions on the complexity of the problem: - simplified mechanistic solution; - only steady-state solution; - no cladding failure criterion; - no model for axial fuel-cladding interaction

  14. 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

  15. Development and application of an asymmetric deformation model to describe the fuel rod behaviour during LOCA

    International Nuclear Information System (INIS)

    Chakraborty, A.K.; Schubert, J.D.

    1983-01-01

    For calculation of clad ballooning from single rod and rod bundle experiments a model considering the influences of azimuthal temperature gradients due to the existing eccentricity of the pellets has been developed. This model is based on the secondary creep model of Norton and on the concentric deformation model ending in cladding burst as proposed by F. Erbacher. The new model considers the azimuthal temperature differences along the cladding and the resulting differences in deformations. With this model, calculations of cladding burst deformations from single rod and rod bundle experiments are performed with good agreement

  16. 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

  17. 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

  18. A light ion four rod RFQ injector

    International Nuclear Information System (INIS)

    Schempp, A.; Ferch, M.; Klein, H.

    1987-01-01

    The four-rod RFQ has been developed in Frankfurt as an alternative solution for ion injectors. A 202 MHz resonator has been built with design parameters taken from the HERA injector (18keV-750keV, 20mA H - ). Properties of this structure are described and applications as light ion accelerator for particles from an EBIS ion source are discussed

  19. Future plans for performance analysis and maintenance/inspection optimization of shutoff rods based on the case study of Bruce Power Unit-3 Shutoff Rod 5 inspection

    International Nuclear Information System (INIS)

    Nasimi, E.; Gabbar, H.A.

    2011-01-01

    Shutdown System 1 (SDS1) is a preferred method for a quick shutdown of nuclear fission process in CANDU (CANada Deuterium Uranium) reactor units. Failure of a routine SDS1 safety test during Fall 2009 outage resulted in the need to develop and execute a new methodology for Shutoff Rod inspection and re-evaluate the known degradation mechanisms and failure modes. This paper describes the development of this methodology and the obtained results. It also proposes several alternative solutions for the future performance analysis and maintenance/inspection optimization for SDS1 Shutoff Rods based on the Bruce Power Unit-3 Shutoff Rod 5 case study. (author)

  20. PWR control rod ejection analysis with the numerical nuclear reactor

    International Nuclear Information System (INIS)

    Hursin, M.; Kochunas, B.; Downar, T. J.

    2008-01-01

    During the past several years, a comprehensive high fidelity reactor LWR core modeling capability has been developed and is referred to as the Numerical Nuclear Reactor (NNR). The NNR achieves high fidelity by integrating whole core neutron transport solution and ultra fine mesh computational fluid dynamics/heat transfer solution. The work described in this paper is a preliminary demonstration of the ability of NNR to provide a detailed intra pin power distribution during a control rod ejection accident. The motivation of the work is to quantify the impact on the fuel performance calculation of a more physically accurate representation of the power distribution within the fuel rod during the transient. The paper addresses first, the validation of the transient capability of the neutronic module of the NNR code system, DeCART. For this purpose, a 'mini core' problem consisting of a 3x3 array of typical PWR fuel assemblies is considered. The initial state of the 'mini core' is hot zero power with a control rod partially inserted into the central assembly which is fresh fuel and is adjacent to once and twice burned fuel representative of a realistic PWR arrangement. The thermal hydraulic feedbacks are provided by a simplified fluids and heat conduction solver consistent for both PARCS and DeCART. The control rod is ejected from the central assembly and the transient calculation is performed with DeCART and compared with the results of the U.S. NRC core simulation code PARCS. Because the pin power reconstruction in PARCS is based on steady state intra assembly pin power distributions which do not account for thermal feedback during the transient and which do not take into account neutron leakage from neighboring assemblies during the transient, there are some small differences in the PARCS and DeCART pin power prediction. Intra pin power density information obtained with DeCART represents new information not available with previous generation of methods. The paper then

  1. 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

  2. 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.

  3. Process and device for identifying nuclear reactor neutron absorber rod etancheity defect

    International Nuclear Information System (INIS)

    Pelletier, J.; Parrat, D.

    1990-01-01

    For identifying defects in the sealing of neutron absorbing rods. The rod is placed in a pressure tight enclosure filled with a chemically agressive solution. After a time the pressure is released to allow the solution come out of the rod. An analysis of the solution allows the detection of radioactive isotopes of metals which are in the rod [fr

  4. 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

  5. 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,

  6. 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

  7. 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

  8. A comparison of macroscopic models describing the collective response of sedimenting rod-like particles in shear flows

    KAUST Repository

    Helzel, Christiane; Tzavaras, Athanasios

    2016-01-01

    We consider a kinetic model, which describes the sedimentation of rod-like particles in dilute suspensions under the influence of gravity, presented in Helzel and Tzavaras (submitted for publication). Here we restrict our considerations to shear flow and consider a simplified situation, where the particle orientation is restricted to the plane spanned by the direction of shear and the direction of gravity. For this simplified kinetic model we carry out a linear stability analysis and we derive two different nonlinear macroscopic models which describe the formation of clusters of higher particle density. One of these macroscopic models is based on a diffusive scaling, the other one is based on a so-called quasi-dynamic approximation. Numerical computations, which compare the predictions of the macroscopic models with the kinetic model, complete our presentation.

  9. A comparison of macroscopic models describing the collective response of sedimenting rod-like particles in shear flows

    KAUST Repository

    Helzel, Christiane

    2016-07-22

    We consider a kinetic model, which describes the sedimentation of rod-like particles in dilute suspensions under the influence of gravity, presented in Helzel and Tzavaras (submitted for publication). Here we restrict our considerations to shear flow and consider a simplified situation, where the particle orientation is restricted to the plane spanned by the direction of shear and the direction of gravity. For this simplified kinetic model we carry out a linear stability analysis and we derive two different nonlinear macroscopic models which describe the formation of clusters of higher particle density. One of these macroscopic models is based on a diffusive scaling, the other one is based on a so-called quasi-dynamic approximation. Numerical computations, which compare the predictions of the macroscopic models with the kinetic model, complete our presentation.

  10. 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.)

  11. 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.

  12. 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)

  13. 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

  14. Describing function theory as applied to thermal and neutronic problems

    International Nuclear Information System (INIS)

    Nassersharif, B.

    1983-01-01

    Describing functions have traditionally been used to obtain the solutions of systems of ordinary differential equations. In this work the describing function concept has been extended to include nonlinear, distributed parameter partial differential equations. A three-stage solution algorithm is presented which can be applied to any nonlinear partial differential equation. Two generalized integral transforms were developed as the T-transform for the time domain and the B-transform for the spatial domain. The thermal diffusion describing function (TDDF) is developed for conduction of heat in solids and a general iterative solution along with convergence criteria is presented. The proposed solution method is used to solve the problem of heat transfer in nuclear fuel rods with annular fuel pellets. As a special instance the solid cylindrical fuel pellet is examined. A computer program is written which uses the describing function concept for computing fuel pin temperatures in the radial direction during reactor transients. The second problem investigated was the neutron diffusion equation which is intrinsically different from the first case. Although, for most situations, it can be treated as a linear differential equation, the describing function method is still applicable. A describing function solution is derived for two possible cases: constant diffusion coefficient and variable diffusion coefficient. Two classes of describing functions are defined for each case which portray the leakage and absorption phenomena. For the specific case of a slab reactor criticality problem the comparison between analytical and describing function solutions revealed an excellent agreement

  15. 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.)

  16. 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.

  17. 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

  18. New soliton solution to the longitudinal wave equation in a magneto-electro-elastic circular rod

    Science.gov (United States)

    Seadawy, Aly R.; Manafian, Jalil

    2018-03-01

    This paper examines the effectiveness of an integration scheme which called the extended trial equation method (ETEM) in exactly solving a well-known nonlinear equation of partial differential equations (PDEs). In this respect, the longitudinal wave equation (LWE) that arises in mathematical physics with dispersion caused by the transverse Poisson's effect in a magneto-electro-elastic (MEE) circular rod, which a series of exact traveling wave solutions for the aforementioned equation is formally extracted. Explicit new exact solutions are derived in different form such as dark solitons, bright solitons, solitary wave, periodic solitary wave, rational function, and elliptic function solutions of the longitudinal wave equation. The movements of obtained solutions are shown graphically, which helps to understand the physical phenomena of this longitudinal wave equation. Many other such types of nonlinear equations arising in non-destructive evaluation of structures made of the advanced MEE material can also be solved by this method.

  19. 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

  20. 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

  1. Heat flux and temperature determination on the control rod outer surface

    International Nuclear Information System (INIS)

    Taler, J.; Cebula, A.; Marcinkiewicz, J.; Tinoco, H.

    2011-01-01

    The paper presents heat transfer calculation results concerning a control rod of Unit 3 of Forsmark Nuclear Power Plant (NPP). The part of the control rod, which is the object of interest, is surrounded by a mixing region of hot and cold flows and, as a consequence, is subjected to thermal fluctuations. The paper describes a numerical test which validates the method based on the solution of the inverse heat conduction problem (IHCP). The comparison of the results achieved by two methods, CFD and IHCP, including a description of the IHCP method used in the calculation process, shows a very good agreement between the methods. (author)

  2. 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.

  3. Rod drive and latching mechanism

    International Nuclear Information System (INIS)

    Veronesi, L.; Sherwood, D.G.

    1982-01-01

    Hydraulic drive and latching mechanisms for driving reactivity control mechanisms in nuclear reactors are described. Preferably, the pressurized reactor coolant is utilized to raise the drive rod into contact with and to pivot the latching mechanism so as to allow the drive rod to pass the latching mechanism. The pressure in the housing may then be equalized which allows the drive rod to move downwardly into contact with the latching mechanism but to hold the shaft in a raised position with respect to the reactor core. Once again, the reactor coolant pressure may be utilized to raise the drive rod and thus pivot the latching mechanism so that the drive rod passes above the latching mechanism. Again, the mechanism pressure can be equalized which allows the drive rod to fall and pass by the latching mechanism so that the drive rod approaches the reactor core. (author)

  4. Rayleigh-Love model of longitudinal vibrations of conical and exponential rods: Exact solutions and numerical simulation by the method of lines

    CSIR Research Space (South Africa)

    Shatalov, M

    2011-07-01

    Full Text Available of conical surface of the rod is described by equation ( ) ( )pr x k x x kx= ? = , where px is coordinate of the pole of the cone, px x x= ? , then ( ) 2 2S x k xpi= , ( ) 4 4 2pI x k xpi= and equation (2) is rewritten as follows: ( ) ( ) ( ) ( ) ( ) 2... rod in accordance with the classi- cal theory, and 2 k c ? ?? = is the wavenumber of the conical rod which has dimension 1m? . Introducing new dimensionless variable z x?= , considering new function ( ) zV z U ? ? ? = ? ?? ? ( )W z z...

  5. 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....

  6. 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

  7. 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.)

  8. 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

  9. 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

  10. Hydraulically driven control rod concept for integral reactors: fluid dynamic simulation and preliminary test

    International Nuclear Information System (INIS)

    Ricotti, M.E.; Cammi, A.; Lombardi, C.; Passoni, M.; Rizzo, C.; Carelli, M.; Colombo, E.

    2003-01-01

    The paper deals with the preliminary study of the Hydraulically Driven Control Rod concept, tailored for PWR control rods (spider type) with hydraulic drive mechanism completely immersed in the primary water. A specific solution suitable for advanced versions of the IRIS integral reactor is under investigation. The configuration of the Hydraulic Control Rod device, made up by an external movable piston and an internal fixed cylinder, is described. After a brief description of the whole control system, particular attention is devoted to the Control Rod characterization via Computational Fluid Dynamics (CFD) analysis. The investigation of the system behavior, including dynamic equilibrium and stability properties, has been carried out. Finally, preliminary tests were performed in a low pressure, low temperature, reduced length experimental facility. The results are compared with the dynamic control model and CFD simulation model, showing good agreement between simulations and experimental data. During these preliminary tests, the control system performs correctly, allowing stable dynamic equilibrium positions for the Control Rod and stable behavior during withdrawal and insertion steps. (author)

  11. 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

  12. 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.

  13. 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

  14. 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

  15. Verification of heat flux and temperature calculation on the control rod outer surface

    Science.gov (United States)

    Taler, Jan; Cebula, Artur

    2011-12-01

    The paper presents heat transfer calculation results concerning a control rod of Forsmark Nuclear Power Plant (NPP). The part of the control rod, which is the object of interest, is surrounded by a mixing region of hot and cold flows and, as a consequence, is subjected to thermal fluctuations. The paper describes a numerical test which validates the method based on the solution of the inverse heat conduction problem (IHCP). The comparison of the results achieved by two methods, computational fluid dynamics (CFD) simulations and IHCP, including a description of the IHCP method used in the calculation process, shows a very good agreement between the methods.

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

    International Nuclear Information System (INIS)

    Thomas, J.K.

    1992-02-01

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

  17. 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)

  18. FEMAXI-III, a computer code for fuel rod performance analysis

    International Nuclear Information System (INIS)

    Ito, K.; Iwano, Y.; Ichikawa, M.; Okubo, T.

    1983-01-01

    This paper presents a method of fuel rod thermal-mechanical performance analysis used in the FEMAXI-III code. The code incorporates the models describing thermal-mechanical processes such as pellet-cladding thermal expansion, pellet irradiation swelling, densification, relocation and fission gas release as they affect pellet-cladding gap thermal conductance. The code performs the thermal behavior analysis of a full-length fuel rod within the framework of one-dimensional multi-zone modeling. The mechanical effects including ridge deformation is rigorously analyzed by applying the axisymmetric finite element method. The finite element geometrical model is confined to a half-pellet-height region with the assumption that pellet-pellet interaction is symmetrical. The 8-node quadratic isoparametric ring elements are adopted for obtaining accurate finite element solutions. The Newton-Raphson iteration with an implicit algorithm is applied to perform the analysis of non-linear material behaviors accurately and stably. The pellet-cladding interaction mechanism is exactly treated using the nodal continuity conditions. The code is applicable to the thermal-mechanical analysis of water reactor fuel rods experiencing variable power histories. (orig.)

  19. Mathematical model of an integrated circuit cooling through cylindrical rods

    Directory of Open Access Journals (Sweden)

    Beltrán-Prieto Luis Antonio

    2017-01-01

    Full Text Available One of the main challenges in integrated circuits development is to propose alternatives to handle the extreme heat generated by high frequency of electrons moving in a reduced space that cause overheating and reduce the lifespan of the device. The use of cooling fins offers an alternative to enhance the heat transfer using combined a conduction-convection systems. Mathematical model of such process is important for parametric design and also to gain information about temperature distribution along the surface of the transistor. In this paper, we aim to obtain the equations for heat transfer along the chip and the fin by performing energy balance and heat transfer by conduction from the chip to the rod, followed by dissipation to the surrounding by convection. Newton's law of cooling and Fourier law were used to obtain the equations that describe the profile temperature in the rod and the surface of the chip. Ordinary differential equations were obtained and the respective analytical solutions were derived after consideration of boundary conditions. The temperature along the rod decreased considerably from the initial temperature (in contatct with the chip surface. This indicates the benefit of using a cilindrical rod to distribute the heat generated in the chip.

  20. Components inspection of Monju, a sodium bonded type control rod

    International Nuclear Information System (INIS)

    Harada, Kiyoshi; Matsushita, Yuichi; Lee, Chunchan; Abe, Hideaki; Watahiki, Naohisa

    2002-03-01

    This Report addresses a result of a sodium test conducted on components of a Double Poral Filter Sodium Bonded Type Control Rod that is expected to be a next generation, long life Control Rod. Upper and lower Poral Filter Sodium Bonded Type Control Rod components were mocked up to conduct a sodium test. During the test, sodium chargeability, formation of Gas Plenum at the upper part of the components, sodium drain-ability and NaOH clean-ability were recognized under actual plant condition. The following are results obtained: (1) Sodium Chargeability at Control Rod Insertion to EVST. Sodium was charged into the components when the mocked-up was inserted in sodium of 190degC, with insertion speed of 6 m/min which is an actual insertion speed to EVST. (2) Formation of Upper Gas Plenum by Helium Gas generated in Control Rod Components Gas Plenum formation within deviation of 9% was confirmed by releasing helium gas into the mocked-up which is immersed in sodium of 620degC and 190degC. Length of Gas Plenum is confirmed to be retained in certain length even if helium gas is further released into formed Gas Plenum. (3) Sodium Drain-ability of Control Rod Components when Drawing from EVST. Drain-ability was confirmed to be sufficient and no sodium residue was found in the mocked-up when the mocked-up was drawn out from sodium of 190degC, with drawing speed of 6 m/min which is an actual drawing speed from EVST. (4) Clean-ability of NaOH Solution against Sodium Residue in Control Rod Components. Sodium and NaOH solution reacted calmly, however, clean-ability was not sufficient. When Sodium fully remained in Control Rod Components, it made circulation of NaOH solution not enough. (author)

  1. Thermodynamic analysis of unimer-micelle and sphere-to-rod micellar transitions of aqueous solutions of sodium dodecylbenzenesulfonate

    International Nuclear Information System (INIS)

    Valente, Artur J.M.; López Cascales, J.J.; Fernández Romero, Antonio J.

    2014-01-01

    Highlights: • Unimer-micelle and sphere-to-rod micellar transitions were observed to sodium dodecylbenzenesulfonate in aqueous solutions. • Two micellar transitions were seen by electrical conductivity and surface tension. • An anomalous ΔS 0 and ΔH 0 increase with T was found for the second critical transition. • More stable aggregates are evidenced for spherical micelles than for the other shapes. - Abstract: Temperature dependence of specific conductivity of sodium dodecylbenzenesulfonate (NaDBS) aqueous solutions was analyzed. Two breaks on the plot appeared for all temperature, which suggest two micellar transitions. This has been corroborated by surface tension measurements. The first transition concentration occurs at the critical micelle concentration (CMC), whilst the second critical concentration (so-called transition micellar concentration, TMC) is due to a sphere-to-rod micelles transition. The dependence of CMC and TMC on the temperature allows the computation of the corresponding thermodynamic functions: Gibbs free energy, enthalpy and entropy changes. For the CMC, enthalpy and entropy increments were found that decrease with the temperature values. However, an anomalous behavior was obtained for the TMC, where both ΔS 0 and ΔH 0 values raised with the temperature increase. However, for both transitions, an (enthalpy + entropy) compensation is observed. These results will be compared with similar systems reported in the literature

  2. 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

  3. 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

  4. 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

  5. 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

  6. 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

  7. 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.

  8. 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

  9. 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.)

  10. 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.)

  11. 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.

  12. 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

  13. 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

  14. 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

  15. Interactions Between Charged Macroions Mediated by Molecules with Rod-like Charged Structures

    Directory of Open Access Journals (Sweden)

    Bohinc, K.

    2014-03-01

    Full Text Available A short review of recent theoretical advances in studies of the interaction between highly charged systems embedded in a solution of rod-like molecules is presented. The system is theoretically described by the functional density theory, where the correlations within the rod-like molecules are accounted for. We show that for sufficiently long molecules and large surface charge densities, an attractive force between like-charged surfaces arises due to the spatially distributed charges within the molecules. The added salt has an influence on the condition for the attractive force between like-charged surfaces. The theoretical results are compared with Monte Carlo simulations. Many phenomena motivate the study of the interaction between like-charged surfaces (DNA condensation, virus aggregation, yeast flocculation, cohesion of cement paste.

  16. 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.

  17. 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

  18. 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

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

    International Nuclear Information System (INIS)

    Rector, D.R.

    1987-10-01

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

  20. 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

  1. 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

  2. Stress-strain state analysis and optimization of rod system under periodic pulse load

    Directory of Open Access Journals (Sweden)

    Grebenyuk Grigory

    2018-01-01

    Full Text Available The paper considers the problem of analysis and optimization of rod systems subjected to combined static and periodic pulse load. As a result of the study the analysis method was developed based on traditional approach to solving homogeneous matrix equations of state and a special algorithm for developing a particular solution. The influence of pulse parameters variations on stress-strain state of a rod system was analyzed. Algorithms for rod systems optimization were developed basing on strength recalculation and statement and solution of optimization problem as a problem of nonlinear mathematical programming. Recommendations are developed for efficient organization of process for optimization of rod systems under static and periodic pulse load.

  3. Modeling and Experimental Tests on the Hydraulically Driven Control Rod option for IRIS Reactor

    International Nuclear Information System (INIS)

    Cammi, Antonio; Ricotti, Marco E.; Vitulo, Alessia

    2004-01-01

    The adoption of Internal Control Rod Drive Mechanisms (ICRDMs) represents a valuable alternative to classical, external CRDMs based on electro-magnetic devices, as adopted in current PWRs. The advantages on the safety features of the reactor are apparent: inherent elimination of the Rod Ejection accidents and of possible concerns about the vessel head penetrations. A further positive feedback on the design is the reduction of the primary system overall dimensions. Within the frame of the ICRDM concepts, the Hydraulically Driven Control Rod solution is investigated as a possible option for the IRIS integral reactor. After a brief comparison of the solutions currently proposed for integral reactors, the configuration of the Hydraulic Control Rod device for IRIS, made up by an external movable piston and an internal fixed cylinder, is described. A description of the whole control system is reported as well. Particular attention is devoted to the Control Rod profile characterization, performed by means of a Computational Fluid Dynamics (CFD) analysis. The investigation of the system behavior has been carried out, including the dynamic equilibrium and its stability properties, the withdrawal and insertion step movement and the sensitivity study on command time periods. A suitable dynamic model has been set up for the mentioned purposes: the models corresponding to the various Control Rod system devices have been written in an Object-Oriented language (Modelica), thus allowing an easy implementation of such a system into the simulator for the whole reactor. Finally, a preliminary low pressure, low temperature, reduced length experimental facility has been built. Tests on HDCR stability and operational transients have been performed. The results are compared with the dynamic system model and CFD simulation model, showing good agreement between simulations and experimental data. During these preliminary tests, the control system performed correctly, allowing stable dynamic

  4. Anisotropic generalization of well-known solutions describing relativistic self-gravitating fluid systems. An algorithm

    Energy Technology Data Exchange (ETDEWEB)

    Thirukkanesh, S. [Eastern University, Department of Mathematics, Chenkalady (Sri Lanka); Ragel, F.C. [Eastern University, Department of Physics, Chenkalady (Sri Lanka); Sharma, Ranjan; Das, Shyam [P.D. Women' s College, Department of Physics, Jalpaiguri (India)

    2018-01-15

    We present an algorithm to generalize a plethora of well-known solutions to Einstein field equations describing spherically symmetric relativistic fluid spheres by relaxing the pressure isotropy condition on the system. By suitably fixing the model parameters in our formulation, we generate closed-form solutions which may be treated as an anisotropic generalization of a large class of solutions describing isotropic fluid spheres. From the resultant solutions, a particular solution is taken up to show its physical acceptability. Making use of the current estimate of mass and radius of a known pulsar, the effects of anisotropic stress on the gross physical behaviour of a relativistic compact star is also highlighted. (orig.)

  5. Solutions to the equations describing materials with competing quadratic and cubic nonlinearities

    International Nuclear Information System (INIS)

    Li-Na, Zhao; Ji, Lin; Zi-Shuang, Tong

    2009-01-01

    The Lie group theoretical method is used to study the equations describing materials with competing quadratic and cubic nonlinearities. The equations share some of the nice properties of soliton equations. From the elliptic functions expansion method, we obtain large families of analytical solutions, in special cases, we have the periodic, kink and solitary solutions of the equations. Furthermore, we investigate the stability of these solutions under the perturbation of amplitude noises by numerical simulation

  6. 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.)

  7. 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

  8. 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.)

  9. 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

  10. On the General Analytical Solution of the Kinematic Cosserat Equations

    KAUST Repository

    Michels, Dominik L.

    2016-09-01

    Based on a Lie symmetry analysis, we construct a closed form solution to the kinematic part of the (partial differential) Cosserat equations describing the mechanical behavior of elastic rods. The solution depends on two arbitrary analytical vector functions and is analytical everywhere except a certain domain of the independent variables in which one of the arbitrary vector functions satisfies a simple explicitly given algebraic relation. As our main theoretical result, in addition to the construction of the solution, we proof its generality. Based on this observation, a hybrid semi-analytical solver for highly viscous two-way coupled fluid-rod problems is developed which allows for the interactive high-fidelity simulations of flagellated microswimmers as a result of a substantial reduction of the numerical stiffness.

  11. On the General Analytical Solution of the Kinematic Cosserat Equations

    KAUST Repository

    Michels, Dominik L.; Lyakhov, Dmitry; Gerdt, Vladimir P.; Hossain, Zahid; Riedel-Kruse, Ingmar H.; Weber, Andreas G.

    2016-01-01

    Based on a Lie symmetry analysis, we construct a closed form solution to the kinematic part of the (partial differential) Cosserat equations describing the mechanical behavior of elastic rods. The solution depends on two arbitrary analytical vector functions and is analytical everywhere except a certain domain of the independent variables in which one of the arbitrary vector functions satisfies a simple explicitly given algebraic relation. As our main theoretical result, in addition to the construction of the solution, we proof its generality. Based on this observation, a hybrid semi-analytical solver for highly viscous two-way coupled fluid-rod problems is developed which allows for the interactive high-fidelity simulations of flagellated microswimmers as a result of a substantial reduction of the numerical stiffness.

  12. 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

  13. 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

  14. 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.

  15. The BWR core simulator COSIMA with 2 group nodal flux expansion and control rod history

    International Nuclear Information System (INIS)

    Hoejerup, C.F.

    1989-08-01

    The boiling water simulator NOTAM has been modified and improved in several aspects: - The ''1 1/2'' energy group TRILUX nodal flux solution method has been exchanged with a 2 group modal expansion method. - Control rod ''history'' has been introduced. - Precalculated instrument factors have been introduced. The paper describes these improvements, which were considered sufficiently large to justify a new name to the programme: COSIMA. (author)

  16. 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

  17. 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)

  18. SCORE-EVET: a computer code for the multidimensional transient thermal-hydraulic analysis of nuclear fuel rod arrays

    International Nuclear Information System (INIS)

    Benedetti, R.L.; Lords, L.V.; Kiser, D.M.

    1978-02-01

    The SCORE-EVET code was developed to study multidimensional transient fluid flow in nuclear reactor fuel rod arrays. The conservation equations used were derived by volume averaging the transient compressible three-dimensional local continuum equations in Cartesian coordinates. No assumptions associated with subchannel flow have been incorporated into the derivation of the conservation equations. In addition to the three-dimensional fluid flow equations, the SCORE-EVET code ocntains: (a) a one-dimensional steady state solution scheme to initialize the flow field, (b) steady state and transient fuel rod conduction models, and (c) comprehensive correlation packages to describe fluid-to-fuel rod interfacial energy and momentum exchange. Velocity and pressure boundary conditions can be specified as a function of time and space to model reactor transient conditions such as a hypothesized loss-of-coolant accident (LOCA) or flow blockage

  19. 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.)

  20. 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

  1. 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

  2. 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

  3. 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.)

  4. 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

  5. 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.)

  6. Electromagnetic methods for measuring materials properties of cylindrical rods and array probes for rapid flaw inspection

    Energy Technology Data Exchange (ETDEWEB)

    Sun, Haiyan [Iowa State Univ., Ames, IA (United States)

    2005-01-01

    The case-hardening process modifies the near-surface permeability and conductivity of steel, as can be observed through changes in alternating current potential drop (ACPD) along a rod. In order to evaluate case depth of case hardened steel rods, analytical expressions are derived for the alternating current potential drop on the surface of a homogeneous rod, a two-layered and a three-layered rod. The case-hardened rod is first modeled by a two-layer rod that has a homogeneous substrate with a single, uniformly thick, homogeneous surface layer, in which the conductivity and permeability values differ from those in the substrate. By fitting model results to multi-frequency ACPD experimental data, estimates of conductivity, permeability and case depth are found. Although the estimated case depth by the two-layer model is in reasonable agreement with the effective case depth from the hardness profile, it is consistently higher than the effective case depth. This led to the development of the three-layer model. It is anticipated that the new three-layered model will improve the results and thus makes the ACPD method a novel technique in nondestructive measurement of case depth. Another way to evaluate case depth of a case hardened steel rod is to use induction coils. Integral form solutions for an infinite rod encircled by a coaxial coil are well known, but for a finite length conductor, additional boundary conditions must be satisfied at the ends. In this work, calculations of eddy currents are performed for a two-layer conducting rod of finite length excited by a coaxial circular coil carrying an alternating current. The solution is found using the truncated region eigenfunction expansion (TREE) method. By truncating the solution region to a finite length in the axial direction, the magnetic vector potential can be expressed as a series expansion of orthogonal eigenfunctions instead of as a Fourier integral. Closed-form expressions are derived for the electromagnetic

  7. 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.

  8. 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

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

    Science.gov (United States)

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

    2016-01-01

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

  10. New interior solution describing relativistic fluid sphere

    Indian Academy of Sciences (India)

    Anewexact solution of embedding class I is presented for a relativistic anisotropicmassive fluid sphere. The new exact solution satisfies Karmarkar condition, is well-behaved in all respects, and therefore is suitable for the modelling of superdense stars. Consequently, using this solution, we have studied in detail two ...

  11. 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.

  12. 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

  13. Solitary waves on nonlinear elastic rods. I

    DEFF Research Database (Denmark)

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

    1984-01-01

    Acoustic waves on elastic rods with circular cross section are governed by improved Boussinesq equations when transverse motion and nonlinearity in the elastic medium are taken into account. Solitary wave solutions to these equations have been found. The present paper treats the interaction betwe...... nonlinearity. The balance between dispersion and nonlinearity in the equation is investigated.......Acoustic waves on elastic rods with circular cross section are governed by improved Boussinesq equations when transverse motion and nonlinearity in the elastic medium are taken into account. Solitary wave solutions to these equations have been found. The present paper treats the interaction between...... the solitary waves numerically. It is demonstrated that the waves behave almost like solitons in agreement with the fact that the improved Boussinesq equations are nearly integrable. Thus three conservation theorems can be derived from the equations. A new subsonic quasibreather is found in the case of a cubic...

  14. 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

  15. Measurement of the fission ratio for several configurations of uranium oxide rod clusters

    International Nuclear Information System (INIS)

    Pattenden, S.K.; Patterson, C.R.

    1962-02-01

    This report describes measurements of the fission ratio for a single fuel channel of oxide rod clusters in an essentially infinite block of graphite. The measurements were made using the 'catcher-foil' technique, the activities of the catcher foils being measured by β-counting. Results are given, for 37-rod; 18-rod; 7-rod and 3-rod clusters, and are compared with theoretical predictions. (author)

  16. 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)

  17. Coexistence of spheres and rods in micellar solution of dodecyldimethylamine oxide

    NARCIS (Netherlands)

    Majhi, P.R.; Dubin, P.L.; Feng, X.H.; Ruo, X.H.; Leermakers, F.A.M.; Tribet, C.

    2004-01-01

    Micelles of dimethyldodecylamine oxide (DMDAO) are known to exhibit sphere-to-rod transitions as a function of pH and ionic strength. Long micelles are stabilized at pH corresponding to half-protonation, because hydrogen bonding between nonionic and protonated monomers yields an effectively

  18. SCORE-EVET: a computer code for the multidimensional transient thermal-hydraulic analysis of nuclear fuel rod arrays. [BWR; PWR

    Energy Technology Data Exchange (ETDEWEB)

    Benedetti, R. L.; Lords, L. V.; Kiser, D. M.

    1978-02-01

    The SCORE-EVET code was developed to study multidimensional transient fluid flow in nuclear reactor fuel rod arrays. The conservation equations used were derived by volume averaging the transient compressible three-dimensional local continuum equations in Cartesian coordinates. No assumptions associated with subchannel flow have been incorporated into the derivation of the conservation equations. In addition to the three-dimensional fluid flow equations, the SCORE-EVET code ocntains: (a) a one-dimensional steady state solution scheme to initialize the flow field, (b) steady state and transient fuel rod conduction models, and (c) comprehensive correlation packages to describe fluid-to-fuel rod interfacial energy and momentum exchange. Velocity and pressure boundary conditions can be specified as a function of time and space to model reactor transient conditions such as a hypothesized loss-of-coolant accident (LOCA) or flow blockage.

  19. Fabrication and characterization of terahertz anisotropic anti-rod dimer planar metamaterials

    DEFF Research Database (Denmark)

    Zalkovskij, Maksim; Malureanu, Radu; Novitsky, Andrey

    2012-01-01

    In this work we describe the fabrication and characterization of free-standing membranes with thick anti-rod dimers metamaterials for terahertz waves. Two different designs with parallel and V-shape anti-rods were analysed. Even though both structures consists of simple elements, namely anti......-rod dimers, they reveal interesting birefringent and dichroic transmission properties....

  20. 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)

  1. 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

  2. Model of cooling nuclear fuel rod in the nuclear reactor

    International Nuclear Information System (INIS)

    Lavicka, David; Polansky, Jiri

    2010-01-01

    The following topics are described: Some basic requirements for nuclear fuel rods; The VVER 1000 fuel rod; Classification of the two-phase flow in the vertical tube; Type of heat transfer crisis in the vertical tube; Experimental apparatus; Model of the nuclear fuel rod and spacers; Potential of the experimental apparatus (velocity profile measurement via PIV; thermal flow field measurement by the PLIF method; cooling graph in dependence on the fuel rod temperature; comparison of the hydrodynamic properties with respect to the design features of the spacers). (P.A.)

  3. The buckling of fuel rods in transportation casks under hypothetical accident conditions

    International Nuclear Information System (INIS)

    Bjorkman, G.S.

    2004-01-01

    The buckling analysis of fuel rods during an end drop impact of a spent fuel transportation cask has traditionally been performed to demonstrate the structural integrity of the fuel rod cladding or the integrity of the fuel geometry in criticality evaluations following a cask drop event. The actual calculation of the fuel rod buckling load, however, has been the subject of some controversy, with estimates of the critical buckling load differing by as much as a factor of 5. Typically, in the buckling analysis of a fuel rod, assumptions are made regarding the percentage of fuel mass that is bonded to or participates with the cladding during the buckling process, with estimates ranging from 0 to 100%. The greater the percentage of fuel mass that is assumed to be bonded to the cladding the higher the inertia loads on the cladding, and, therefore, the lower the ''g'' value at which buckling occurs. Current published solutions do not consider displacement compatibility between the fuel and the cladding. By invoking displacement compatibility between the fuel column and the cladding column, this paper presents an exact solution for the buckling of fuel rods under inertia loading. The results show that the critical inertia load magnitude for the buckling of a fuel rod depends on the weight of the cladding and the total weight of the fuel, regardless of the percentage of fuel mass that is assumed to be attached to or participate with the cladding in the buckling process. Therefore, 100% of the fuel always participates in the buckling of a fuel rod under inertia loading

  4. 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

  5. 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.)

  6. Synthesis and characterization of nano ZnO rods via microwave assisted chemical precipitation method

    Energy Technology Data Exchange (ETDEWEB)

    Uma Sangari, N., E-mail: umasangariselvakumar@gmail.com [Department of Chemistry, S.F.R. College for Women, Sivakasi 626123 (India); Chitra Devi, S. [Department of Chemistry, S.F.R. College for Women, Sivakasi 626123 (India)

    2013-01-15

    A microwave assisted chemical precipitation method has been employed for the synthesis of nano zinc oxide rods by reacting zinc nitrate and potassium hydroxide. The amount of potassium hydroxide was adjusted for three different pHs to achieve ZnO nano rods with varying aspect ratio. The mechanism of growth of nano rods is explained briefly. The average crystallite size of the as synthesized samples was analyzed by means of powder XRD pattern and estimated to vary from 25.6 nm to 43.1 nm. The existence of rods was confirmed using scanning electron microscopy (SEM). The samples were also analyzed using FT-IR. The optical properties of the samples were also studied by means of UV-visible spectra and Room Temperature Photo Luminescence studies. The band gap of the samples was determined from the DRS spectrum. A strong near band emission peaks due to surface defects are observed in the PL spectrum. - Graphical abstract: At the solution pH of 11 and 9, tetrapod-like and flower-like ZnO nano rods were formed along with separated rods respectively due to the formation of activated nuclei of different sizes. Highlights: Black-Right-Pointing-Pointer Increase in alkalinity of the precursor solution results in longer rods. Black-Right-Pointing-Pointer Beyond a saturation limit, the excess of added OH{sup -} ions inhibited the growth of rods. Black-Right-Pointing-Pointer Keeping all parameters the same, the alkalinity can only modify the aspect ratio of the rods and not their morphology.

  7. 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.)

  8. Stability of exact solutions describing two-layer flows with evaporation at the interface

    Energy Technology Data Exchange (ETDEWEB)

    Bekezhanova, V B [Institute of Computational Modelling SB RAS, Akademgorodok, 50/44, Krasnoyarsk, 660036 (Russian Federation); Goncharova, O N, E-mail: bekezhanova@mail.ru, E-mail: gon@math.asu.ru [Altai State University, Lenina 61, Barnaul, 656049 (Russian Federation)

    2016-12-15

    A new exact solution of the equations of free convection has been constructed in the framework of the Oberbeck–Boussinesq approximation of the Navier–Stokes equations. The solution describes the joint flow of an evaporating viscous heat-conducting liquid and gas-vapor mixture in a horizontal channel. In the gas phase the Dufour and Soret effects are taken into account. The consideration of the exact solution allows one to describe different classes of flows depending on the values of the problem parameters and boundary conditions for the vapor concentration. A classification of solutions and results of the solution analysis are presented. The effects of the external disturbing influences (of the liquid flow rates and longitudinal gradients of temperature on the channel walls) on the stability characteristics have been numerically studied for the system HFE7100-nitrogen in the common case, when the longitudinal temperature gradients on the boundaries of the channel are not equal. In the system both monotonic and oscillatory modes can be formed, which damp or grow depending on the values of the initial perturbations, flow rates and temperature gradients. Hydrodynamic perturbations are most dangerous under large gas flow rates. The increasing oscillatory perturbations are developed due to the thermocapillary effect under large longitudinal gradients of temperature. The typical forms of the disturbances are shown. (paper)

  9. Validation of Westinghouse integrated code POLCA-T against OECD NEACRP-L-335 rod ejection benchmark

    Energy Technology Data Exchange (ETDEWEB)

    Panayotov, Dobromir [Westinghouse Electric Sweden AB, Vaesteraas, SE-721 63 (Sweden)

    2008-07-01

    This paper describes the work performed and results obtained in the validation of the POLCA-T code against NEACRP PWR Rod Ejection Transients Benchmark. Presented work is a part of the POLCA-T licensing Assessment Data Base for BWR Control Rod Drop Accident (CRDA) Application. The validation against a PWR Rod Ejection Accidents (REA) Benchmark is relevant for the validation of the code for BWR CRDA, as the analyses of both transients require identical phenomena to be modelled. All six benchmark cases have been analyzed in the presented work. Initial state steady-state calculations including boron search, control rod worth, and final state power search have been performed by POLCA7 code. Initial state boron adjustment and steady-state CR worth as well as the transient analyses were performed by POLCA-T code. Benchmark results including 3D transient power distributions are compared with reference PANTHER solutions and published results of other codes. Given the similarity of the kinetics modelling for a BWR CRDA and a PWR REA and the fact that POLCA-T accurately predicts the local transient power and thus, the resulting fuel enthalpy, it is concluded that POLCA-T is a state-of-art tool also for BWR CRDA analysis. (author)

  10. Validation of Westinghouse integrated code POLCA-T against OECD NEACRP-L-335 rod ejection benchmark

    International Nuclear Information System (INIS)

    Panayotov, Dobromir

    2008-01-01

    This paper describes the work performed and results obtained in the validation of the POLCA-T code against NEACRP PWR Rod Ejection Transients Benchmark. Presented work is a part of the POLCA-T licensing Assessment Data Base for BWR Control Rod Drop Accident (CRDA) Application. The validation against a PWR Rod Ejection Accidents (REA) Benchmark is relevant for the validation of the code for BWR CRDA, as the analyses of both transients require identical phenomena to be modelled. All six benchmark cases have been analyzed in the presented work. Initial state steady-state calculations including boron search, control rod worth, and final state power search have been performed by POLCA7 code. Initial state boron adjustment and steady-state CR worth as well as the transient analyses were performed by POLCA-T code. Benchmark results including 3D transient power distributions are compared with reference PANTHER solutions and published results of other codes. Given the similarity of the kinetics modelling for a BWR CRDA and a PWR REA and the fact that POLCA-T accurately predicts the local transient power and thus, the resulting fuel enthalpy, it is concluded that POLCA-T is a state-of-art tool also for BWR CRDA analysis. (author)

  11. DANCOFF-3, Dancoff Correction for Cylindrical Fuel Rod at H2O Gaps and for Fuel Clusters

    International Nuclear Information System (INIS)

    Sauer, A.

    1989-01-01

    1 - Nature of physical problem solved: Calculation of the Dancoff correction for cylindrical fuel rods in square and hexagonal infinite lattices, for fuel element rods near water gaps, and for fuel rod clusters. 2 - Method of solution: Evaluation by direct numerical integration over the moderator region. 3 - Restrictions on the complexity of the problem: For every rod arrangement at most 100 cases with different materials cross- sections

  12. 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.)

  13. Exact solutions of magnetohydrodynamics for describing different structural disturbances in solar wind

    Science.gov (United States)

    Grib, S. A.; Leora, S. N.

    2016-03-01

    We use analytical methods of magnetohydrodynamics to describe the behavior of cosmic plasma. This approach makes it possible to describe different structural fields of disturbances in solar wind: shock waves, direction discontinuities, magnetic clouds and magnetic holes, and their interaction with each other and with the Earth's magnetosphere. We note that the wave problems of solar-terrestrial physics can be efficiently solved by the methods designed for solving classical problems of mathematical physics. We find that the generalized Riemann solution particularly simplifies the consideration of secondary waves in the magnetosheath and makes it possible to describe in detail the classical solutions of boundary value problems. We consider the appearance of a fast compression wave in the Earth's magnetosheath, which is reflected from the magnetosphere and can nonlinearly overturn to generate a back shock wave. We propose a new mechanism for the formation of a plateau with protons of increased density and a magnetic field trough in the magnetosheath due to slow secondary shock waves. Most of our findings are confirmed by direct observations conducted on spacecrafts (WIND, ACE, Geotail, Voyager-2, SDO and others).

  14. Review of FFTF and CRBRP control rod systems designs

    International Nuclear Information System (INIS)

    Pitterle, T.A.; Lagally, H.O.

    1977-01-01

    The evolution of the primary control rod system design for FFTF and CRBR, beginning with the initial choice of the basic concepts, is described. The significant component and systems tests are reviewed together with the test results which referenced the development of the CRBR primary control rod system design. Modifications to the concepts and detail designs of the FFTF control rod system were required principally to satisfy the requirements of CRBR, and at the same time incorporating design refinements shown desirable by the tests

  15. 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.)

  16. 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.)

  17. 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

  18. 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.

  19. Nuclear fuel rod loading apparatus

    International Nuclear Information System (INIS)

    King, H.B.

    1981-01-01

    A nuclear fuel loading apparatus, incorporating a microprocessor control unit, is described which automatically loads nuclear fuel pellets into dual fuel rods with a minimum of manual involvement and in a manner and sequence to ensure quality control and accuracy. (U.K.)

  20. Critical experiments supporting underwater storage of tightly packed configurations of spent fuel rods

    International Nuclear Information System (INIS)

    Hoovler, G.S.; Baldwin, M.N.

    1981-04-01

    Criticla arrays of 2.5%-enriched UO 2 fuel rods that simulate underwater rod storage of spent power reactor fuel are being constructed. Rod storage is a term used to describe a spent fuel storage concept in which the fuel bundles are disassembled and the rods are packed into specially designed cannisters. Rod storage would substantially increase the amount of fuel that could be stored in available space. These experiments are providing criticality data against which to benchmark nuclear codes used to design tightly packed rod storage racks

  1. Embedded cladding surface thermocouples on Zircaloy-sheathed heater rods

    International Nuclear Information System (INIS)

    Wilkins, S.C.

    1977-06-01

    Titanium-sheathed Type K thermocouples embedded in the cladding wall of zircaloy-sheathed heater rods are described. These thermocouples constitute part of a program intended to characterize the uncertainty of measurements made by surface-mounted cladding thermocouples on nuclear fuel rods. Fabrication and installation detail, and laboratory testing of sample thermocouple installations are included

  2. 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

  3. 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

  4. On the Construction and Properties of Weak Solutions Describing Dynamic Cavitation

    KAUST Repository

    Miroshnikov, Alexey

    2014-08-21

    We consider the problem of dynamic cavity formation in isotropic compressible nonlinear elastic media. For the equations of radial elasticity we construct self-similar weak solutions that describe a cavity emanating from a state of uniform deformation. For dimensions d=2,3 we show that cavity formation is necessarily associated with a unique precursor shock. We also study the bifurcation diagram and do a detailed analysis of the singular asymptotics associated to cavity initiation as a function of the cavity speed of the self-similar profiles. We show that for stress free cavities the critical stretching associated with dynamically cavitating solutions coincides with the critical stretching in the bifurcation diagram of equilibrium elasticity. Our analysis treats both stress-free cavities and cavities with contents.

  5. Process development for fabrication of Ag-15% In-5% Cd alloys and rods for the control rods of IPEN critical unit

    International Nuclear Information System (INIS)

    Figueredo, A.M. de.

    1985-12-01

    The development of two process at the Nuclear and Energetic Research Institute (IPEN-Brazil) are described. - the production of Ag-15% In-5%. Cd alloys with nuclear grade. The fabrication of rods from Ag-15% In-5% Cd alloy for use at the critical unit. The methods for quality control of alloy and rod are presented, and main problems are identified. (C.G.C.)

  6. 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.)

  7. Cleaning and excavating tool for control rod canopy seals

    International Nuclear Information System (INIS)

    Kucera, R.A.

    1991-01-01

    This patent describes a device for servicing a weld site about the periphery of a control rod drive mechanism. It comprises a housing adapted to be rotated about the periphery of the control rod drive mechanism; a carriage reciprocably received within the housing; first movement means for reciprocating the carriage in a first direction; a tool attachment reciprocably received within the carriage; and second movement means for reciprocating the tool in a second direction; wherein the tool attachment is positioned relative to the control rod drive mechanism by the first and second movement means

  8. 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

  9. Development and assessment of modular models of calculation for the interpretation of rod-melting experiments

    International Nuclear Information System (INIS)

    Tuerk, W.

    1980-01-01

    By the example of recalculations of rod-melting experiment it is shown how a modular simulation model for complex systems can be formulated within the scope of RSYST1. The procedure of code development as well as the physical and numerical methods and approximations of the simulation model are described. To each important physical process a code module is assigned. The individual moduls describe heat production, rod heat-up, rod oxidation, rod environment, rod deformation by thermal expansion and can buckling, melting of the rod, rod failure, and flowing off of the melted mass. A comparison of the results for the overall model with the result of different experiments indicates that the phenomena during heat-up and melting of the rod are treated in agreement with the experiments. The results of the calculation model and its submodels are thus largely supported by experiments. Therefore further predictions with a high level of confidence can be made with the model within the scope of reactor safety research. (orig.) [de

  10. FEMAXI-III: a computer code for the analysis of thermal and mechanical behavior of fuel rods

    International Nuclear Information System (INIS)

    Nakajima, Tetsuo; Ichikawa, Michio; Iwano, Yoshihiko; Ito, Kenichi; Saito, Hiroaki; Kashima, Koichi; Kinoshita, Motoyasu; Okubo, Tadatsune.

    1985-12-01

    FEMAXI-III is a computer code to predict the thermal and mechanical behavior of a light water fuel rod during its irradiation life. It can analyze the integral behavior of a whole fuel rod throughout its life, as well as the localized behavior of a small part of fuel rod. The localized mechanical behavior such as the cladding ridge deformation is analyzed by the two-dimensional axisymmetric finite element method. FEMAXI-III calculates, in particular, the temperature distribution, the radial deformation, the fission gas release, and the inner gas pressure as a function of irradiation time and axial position, and the stresses and strains in the fuel and cladding at a small part of fuel rod as a function of irradiation time. For this purpose, Elasto-plasticity, creep, thermal expansion, fuel cracking and crack healing, relocation, densification, swelling, hot pressing, heat generation distribution, fission gas release, and fuel-cladding mechanical interaction are modelled and their interconnected effects are considered in the code. Efforts have been made to improve the accuracy and stability of finite element solution and to minimize the computer memory and running time. This report describes the outline of the code and the basic models involved, and also includes the application of the code and its input manual. (author)

  11. 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)

  12. 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.)

  13. 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.)

  14. 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

  15. Temperature distribution on fuel rods: a study on the effect of eccentricity in the position of UO2 pellets

    International Nuclear Information System (INIS)

    Gaspar Junior, Joao Carlos Aguiar

    2010-01-01

    This work proposes the development of a method of solving equations of heat transfer applied in fuel rods using the finite element method, in order to evaluate the performance and safety of the nuclear system. Was prepared in a Fortran program to evaluate the equations governing the problem, the boundary conditions and apply the properties of materials on a steady state. This program uses the mesh generation input and graphical output generated by the program GID. The method was validated against the analytical solution found in the book Todreas and Kazimi with error less than 0.2% and with respect to the improved analytical solution of Nijsing for axisymmetry rod and eccentricity rod with error less than a 3.6%. Applications have been developed with the use of correlations for properties with the temperature dependence of resolution axisymmetry rod and the resolution of a rod with eccentricity. The method developed, should it be implemented, would allow the assessment of fuel rods in the given situations and other scenarios, as well as adding a tool of substantial value in the analysis of rods. (author)

  16. New models of droplet deposition and entrainment for prediction of CHF in cylindrical rod bundles

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Haibin, E-mail: hb-zhang@xjtu.edu.cn [School of Chemical Engineering and Technology, Xi’an Jiaotong University, Xi’an 710049 (China); Department of Chemical Engineering, Imperial College, London SW7 2BY (United Kingdom); Hewitt, G.F. [Department of Chemical Engineering, Imperial College, London SW7 2BY (United Kingdom)

    2016-08-15

    Highlights: • New models of droplet deposition and entrainment in rod bundles is developed. • A new phenomenological model to predict the CHF in rod bundles is described. • The present model is well able to predict CHF in rod bundles. - Abstract: In this paper, we present a new set of model of droplet deposition and entrainment in cylindrical rod bundles based on the previously proposed model for annuli (effectively a “one-rod” bundle) (2016a). These models make it possible to evaluate the differences of the rates of droplet deposition and entrainment for the respective rods and for the outer tube by taking into account the geometrical characteristics of the rod bundles. Using these models, a phenomenological model to predict the CHF (critical heat flux) for upward annular flow in vertical rod bundles is described. The performance of the model is tested against the experimental data of Becker et al. (1964) for CHF in 3-rod and 7-rod bundles. These data include tests in which only the rods were heated and data for simultaneous uniform and non-uniform heating of the rods and the outer tube. It was shown that the predicted CHFs by the present model agree well with the experimental data and with the experimental observation that dryout occurred first on the outer rods in 7-rod bundles. It is expected that the methodology used will be generally applicable in the prediction of CHF in rod bundles.

  17. 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

  18. Subchannel friction factors for rod bundles: laminar flow predictions and their application to turbulent flows

    International Nuclear Information System (INIS)

    Robinson, D.P.

    1979-02-01

    For the calculation of friction factors the use of correlations validated for smooth circular tubes along with the duct hydraulic diameter is known to be inappropriate for certain non-circular geometries. In order to test the validity and range of application of such correlations to the subchannels of rod bundles a computer programme has been written for the prediction of subchannel laminar velocity distributions and friction coefficients for fully developed flow. The theoretical basis and development of the programme is described along with comparisons between predictions and existing solutions for some simple geometries. Using the computer programme a wide range of calculations have been carried out for flow sections representing edge, corner and internal subchannels of rod bundles with particular emphasis on those of in-line pin bundle geometries. Where comparison can be made the predicted laminar coefficients are in excellent agreement with existing solutions. Although the approach adopted here could be used as the basis of a model for the subchannel axial friction factor, careful account should be taken of enhanced turbulent momentum transfer in situations where the flow is not unidirectional. (UK)

  19. 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.)

  20. 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

  1. 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

  2. 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

  3. COMETHE III-M for transient fuel rod behaviour prediction

    International Nuclear Information System (INIS)

    Billaux, M.; Vliet, J. van

    1983-01-01

    The COMETHE III-M version is being developed in order to provide fuel rod behaviour prediction capability both in steady-state and in transient situations. It also allows to estimate the fuel rod enthalpy evolution versus time or burnup which may be important in core-related safety studies. This paper describes the transient heat transfer models, including transient heat conduction inside the fuel rod, and a subchannel model providing transient flow as well as enthalpy calculation capability. Transient fission gas release is also modelled on basis of the change rate of oxide temperature. The models are illustrated by a few calculation examples. (author)

  4. Nuclear fuel assembly with improved spectral shift-producing rods

    International Nuclear Information System (INIS)

    Ferrari, H.M.

    1987-01-01

    This patent describes a nuclear reactor having fuel assemblies and a moderator-coolant liquid flowing through the fuel assemblies, each fuel assembly including an organized array of nuclear fuel rods wherein the moderator-coolant liquid flows along the fuel rods, at least one improved spectral shift-producing rod disposed among the fuel rods. The spectra shift-producing rod consists of: (a) an elongated hollow hermetically-sealed tubular member; (b) a weakened region formed in a portion of the member, the portion being subject to rupture at a given level of internal pressure; and (c) burnable poison material contained in the member which generates gas in the member as operation of the reactor proceeds normally, the material being soluble in the moderator-coolant liquid when brought into contact therewith; (d) the given level of internal pressure being less than the maximum level of internal pressure normally expected to be generated within the member by the poison material by normal operation of the reactor

  5. 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

  6. Wave Stresses in the Anvil Hammer Rods under Impact Including Ram Mass and Deformation Force of Forgings

    Directory of Open Access Journals (Sweden)

    V. M. Sinitskiy

    2016-01-01

    Full Text Available When operating the anvil hammers there occur impacts of die tooling and as a consequence, virtually instantaneous impact stops of motion of drop hammer parts. Such operating conditions come with accelerated failures of the anvil hammer rods because of emerging significant wave stresses. Engineering practice widely uses variation, difference, and integral methods to calculate wave stresses. However, to use them a researcher has to acquire certain skills, and the special programs should be available. The paper considers a method for estimating the wave stress changes in the anvil hammer rods, which is based on the wave equation of the Laplace transform. It presents a procedure for generating differential equations and their solution using the operator method. These equations describe the wave processes of strain and stress propagation in the anvil hammer rod under non-rigid impact with the compliance obstacle of the drop hammer parts. The work defines how the piston and rod mass and also the mechanical and geometric parameters of the rod influence on the stress level in the rod sealing of the hammer ram. Analysis of the results shows that the stresses in the rod sealing are proportional to the total amount of wave stresses caused by the rod and piston impact included in the total weight of the system. The piston influence on the stresses in the rod under impact is in direct proportion to the ratio of its mass to the mass of the rod. Geometric parameters of the rod and speed of drop parts before the impact influence on the stress value as well. It was found that if the time of impact is less than the time of the shock wave running in forward and backward direction, the impact with a compliance obstacle is equivalent to that of with a rigid obstacle, and the dependence of the wave stresses follows the Zhukovsky formula of direct pressure shock. The presented method of stress calculation can be successfully used to select the optimal mass and the rod

  7. 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

  8. Study on shadowing effect caused by transient rods at NSRR

    International Nuclear Information System (INIS)

    Nakamura, T.; Yachi, S.; Ishijima, K.

    1992-01-01

    Irregularly inserted three control rods created so called shadowing effects on some of the neutronic instruments at the Nuclear Safety Research Reactor (NSRR). During operations at the reactor power of up to 10 MW, the three control rods called transient rods, could be fully or partly inserted into the NSRR core. Reactor power monitors located outside of the core at the direction of deeply inserted transient rods indicated lower power in such operations. Power profiles of the reactor and neutron fluxes at power monitor locations were calculated with a three dimensional neutron diffusion code, CITATION. The calculation indicated that the real reactor power could be smaller than the measured maximum power by as mush as 30 % in such operations. The calculated neutron fluxes well described the changes in the apparent power monitor indications as a function of the transient rod position. (author)

  9. 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.)

  10. A study on the thermal hydraulics in rod bundles

    International Nuclear Information System (INIS)

    Chung, Moon Ki; Yang, Sun Kyu

    1989-03-01

    In order to improve the thermal hydraulic characteristics of the nuclear reactor core, it is necessary to obtain better understanding of the coolant flow and the enthalpy distribution in complex rod bundle geometries. The purpose of this report is to obtain a comprehensive survey on the thermal hydraulic in rod bundles from both experimental and numerical point of view. From references on experimental study, measurement methods and results of the flow velocity and the pressure drop in the subchannels of rod bundles are expressed. The microscopic flow characteristics of the subchannels and spacer grid effect on the flow structure are described. Physical phenomena and measurement methods of the secondary flow are also described. From references on the numerical study, general numerical methods are expressed. Numerical studies on the laminar flow and turbulent flow such as 1-equation and 2-equation model are reviewed.(Author)

  11. Large test rigs verify Clinch River control rod reliability

    International Nuclear Information System (INIS)

    Michael, H.D.; Smith, G.G.

    1983-01-01

    The purpose of the Clinch River control test programme was to use multiple full-scale prototypic control rod systems for verifying the system's ability to perform reliably during simulated reactor power control and emergency shutdown operations. Two major facilities, the Shutdown Control Rod and Maintenance (Scram) facility and the Dynamic and Seismic Test (Dast) facility, were constructed. The test programme of each facility is described. (UK)

  12. Critical experiment program of heterogeneous core composed for LWR fuel rods and low enriched uranyl nitrate solution

    International Nuclear Information System (INIS)

    Miyoshi, Yoshinori; Yamamoto, Toshihiro; Watanabe, Shouichi; Nakamura, Takemi

    2003-01-01

    In order to stimulate the criticality characteristics of a dissolver in a reprocessing plant, a critical experiment program of heterogeneous cores is under going at a Static Critical Experimental Facility, STACY in Japan Atomic Energy Research Institute, JAERI. The experimental system is composed of 5w/o enriched PWR-type fuel rod array immersed in 6w/o enriched uranyl nitrate solution. First series of experiments are basic benchmark experiments on fundamental critical data in order to validate criticality calculation codes for 'general-form system' classified in the Japanese Criticality Safety Handbook, JCSHB. Second series of experiments are concerning the neutron absorber effects of fission products related to the burn-up credit Level-2. For demonstrating the reactivity effects of fission products, reactivity effects of natural elements such as Sm, Nd, Eu and 103 Rh, 133 Cs, solved in the nitrate solution are to be measured. The objective of third series of experiments is to validate the effect of gadolinium as a soluble neutron poison. Properties of temperature coefficients and kinetic parameters are also studied, since these parameters are important to evaluate the transient behavior of the criticality accident. (author)

  13. Application of COMSOL in the solution of the neutron diffusion equations for fast nuclear reactors in stationary state

    International Nuclear Information System (INIS)

    Silva A, L.; Del Valle G, E.

    2012-10-01

    This work shows an application of the program COMSOL Multi physics Ver. 4.2a in the solution of the neutron diffusion equations for several energy groups in nuclear reactors whose core is formed by assemblies of hexagonal transversal cut as is the cas of fast reactors. A reference problem of 4 energy groups is described of which takes the cross sections which are processed by means of a program that prepares the definition of the constants utilized in COMSOL for the generic partial differential equations that this uses. The considered solution domain is the sixth part of the core which is applied frontier conditions of reflection and incoming flux zero. The discretization mesh is elaborated in automatic way by COMSOL and the solution method is one of finite elements of Lagrange grade two. The reference problem is known as the Knk with and without control rod which led to propose the calculation of the effective multiplication factor in function of the control rod fraction from a value 0 (completely inserted control rod) until the value 1 (completely extracted control rod). Besides this the reactivity was determined as well as the change of this in function of control rod fraction. The neutrons scalar flux for each energy group with and without control rod is proportioned. The reported results show a behavior similar to the one reported in other works but using the discreet ordinates S 2 approximation. (Author)

  14. Electrochemical catalytic activities of nanoporous palladium rods for methanol electro-oxidation

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Xiaoguang; Wang, Weimin; Qi, Zhen; Zhao, Changchun; Ji, Hong; Zhang, Zhonghua [Key Laboratory for Liquid-Solid Structural Evolution and Processing of Materials (MOE), School of Materials Science and Engineering, Shandong University, Jingshi Road 73, Jinan 250061 (China)

    2010-10-01

    A novel electrocatalyst, nanoporous palladium (npPd) rods can be facilely fabricated by dealloying a binary Al{sub 80}Pd{sub 20} alloy in a 5 wt.% HCl aqueous solution under free corrosion conditions. The microstructure of these nanoporous palladium rods has been characterized using scanning electron microscopy and transmission electron microscopy. The results show that each Pd rod is several microns in length and several hundred nanometers in diameter. Moreover, all the rods exhibit a typical three-dimensional bicontinuous interpenetrating ligament-channel structure with length scale of 15-20 nm. The electrochemical experiments demonstrate that these peculiar nanoporous palladium rods (mixed with Vulcan XC-72 carbon powders to form a npPd/C catalyst) reveal a superior electrocatalytic performance toward methanol oxidation in the alkaline media. In addition, the electrocatalytic activity obviously depends on the metal loading on the electrode and will reach to the highest level (223.52 mA mg{sup -1}) when applying 0.4 mg cm{sup -2} metal loading on the electrode. Moreover, a competing adsorption mechanism should exist when performing methanol oxidation on the surface of npPd rods, and the electro-oxidation reaction is a diffusion-controlled electrochemical process. Due to the advantages of simplicity and high efficiency in the mass production, the npPd rods can act as a promising candidate for the anode catalyst for direct methanol fuel cells (DMFCs). (author)

  15. Force analysis of the advanced neutron source control rod drive latch mechanism

    International Nuclear Information System (INIS)

    Damiano, B.

    1989-01-01

    The Advanced Neutron Source reactor (ANS), a proposed Department of Energy research reactor currently undergoing conceptual design at the Oak Ridge National Laboratory (ORNL), will generate a thermal neutron flux approximating 10 30 M -2 emdash S -1 . The compact core necessary to produce this flux provides little space for the shim safety control rods, which are located in the central annulus of the core. Without proper control rod drive design, the control rod drive magnets (which hold the control rod latch in a ready-to-scram position) may be unable to support the required load due to their restricted size. This paper describes the force analysis performed on the control rod latch mechanism to determine the fraction of control rod weight transferred to the drive magnet. This information will be useful during latch, control rod drive and magnet design. 5 refs., 12 figs

  16. 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.)

  17. 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.)

  18. An Examination Of Fracture Splitting Parameters Of Crackable Connecting Rods

    Directory of Open Access Journals (Sweden)

    Zafer Özdemir

    2000-06-01

    Full Text Available Fracture splitting method is an innovative processing technique in the field of automobile engine connecting rod (con/rod manufacturing. Compared with traditional method, this technique has remarkable advantages. Manufacturing procedures, equipment and tools investment can be decreased and energy consumption reduced remarkably. Furthermore, product quality and bearing capability can also be improved. It provides a high quality, high accuracy and low cost route for producing connecting rods (con/rods. With the many advantages mentioned above, this method has attracted manufacturers attention and has been utilized in many types of con/rod manufacturing. In this article, the method and the advantages it provides, such as materials, notches for fracture splitting, fracture splitting conditions and fracture splitting equipment are discussed in detail. The paper describes an analysis of examination of fracture splitting parameters and optik-SEM fractography of C70S6 crackable connectıng rod. Force and velocity parameters are investigated. That uniform impact force distrubition starting from the starting notch causes brittle and cleavage failure mode is obtained as a result. This induces to decrease the toughness.

  19. Coupler for nuclear reactor absorber rods

    International Nuclear Information System (INIS)

    Kerz, K.

    1984-01-01

    A coupler is described for absorber rods being suspended during operation of nuclear reactors which includes plurality of actuating elements being movable for individually and jointly releasing the coupler, the movement of each of the actuating elements for releasing the coupler being independently controllable

  20. A two-dimensional finite element method for analysis of solid body contact problems in fuel rod mechanics

    International Nuclear Information System (INIS)

    Nissen, K.L.

    1988-06-01

    Two computer codes for the analysis of fuel rod behavior have been developed. Fuel rod mechanics is treated by a two-dimensional, axisymmetric finite element method. The program KONTAKT is used for detailed examinations on fuel rod sections, whereas the second program METHOD2D allows instationary calculations of whole fuel rods. The mechanical contact of fuel and cladding during heating of the fuel rod is very important for it's integrity. Both computer codes use a Newton-Raphson iteration for the solution of the nonlinear solid body contact problem. A constitutive equation is applied for the dependency of contact pressure on normal approach of the surfaces which are assumed to be rough. If friction is present on the contacting surfaces, Coulomb's friction law is used. Code validation is done by comparison with known analytical solutions for special problems. Results of the contact algorithm for an elastic ball pressing against a rigid surface are confronted with Hertzian theory. Influences of fuel-pellet geometry as well as influences of discretisation of displacements and stresses of a single fuel pellet are studied. Contact of fuel and cladding is calculated for a fuel rod section with two fuel pellets. The influence of friction forces between fuel and cladding on their axial expansion is demonstrated. By calculation of deformations and temperatures during an instationary fuel rod experiment of the CABRI-series the feasibility of two-dimensional finite element analysis of whole fuel rods is shown. (orig.) [de

  1. Determination of equivalent cross sections for representation of control rod regions in diffusion calculations

    International Nuclear Information System (INIS)

    Scherer, W.; Neef, H.J.

    1976-07-01

    The representation of control rod regions in reactor calculations requires a combination of transport and diffusion theory calculations. A method is described which produces equivalent cross sections for a rodded region. These cross sections used in a diffusion theory calcualtion yield the same rod efficiency and reaction rate distribution as the transport theory calculation for the explicit heterogeneous control rod. The description of the method is complemented by sample problems. (orig.) [de

  2. Construction of supporting grids for fuel rods (or tubes in a heat exchanger)

    International Nuclear Information System (INIS)

    1975-01-01

    The construction of supporting grids for fuel rods (or tubes in heat exchangers) is described. It is a modification of a former French patent. The modification consists in the use of different material for the springs keeping the rod in place and describes another way of fixing these blade-shaped springs. Advantages of the specific spring characteristics were taken into consideration

  3. Simulation of the fuel rod thermal hydraulic performance during the blow down phase in a PWR

    International Nuclear Information System (INIS)

    Gadelha, J.A.M.

    1982-10-01

    A digital computer code to predict the fuel rod thermalhydraulic performance during a postulated loss-of-coolant accident (LOCA) in the primary circuit of a PWR nuclear power plant is developed. The fuel rod corresponds to that in an average channel in the core. Only the blowdown phase is considered during the accident. The conservation equations of mass, momentum, and energy, and the heat conduction equation are solved to determine the fuel rod conditions during the accident. Finite differences are applied as a numerical method in the solution of the equations modelling the rod and coolant conditions. (Author) [pt

  4. Nuclear fuel rod helium leak inspection apparatus and method

    International Nuclear Information System (INIS)

    Ahmed, H.J.

    1991-01-01

    This patent describes an inspection apparatus for testing nuclear fuel rods for helium leaks. It comprises a test chamber being openable and closable for receiving at least one nuclear fuel rod; means separate from the fuel rod for supplying helium and constantly leaking helium at a predetermined known positive value into the test chamber to constantly provide an atmosphere of helium at the predetermined known positive value in the test chamber; and means for sampling the atmosphere within the chamber and measuring the helium in the atmosphere such that a measured helium value below a preset minimum helium value substantially equal to the predetermined known positive value of the atmosphere of helium being constantly provided in the test chamber indicates a malfunction in the inspection apparatus, above a preset maximum helium value greater than the predetermined known positive in the test chamber indicates the existence of a helium leak from the fuel rod, or between the preset minimum and maximum helium values indicates the absence of a helium leak from the fuel rod

  5. Thermal behavior simulation of a nuclear fuel rod through an eletrically heated rod

    International Nuclear Information System (INIS)

    Lima, R. de C.F. de.

    1984-01-01

    In thermalhydraulic loops the nuclear industry often uses electrically heated rods to simulate power transients, which occur in nuclear fuel rods. The development and design of a electrically heated rod, by supplying the dimensions and materials which should be used in order to yeld the same temperature and heat flux at the surfaces of the nuclear rod and the electrically heated rod are presented. To a given nuclear transient this equality was obtained by fitting the linear power through the lumped parameters technique. (Author) [pt

  6. Apparatus and method for loading fuel rods into grids of a fuel assembly

    International Nuclear Information System (INIS)

    De Mario, E.E.; Burman, D.L.; Olson, C.A.; Secker, J.R.

    1987-01-01

    This patent describes a fuel assembly having fuel rods and at least one grid formed of interleaved straps and yieldable springs, the interleaved straps defining hollow cells aligned in rows and columns thereof for receiving the respective fuel rods. A pair of the springs are disposed within each of the cells for engaging and supporting one of the fuel rods when received in the cell. An apparatus is described for facilitating the loading of the fuel rods into the grid of the fuel assembly, comprising: (a) first mean insertable concurrently into the cells of the grid for engaging and moving the springs from respective first positions in which each pair of springs will engage a respective fuel rod when disposed within the grid cell to respective second positions in which each pair of springs is disengaged from the respective fuel rod when disposed within the grid cell; (b) a pair of second means, one of the pair of the second means being insertable concurrently into the rows of the cells of the grid and the other of the pair of second means being insertable concurrently into the column of the cells

  7. 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

  8. 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

  9. 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.)

  10. A theory of longitudinally polarised piezocomposite rod based on Mindlin-Herrmann model

    CSIR Research Space (South Africa)

    Shatalov, M

    2010-09-01

    Full Text Available The conventional theory of the piezoelectric rod is based on an assumption that its lateral vibrations are negligible. In this case the rod, vibration could be described in terms of one-dimensional wave equation and a set of mechanical and electric...

  11. A Patient With Pyruvate Carboxylase Deficiency and Nemaline Rods on Muscle Biopsy

    DEFF Research Database (Denmark)

    Unal, Ozlem; Orhan, Diclehan; Ostergaard, Elsebet

    2013-01-01

    Nemaline rods are the pathologic hallmark of nemaline myopathy, but they have also been described as a secondary phenomenon in a variety of other disorders. Nemaline rods have not been reported in pyruvate carboxylase deficiency before. Here we present a patient with pyruvate carboxylase deficiency...

  12. Corrugated thimble tube for controlling control rod descent in nuclear reactor

    International Nuclear Information System (INIS)

    Luetzow, H.J.

    1981-01-01

    A thimble tube construction is described which will provide a controlled descent for a control rod while minimizing the reaction forces which must be absorbed by the thimble tube and reducing the possibility that a foreign particle could interfere with the free descent of a control rod. A thimble tube is formed with helically-corrugate internal walls which cooperate with a control rod contained in the tube in an emergency situation to provide a progressively-increasing hydraulic restraining force as each adjacent corrugation is encountered

  13. 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

  14. Recursive analytical solution describing artificial satellite motion perturbed by an arbitrary number of zonal terms

    Science.gov (United States)

    Mueller, A. C.

    1977-01-01

    An analytical first order solution has been developed which describes the motion of an artificial satellite perturbed by an arbitrary number of zonal harmonics of the geopotential. A set of recursive relations for the solution, which was deduced from recursive relations of the geopotential, was derived. The method of solution is based on Von-Zeipel's technique applied to a canonical set of two-body elements in the extended phase space which incorporates the true anomaly as a canonical element. The elements are of Poincare type, that is, they are regular for vanishing eccentricities and inclinations. Numerical results show that this solution is accurate to within a few meters after 500 revolutions.

  15. 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.)

  16. 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.)

  17. A digital position-indication system for control rods

    International Nuclear Information System (INIS)

    Nishizawa, Yukio; Hayakawa, Toshifumi

    1979-01-01

    Systems that detect and indicate the position of the control rods that regulate the thermal output of a nuclear reactor play a particularly important role in monitoring its operational status. Conventionally, control rod position indication in pressurized water reactors has been of the analog type, utilizing the principle of the differential transformer. The present digital system was developed with the objective of achieving greater stability, greater accuracy, and higher reliability. The article gives a general description of the system and describes its advantages. (author)

  18. Gap conductance in Zircaloy-clad LWR fuel rods

    International Nuclear Information System (INIS)

    Ainscough, J.B.

    1982-04-01

    This report describes the procedures currently used to calculate fuel-cladding gap conductance in light water reactor fuel rods containing pelleted UO 2 in Zircaloy cladding, under both steady-state and transient conditions. The relevant theory is discussed together with some of the approximations usually made in performance modelling codes. The state of the physical property data which are needed for heat transfer calculations is examined and some of the relevant in- and out-of-reactor experimental work on fuel rod conductance is reviewed

  19. BRIGITTE, Dose Rate and Heat Source and Energy Flux for Self-Absorbing Rods

    International Nuclear Information System (INIS)

    Jegu, M.; Clement, M.

    1978-01-01

    1 - Nature of physical problem solved: Calculation of dose rate, heat sources or energy flux. The sources are self-absorbing radioactive rods. The shielding consists of blocks of which the cross section can be defined. 2 - Method of solution: Exponential attenuation and build-up factor between source points and detector points. Source integration with error estimate. Automatic or controlled build-up with monitor print-out. 3 - Restrictions on the complexity of the problem: Number of energy points, regions, detector points, abscissa points of the rod, vertical position of the rod, are all limited to ten. The maximum total number of vertical steps is 124

  20. Potential impacts of crud deposits on fuel rod behaviour on high powered PWR fuel rods

    International Nuclear Information System (INIS)

    Wilson, W.; Comstock, R.J.

    1999-01-01

    Fuel assemblies operating with significant sub-cooled boiling are subject to deposition of surface deposits commonly referred to as crud. This crud can potentially cause concentration of chemical species within the deposits which can be detrimental to cladding performance in PWRs. In addition, these deposits on the surface of the cladding can result in power anomalies and erroneous reporting of fuel rod oxide thickness which can substantially hamper corrosion and core performance modeling efforts. Data is presented which illustrates the importance of accounting for the presence of crud on fuel cladding surfaces. Several methods used to correct for this phenomenon when collecting and analyzing zirconium alloy field oxide thickness measurements are described. Various observations related to crud characteristics and its impact on fuel rod performance are also addressed. (author)

  1. Controlling a nuclear reactor with dropped control rods

    International Nuclear Information System (INIS)

    Mc Atee, K.R.; Alsop, B.H.

    1987-01-01

    A control system is described for a nuclear power plant including a reactor with a core having an upper portion and a lower portion and control rods which are inserted into and withdrawn from the core of the reactor vertically to control reactivity in the core. The system comprises: means to measure neutron flux separately in the upper portion and the lower portion of the reactor and to generate from such measurements a signal representative of axial distribution of power between the upper and lower portions of the reactor core; means to detect a dropped control rod in the reactor and to generate a dropped rod signal in response thereto; means to generate an axial power distribution limit signal representative of a critical axial power distribution for a dropped rod condition; means to compare the axial power distribution signal to the axial power distribution limit signal and to generate an axial power distribution out of limits signal when the axial power distribution signal exceeds the axial power distribution limit signal; and means responsive only to the presence of both the dropped rod signal and the axial power distribution out of limits signal to generate a signal for shutting the reactor down

  2. Development of a 60Co radioactive rod source used for γ-ray level gauge

    International Nuclear Information System (INIS)

    Lin Yibing; Pan Liangcai; Yin Shunjiu

    1991-09-01

    The installation of level gauge used for urea stripping tower, the structure and forming of radioactive rod source, and the calculation of its approximate linear graduation are described. The theoretical and practical feasibility has been confirmed from the test results of comparing the imported radioactive rod source to the developed radioactive rod source. The technological process of production, method for obtaining distribution of radioactivity along the axis, and the test and operation of developed rod source on site are also presented

  3. 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...

  4. Mathematical model of rod oscillations with account of material relaxation behaviour

    Science.gov (United States)

    Kudinov, I. V.; Kudinov, V. A.; Eremin, A. V.; Zhukov, V. V.

    2018-03-01

    Taking into account the bounded velocity of strains and deformations propagation in the formula given in the Hooke’s law, the authors have obtained the differential equation of rod damped oscillations that includes the first and the third time derivatives of displacement as well as the mixed derivative (with respect to space and time variables). Study of its precise analytical solution found by means of separation of variables has shown that rod recovery after being disturbed is accompanied by low-amplitude damped oscillations that occur at the start time and only within the range of positive displacement values. The oscillations amplitude decreases with increase of relaxation factor. Rod is recovered virtually without an oscillating process both in the limit and with any high values of the relaxation factor.

  5. Analysis of fuel rod behaviour within a rod bundle of a pressurized water reactor under the conditions of a loss of coolant accident (LOCA) using probabilistic methodology

    International Nuclear Information System (INIS)

    Sengpiel, W.

    1980-12-01

    The assessment of fuel rod behaviour under PWR LOCA conditions aims at the evaluation of the peak cladding temperatures and the (final) maximum circumferential cladding strains. Moreover, the estimation of the amount of possible coolant channel blockages within a rod bundle is of special interest, as large coplanar clad strains of adjacent rods may result in strong local reductions of coolant channel areas. Coolant channel blockages of large radial extent may impair the long-term coolability of the corresponding rods. A model has been developed to describe these accident consequences using probabilistic methodology. This model is applied to study the behaviour of fuel rods under accident conditions following the double-ended pipe rupture between collant pump and pressure vessel in the primary system of a 1300 MW(el)-PWR. Specifically a rod bundle is considered consisting of 236 fuel rods, that is subjected to severe thermal and mechanical loading. The results obtained indicate that plastic clad deformations with circumferential clad strains of more than 30% cannot be excluded for hot rods of the reference bundle. However, coplanar coolant channel blockages of significant extent seem to be probable within that bundle only under certain boundary conditions which are assumed to be pessimistic. (orig./RW) [de

  6. Confinement stabilises single crystal vaterite rods.

    OpenAIRE

    Schenk, AS; Albarracin, EJ; Kim, YY; Ihli, J; Meldrum, FC

    2014-01-01

    Single-crystals of vaterite, the least-stable anhydrous polymorph of CaCO3, are rare in biogenic and synthetic systems. We here describe the synthesis of high aspect ratio single crystal vaterite rods under additive-free conditions by precipitating CaCO3 within the cylindrical pores of track-etch membranes.

  7. Determination of transient temperature and heat flux on the surface of a reactor control rod based on temperature measurements at the interior points

    International Nuclear Information System (INIS)

    Cebula, Artur; Taler, Jan

    2014-01-01

    The paper presents heat transfer calculation results concerning a control rod of nuclear power plant. Apart from numerical calculation results, experimental heat transfer measurements of the control rod model are also presented. The control rod that is the object of interest is surrounded by a mixing region of hot and cold streams and, as a consequence, is subjected to thermal fluctuations. The paper describes a method based on the solution of the inverse heat conduction problem (IHCP) for determining heat flux on the outer surface of the rod. Numerical tests were conducted to validate the method by comparison of the results with the time changes of surface temperature and heat flux which were obtained from the computational fluid dynamics (CFD) simulation of the mixing process. A measuring instrument was designed to measure the heat flux at the outer surface of the control rod model. In addition, the principle of operation and construction of heat flux meter is presented in detail. -- Highlights: • Temperature and heat flux estimation during cooling of control rod are presented. • The inverse technique is based on the space marching method. • The instrument for surface heat flux measurement was manufactured and tested. • CFD simulations were used to validate the developed inverse technique. • Actual data were used to demonstrate practical applicability of the method

  8. Investigation of a frequency dependent transfer function and its application to control rod localization

    International Nuclear Information System (INIS)

    Garis, N.S.; Pazsit, I.

    1997-02-01

    Control rod vibrations can be detected via the fluctuations they generate in the neutron flux, i.e. the neutron noise. In a previous paper, a neural network-based algorithm for locating a vibrating control rod from the measured neutron noise was developed. The transfer function used for the core model was based on the so called power-reactor approximation resulting in a simple, real-valued solution which means that the phase delay of the signal propagation is neglected. In the present work a more realistic transfer function is used, without the approximations of the previous model. The transfer function is calculated from the Fourier transformed diffusion equation with a complex, frequency dependent buckling leading to a complex solution. In physical terms, this means that the phase delay of the signal propagation is accounted for. Using such a complex core model, the present paper investigates the effectiveness of applying neural networks for control rod localisation. 7 refs, 4 figs

  9. 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.

  10. Radiological characterization of spent control rod assemblies

    International Nuclear Information System (INIS)

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

    1995-10-01

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

  11. In-pile experiments on fuel rod behaviour during a LOCA

    International Nuclear Information System (INIS)

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

    1981-07-01

    This report describes the results of the Test Series G2/3 within the in-pile experimental program for the investigation of LWR fuel rod behavior. The results were obtained with single rods of a PWR design in the DK loop of the FR2 reactor at the Kernforschungszentrum Karlsruhe (KfK). 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 rods. The main parameter of the test program ist the burnup, ranging from 2500 to 35000 MWd/t. The results of test series G2/3 (35000 MWd/t) with respect to the burst data, i.e. burst temperature, burst pressure, and burst strain, do not indicate major differences from the in-pile tests with unirradiated test specimens. (orig.) [de

  12. In-pile experiemts on fuel rod behavior during a LOCA

    International Nuclear Information System (INIS)

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

    1981-02-01

    This report describes the results of the Test Series G1 within the in-pile experimental program for the investigation of LWR fuel rod behavior. The results were obtained with single rods of a PWR design in the DK loop of the FR2 reactor at the Kernforschungszentrum Karlsruhe (KfK). The in-pile tests with the objective of investigating the influence of a nuclear environment on the mechansims of fuel rod failure were being performed with irradiated and unirradiated rods. The main parameter of the test program is the burnup ranging from 2500 to 35 000 MWd/t. The results of test series G1 (35 000 MWd/t) with respect to the burst data, i.e. burst temperature, burst pressure, and burst strain, do not indicate major differences from the in-pile tests with unirradiated test specimens. (orig.) [de

  13. Axial transport of fission gas in LWR fuel rods

    International Nuclear Information System (INIS)

    Kinoshita, M.

    1983-01-01

    With regard to fission gas transportation inside the fuel rod, the following three mechanisms are important: (1) a localized and time dependent fission gas release from UO 2 fuel to pellet/clad gap, (2) the consequent gas pressure difference between the gap and the plenum, and (3) the inter-diffusion of initially filled Helium and released fission gas such as Xenon. Among these three mechanisms, the 2nd mechanism would result in the one dimensional flow through P/C gap in the axial direction, while the 3rd would average the local fission gas concentration difference. In this paper, an attempt was made to develop a computerized model, LINUS (LINear flow and diffusion under Un-Steady condition) describing the above two mechanisms, items (2) and (3). The item (1) is treated as an input. The code was applied to analyse short length experimental fuel rods and long length commercial fuel rods. The calculated time evolution of Xe concentration along the fuel column shows that the dilution rate of Xe in commercial fuel rods is much slower than that in short experimental fuel rods. Some other sensitivity studies, such as the effect of pre-pressurization, are also presented. (author)

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

    International Nuclear Information System (INIS)

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

    1992-01-01

    The reactor safety rods may be subjected to high temperatures due to gamma heating after the core coolant level has dropped during the ECS phase of a hypothetical LOCA event. Accordingly, an experimental cadmium safety rod testing subtask was established as part of a task to address the response of reactor core components to this accident. Companion reports describe the experiments and a structural evaluation (finite element analysis) of the safety rod. This report deals primarily with the examination of the test specimens, evaluation of possible failure mechanisms, and confirmatory separate effects experiments. It is concluded that the failures observed in the cadmium safety rod thermal tests which occurred at low temperature (T 800 degrees C) with fast thermal ramp rates are concluded to be mechanical in nature without significant environmental degradation. Based on these tests, tasks were initiated to design and manufacture B 4 C safety rods to replace the cadmium safety rods. The B 4 C safety rods have been manufactured at this time and it is currently planned to charge them to the reactor in the near future. 60 refs

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

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  16. The SABRE code for fuel rod cluster thermohydraulics

    International Nuclear Information System (INIS)

    Macdougall, J.D.; Lillington, J.N.

    1984-01-01

    This paper describes the capabilities of the SABRE code for the calculation of single phase and two phase fluid flow and temperature in fuel pin bundles, discusses the methods used in the modelling and solution of the problem, and presents some results including comparison with experiments. The SABRE code permits calculation of steady-state or transient, single or two phase flows and the geometrical options include general representation of grids, wire wraps, multiple blockages, bowed pins, etc. The derivation and solution of the difference equations is discussed. Emphasis is given to the derivation of the spatial differences in triangular subchannel geometry, and the use of central, upward or vector upwind schemes. The method of solution of the difference equations is described for both steady state and transient problems. Together with these topics we consider the problems involved in turbulence modelling and how it is implemented in SABRE. This includes supporting work with a fine scale curvilinear coordinate programme to provide turbulence source data. The problem of modelling boiling flows is discussed, with particular reference to the numerical problems caused by the rapid density change on boiling. The final part of the paper presents applications of the code to the analysis of blockage situations, the study of flow and power transients and analysis of natural circulation within clusters to demonstrate the scope of the code and compare with available experimental results. The comparisons include the calculation of a flow pressure drop characteristic of a boiling channel showing the Ledinegg instability, examples of overpower and flow rundown transients which lead to coolant boiling, and calculation of natural circulation within a rod cluster. (orig./GL)

  17. 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 8 discusses Control System SOT Tests Results and Analysis Report. This is a continuation of Book 7

  18. 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.)

  19. 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.)

  20. 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

  1. Pressure loss in two-phase flow through a microchannel rod bundle

    International Nuclear Information System (INIS)

    Smith, A.C.; Hamm, L.L.; Qureshi, Z.; Steeper, T.J.

    1998-01-01

    The purpose of the microchannel rod bundle two-phase flow test described here was to provide data for benchmarking safety analyses for the accelerator production of tritium (APT). The objective was to obtain pressure loss data for a typical accelerator target rod bundle over a wide range of two-phase flow conditions. The test rod bundle assembly was fabricated for single-phase pressure drop tests conducted at Los Alamos National Laboratory (LANL) and subsequently used for the two-phase flow testing described here. The results for a typical case are given. These results fall generally in the slug flow regime for the horizontal flow results of Fukano and Kariyasaki for a 1.0-mm circular channel. Fukano and Kariyasaki found that surface tension effects were dominant in the 1-mm channel and report no churn regime. The results were also compared with the flow regime maps given by Triplett et al. for flow in discrete microchannels. Triplett employed both circular and trapezoidal channels, the latter to approximate the rod bundle interstitial flow channel shape. It was found that the rod bundle flow fell across the slug-to-churn flow regime transition reported by Triplett. This is consistent with the expectation that cross flow among channels would result in turbulent mixing and would suppress the formation of large discrete bubbles

  2. Development of non-destructive examination system for irradiated fuel rods

    International Nuclear Information System (INIS)

    Sumerling, R.; Goldsmith, L.A.; Cross, M.T.; McKee, F.

    1978-12-01

    The development of non-destructive examination (NDE) system for irradiated fuel rods is described. The system is used for testing rods within a concrete cave and consists of three parts: a fully-automated fuel rod-drive machine, designed for easy maintenance; a series of plug-in NDE modules which fit into the central space provided in the machine, plus optical/TV viewing devices and gamma-scan equipment lined up on the rod; and on electronic control equipment situated outside the concrete shielding. The equipment is at present routinely used for viewing, eddy-current testing, gamma-scanning and diameter measurement of rods. The system is flexible in that additional modules can be added later as they are developed, since there is room for three modules of standard size (about 10cm x 10 cm x 3cm) in the machine or one large module taking the full space. New developments include the use of dual frequency eddy-current testing, which allows much greater discrimination against unwanted signals, and measurement of oxide thickness using a high frequency eddy-current probe. (author)

  3. 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.)

  4. Anisotropy in CdSe quantum rods

    Energy Technology Data Exchange (ETDEWEB)

    Li, Liang-shi [Univ. of California, Berkeley, CA (United States)

    2003-01-01

    concentration. The preparation of the stable liquid crystalline solution of CdSe nanorods is described, as well as the large-scale alignment of the nanorods by taking advantage of the long-range orientational correlation in the liquid crystals. In addition, we investigated the phase diagram of the nanorod solution, as a step toward understanding the possible role of the long-range attractive interaction between the nanorods in the formation of lyotropic liquid crystals.

  5. Apparatus and method for applying an end plug to a fuel rod tube end

    International Nuclear Information System (INIS)

    Rieben, S.L.; Wylie, M.E.

    1987-01-01

    An apparatus is described for applying an end plug to a hollow end of a nuclear fuel rod tube, comprising: support means mounted for reciprocal movement between remote and adjacent positions relative to a nuclear fuel rod tube end to which an end plug is to be applied; guide means supported on the support means for movement; and drive means coupled to the support means and being actuatable for movement between retracted and extended positions for reciprocally moving the support means between its respective remote and adjacent positions. A method for applying an end plug to a hollow end of a nuclear fuel rod tube is also described

  6. 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.

  7. 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)

  8. 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.)

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

    Science.gov (United States)

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

    2011-11-01

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

  10. Behavior of a nine-rod PWR bundle under power-cooling-mismatch conditions

    International Nuclear Information System (INIS)

    Gunnerson, F.S.; Sparks, D.T.

    1979-01-01

    An experiment to characterize the behavior of a nine-rod pressurized water reactor (PWR) fuel bundle operating during power-cooling-mismatch (PCM) conditions has been conducted in the Power Burst Facility (PBF) at the Idaho National Engineering Laboratory (INEL). The experiment, designated Test PCM-5, is part of a series of PCM experiments designed to evaluate light water reactor (LWR) fuel rod response under postulated accident conditions. Test PCM-5 was the first nine-rod bundle experiment in the PCM test series. The primary objectives and the results of the experiment are described

  11. 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.)

  12. 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

  13. 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

  14. Longitudinally Vibrating Elastic Rods with Locally and Non-Locally Reacting Viscous Dampers

    Directory of Open Access Journals (Sweden)

    Şefaatdin Yüksel

    2005-01-01

    Full Text Available Eigencharacteristics of a longitudinally vibrating elastic rod with locally and non-locally reacting damping are analyzed. The rod is considered as a continuous system and complex eigenfrequencies are determined as solution of a characteristic equation. The variation of the damping ratios with respect to damper locations and damping coefficients for the first four eigenfrequencies are obtained. It is shown that at any mode of locally or non-locally damped elastic rod, the variation of damping ratio with damper location is linearly proportional to absolute value of the mode shape of undamped system. It is seen that the increasing damping coefficient does not always increase the damping ratio and there are optimal values for the damping ratio. Optimal values for external damping coefficients of viscous dampers and locations of the dampers are presented.

  15. Liquid crystallinity in flexible and rigid rod polymers

    International Nuclear Information System (INIS)

    Pickett, Galen T.; Schweizer, Kenneth S.

    2000-01-01

    We apply an anisotropic version of the polymer reference interaction site model (PRISM) integral equation description of flexible polymers to analyze athermal liquid crystallinity. The polymers are characterized by a statistical segment length, σ o , and by a physical hard-core thickness, d, that prevents the overlap of monomers on different chains. At small segment densities, ρ, the microscopic length scale d is irrelevant (as it must be in the universal semidilute regime), but becomes important in concentrated solutions and melts. Under the influence of the excluded volume interactions alone, the chains undergo a lyotropic, first-order isotropic-nematic transition at a concentration dependent upon the dimensionless ''aspect ratio,'' σ o /d. The transition becomes weaker as d→0, becoming second order, as has been previously shown. We extend the theory to describe the transition of rigid, thin rods, and discuss the evolution of the anisotropic liquid structure in the ordered phase. (c) 2000 American Institute of Physics

  16. 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

  17. Removable control rod drive shaft guide

    International Nuclear Information System (INIS)

    Ales, M.W.; Brown, S.K.; Dixon, L.D.

    1988-01-01

    A removable control rod drive shaft guide is described for a control rod ''guide'' structure card, comprising: a. a substantially annular shaped main body portion having a central axial bore for receiving a control rod drive shaft and an upper exterior groove for receiving removal tooling; b. the main body portion having a reduced outer diameter at its lower section; c. a shoulder portion integral with the main body portion for supporting the main body portion on the guide structure card; d. the shoulder portion having a substantially radial bore and the reduced outer diameter lower section having a slot in alignment with the radial bore; e. a locking arm ''pivotaly'' mounted in the radial bore which protrudes into the slot and is movable between a first normal locking position for engaging the guide structure card and a second release position; f. a spring received within a second axial bore in the main body portion and biased against the locking arm for urging and locking arm into the first normal locking position; and g. a release tab at one end of the locking arm for moving the locking arm into the second release position

  18. Rod-Shaped Magnetite Nano/Microparticles Synthesis at Ambient Temperature

    Directory of Open Access Journals (Sweden)

    Balaprasad Ankamwar

    2013-01-01

    Full Text Available Here, we reported room temperature synthesis of Fe3O4 rod-shaped nano/microparticles by chemical reduction method from FeCl3 precursor and NaBH4 as the reducing agent in the presence of the pyrrole as a capping agent. The magnetic Fe3O4 particles were characterized by several methods, such as SEM, XRD, FTIR, and TGA. The average aspect ratio of Fe3O4 rod-shaped particles was ~2.8. These particles were redispersed in deionised water to form a colloidal solution and showed magnetic properties. This economical synthesis route is scalable, and Fe3O4 particles can be exploited for various applications such as MRI contrast enhancement, biodiseperations, Ni-Fe batteries, and as a catalyst.

  19. 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.)

  20. 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

  1. 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.

  2. Wave-Breaking Phenomena and Existence of Peakons for a Generalized Compressible Elastic-Rod Equation

    Directory of Open Access Journals (Sweden)

    Xiaolian Ai

    2014-01-01

    Full Text Available Consideration in this paper is the Cauchy problem of a generalized hyperelastic-rod wave equation. We first derive a wave-breaking mechanism for strong solutions, which occurs in finite time for certain initial profiles. In addition, we determine the existence of some new peaked solitary wave solutions.

  3. 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.)

  4. 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.)

  5. Transient calculation performance of the MASTER code for control rod ejection problem

    International Nuclear Information System (INIS)

    Cho, B. O.; Joo, H. G.; Yoo, Y. J.; Park, S. Y.; Zee, S. Q.

    1999-01-01

    The accuracy and the effectiveness of the solution methods of the MASTER code for reactor transient problems were analyzed with a set of NEACRP PWR control rod ejection benchmark problems. A series of sensitivity study for the effects on the solution by the neutronic solution methods and the neutronic and thermal-hydraulic model parameters were thus investigated. The MASTER results were then compared with the reference PANTHER results. This indicates that the MASTER solution is sufficiently accurate and the computing time is fast enough for nuclear design application

  6. Transient calculation performance of the MASTER code for control rod ejection problem

    Energy Technology Data Exchange (ETDEWEB)

    Cho, B. O.; Joo, H. G.; Yoo, Y. J.; Park, S. Y.; Zee, S. Q. [KAERI, Taejon (Korea, Republic of)

    1999-10-01

    The accuracy and the effectiveness of the solution methods of the MASTER code for reactor transient problems were analyzed with a set of NEACRP PWR control rod ejection benchmark problems. A series of sensitivity study for the effects on the solution by the neutronic solution methods and the neutronic and thermal-hydraulic model parameters were thus investigated. The MASTER results were then compared with the reference PANTHER results. This indicates that the MASTER solution is sufficiently accurate and the computing time is fast enough for nuclear design application.

  7. Methods of assembling and disassembling spider and burnable poison rod structures for nuclear reactors

    International Nuclear Information System (INIS)

    Walton, L.A.

    1981-01-01

    A method is described of joining burnable poison rods to the spider arms of a pressurised water power reactor fuel assembly which is proof against the reactor core environment but permits these rods to be removed from the spider simply, swiftly and delicately. (U.K.)

  8. Analytical solution describing pesticide volatilization from soil affected by a change in surface condition.

    Science.gov (United States)

    Yates, S R

    2009-01-01

    An analytical solution describing the fate and transport of pesticides applied to soils has been developed. Two pesticide application methods can be simulated: point-source applications, such as idealized shank or a hot-gas injection method, and a more realistic shank-source application method that includes a vertical pesticide distribution in the soil domain due to a soil fracture caused by a shank. The solutions allow determination of the volatilization rate and other information that could be important for understanding fumigant movement and in the development of regulatory permitting conditions. The solutions can be used to characterize differences in emissions relative to changes in the soil degradation rate, surface barrier conditions, application depth, and soil packing. In some cases, simple algebraic expressions are provided that can be used to obtain the total emissions and total soil degradation. The solutions provide a consistent methodology for determining the total emissions and can be used with other information, such as field and laboratory experimental data, to support the development of fumigant regulations. The uses of the models are illustrated by several examples.

  9. Study of heat transfer in 3D fuel rods of the EPRI-9R reactor modified

    International Nuclear Information System (INIS)

    Affonso, Renato Raoni Werneck; Lava, Deise Diana; Borges, Diogo da Silva; Sampaio, Paulo Augusto Berquo de; Moreira, Maria de Lourdes

    2014-01-01

    This paper aims to conduct a case study of the fuel rods that have the highest and the lowest average power of the EPRI-9R 3D reactor modified , for various positions of the control rods banks. For this, will be addressed the verification of computer code, comparing the results obtained with analytical solutions. This check is important so that, subsequently, it is possible use the program to understand the behavior of the fuel rods and the coolant channel of the EPRI-9R 3D reactor modified. Thus, in view of the scope of this paper, first a brief introducing on the heat transfer is done, including the rod equations and the equation of energy in the channel to allow the analysis of the results

  10. 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

  11. Solutions Stability of Initial Boundary Problem, Modeling of Dynamics of Some Discrete Continuum Mechanical System

    Directory of Open Access Journals (Sweden)

    D. A. Eliseev

    2015-01-01

    Full Text Available The solution stability of an initial boundary problem for a linear hybrid system of differential equations, which models the rotation of a rigid body with two elastic rods located in the same plane is studied in the paper. To an axis passing through the mass center of the rigid body perpendicularly to the rods location plane is applied the stabilizing moment proportional to the angle of the system rotation, derivative of the angle, integral of the angle. The external moment provides a feedback. A method of studying the behavior of solutions of the initial boundary problem is proposed. This method allows to exclude from the hybrid system of differential equations partial differential equations, which describe the dynamics of distributed elements of a mechanical system. It allows us to build one equation for an angle of the system rotation. Its characteristic equation defines the stability of solutions of all the system. In the space of feedback-coefficients the areas that provide the asymptotic stability of solutions of the initial boundary problem are built up.

  12. On the perfect hexagonal packing of rods

    International Nuclear Information System (INIS)

    Starostin, E L

    2006-01-01

    In most cases the hexagonal packing of fibrous structures or rods extremizes the energy of interaction between strands. If the strands are not straight, then it is still possible to form a perfect hexatic bundle. Conditions under which the perfect hexagonal packing of curved tubular structures may exist are formulated. Particular attention is given to closed or cycled arrangements of the rods like in the DNA toroids and spools. The closure or return constraints of the bundle result in an allowable group of automorphisms of the cross-sectional hexagonal lattice. The structure of this group is explored. Examples of open helical-like and closed toroidal-like bundles are presented. An expression for the elastic energy of a perfectly packed bundle of thin elastic rods is derived. The energy accounts for both the bending and torsional stiffnesses of the rods. It is shown that equilibria of the bundle correspond to solutions of a variational problem formulated for the curve representing the axis of the bundle. The functional involves a function of the squared curvature under the constraints on the total torsion and the length. The Euler-Lagrange equations are obtained in terms of curvature and torsion and due to the existence of the first integrals the problem is reduced to the quadrature. The three-dimensional shape of the bundle may be readily reconstructed by integration of the Ilyukhin-type equations in special cylindrical coordinates. The results are of universal nature and are applicable to various fibrous structures, in particular, to intramolecular liquid crystals formed by DNA condensed in toroids or packed inside the viral capsids

  13. Development of design technology on thermal-hydraulic performance in tight-lattice rod bundles. II-rod bowing effect on boiling transition

    International Nuclear Information System (INIS)

    Liu, Wei; Tamai, Hidesada; Kureta, Masatoshi; Ohnuki, Akira; Takase, Kazuyuki; Akimoto, Hajime

    2007-01-01

    A thermal-hydraulic feasibility project for an Innovative Water Reactor for Flexible fuel cycle (FLWR) has been performed since 2002. In this R and D project, large-scale thermal-hydraulic tests, several model experiments and development of advanced numerical analysis codes have been carried out. In this paper, we will describe the critical power characteristics in a 37-rod tight-lattice bundle with rod-bowing under both steady and transient states. It is observed that no matter it is run under a steady or a transient state, boiling transition (BT) always occurs axially at exit elevation of upper high-heat-flux region and transversely in the central area of the bundle. Steady critical power increases monotonically with the increase of mass velocity, with the decrease of inlet water temperature and with the decrease of exit pressure. These trends are same as those in the base case test without rod-bowing. The steady critical power with rod-bowing is about 10% lower than that without rod-bowing. For the postulated power increase and flow decrease cases that may be possibly met in a normal operation of the FLWR, it is confirmed that no BT occurs when Initial Critical Power Ratio (ICPR) is 1.3. Moreover, when the transitions are run under severer ICPR that causes BT, the transient critical powers are generally same as the steady ones. The experiments are analyzed with TRAC-BF1 code. The TRAC-BF1 code shows good prediction for the occurrence or the non occurrence of the BT and predicts the BT starting time within the accuracy of critical power correlation. Traditional quasi - steady state prediction of the transient BT is confirmed being applicable for the postulated abnormal transient processes in the tight lattice bundle with rod - bowing. (author)

  14. 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)

  15. Thermophysical instruments for non-destructive examination of tightness and internal gas pressure or irradiated power reactor fuel rods

    International Nuclear Information System (INIS)

    Pastoushin, V.V.; Novikov, A.Yu.; Bibilashvili, Yu.K.

    1998-01-01

    The developed thermophysical method and technical instruments for non-destructive leak-tightness and gas pressure inspection inside irradiated power reactor fuel rods and FAs under poolside and hot cell conditions are described. The method of gas pressure measuring based on the examination of parameters of thermal convection that aroused in gas volume of rod plenum by special technical instruments. The developed method and technique allows accurate value determination of not only one of the main critical rod parameters, namely total internal gas pressure, that forms rod mean life in the reactor core, but also the partial pressure of every main constituent of gaseous mixture inside irradiated fuel rod, that provides the feasibility of authentic and reliable leak-tightness detection. The described techniques were experimentally checked during the examination of all types power reactor fuel rods existing in Russia (WWER, BN, RBMK) and could form the basis for new technique development for non-destructive examination of PWR (and other) type rods and FAs having gas plenum filled with spring or another elements of design. (author)

  16. 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 9 discusses the following topics: Integrated System Normal Operations Test Results and Analysis Report; Integrated System Off-Normal Operations Test Results and Analysis Report; and Integrated System Maintenance Operations Test Results and Analysis Report

  17. 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 1 discusses the following topics: the background of the project; test program description; summary of tests and test results; problem evaluation; functional requirements confirmation; recommendations; and completed test documentation for tests performed in Phase 3

  18. Prototypical Rod Construction 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 3 discusses the following topics: Downender Test Results and Analysis Report; NFBC Canister Upender Test Results and Analysis Report; Fuel Assembly Handling Fixture Test Results and Analysis Report; and Fuel Canister Upender Test Results and Analysis Report

  19. BWR fuel assembly having fuel rod spacers axially positioned by exterior springs

    International Nuclear Information System (INIS)

    Taleyarkhan, R.P.

    1988-01-01

    In a fuel assembly having spaced fuel rods, an outer hollow tubular flow channel surrounding the fuel rods so as to direct flow of coolant/moderator fluid there-along, and at least one spacer being disposed along the channel and about the fuel rods so as to maintain them in side-by-side spaced relationship, an arrangement for disposing the spacer in a desired axial position along the fuel rods is described comprising: yieldably resilient springs disposed between an interior side of the outer channel and an exterior side of the spacer. The springs have an inherent spring bias directed away from the exterior sides of the spacers and toward the interior side of the channel such that by contact with the channel and spacer the springs assume states in which they are deflected away from the channel interior side so as to exert sufficient compressive contacting force thereon to maintain the spacer substantially stationary in the desired axial position along the fuel rods

  20. Analysis of Double-encapsulated Fuel Rods

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-09-01

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

  1. Parametric study of a reactivity accident in a pressurized water reactor: control rod cluster ejection

    International Nuclear Information System (INIS)

    Chesnel, A.

    1985-01-01

    This research thesis concerns a class 4 accident in a PWR: the ejection of a control rod cluster from the reactor core. It aims at defining, for such an accident, the envelope values which relate the reactivity to the hot spot factor within the frame of a mode A control. The report describes the physical phenomena and their modelling during the considered transient. It presents a simple mathematical solution of the accident which shows that the main neutron parameters are the released reactivity, the delayed neutron fraction, the Doppler coefficient, and the hot spot factor. It reports a temperature sensitivity study, and discusses three-dimensional calculations of irradiation distributions

  2. 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.)

  3. 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)

  4. Simulation of the control rod drop under seismic excitations. Experimental program

    International Nuclear Information System (INIS)

    Chaudat, Th.

    2001-01-01

    This paper describes the experimental program that will be performed at the end of 1998 at the CEA Saclay on a specially constructed analytical mock-up of a control rod. The purpose of these tests is to partially validate the current methodology of the drop time numerical calculations of a PWR (pressurized water reactor) control rod under seismic excitations. The French nuclear partners (EDF and FRAMATOME) are involved in this program. (author)

  5. The analysis of solutions behaviour of Van der Pol Duffing equation describing local brain hemodynamics

    Science.gov (United States)

    Cherevko, A. A.; Bord, E. E.; Khe, A. K.; Panarin, V. A.; Orlov, K. J.

    2017-10-01

    This article proposes the generalized model of Van der Pol — Duffing equation for describing the relaxation oscillations in local brain hemodynamics. This equation connects the velocity and pressure of blood flow in cerebral vessels. The equation is individual for each patient, since the coefficients are unique. Each set of coefficients is built based on clinical data obtained during neurosurgical operation in Siberian Federal Biomedical Research Center named after Academician E. N. Meshalkin. The equation has solutions of different structure defined by the coefficients and right side. We investigate the equations for different patients considering peculiarities of their vessel systems. The properties of approximate analytical solutions are studied. Amplitude-frequency and phase-frequency characteristics are built for the small-dimensional solution approximations.

  6. Conceptual design of stepper motor replacing servo motor for control rod controller

    International Nuclear Information System (INIS)

    Mohd Dzul Aiman Aslan; Mohd Idris Taib; Izhar Abu Hussin; Mohd Khairulezwan Abdul Manan; Mohd Sabri Minhat

    2010-01-01

    In PUSPATI TRIGA Reactor, current control rod controller are using servo motor to control the movement. Control rod is a very important safety element and measure in every nuclear reactor. So, precision is very important in measurement of security in the nuclear reactor. In this case, there are a few disadvantages when using the servo motor is measurement of the motor is not precise. One solution to overcome this is by shifting servo motor with stepper motor. A stepper motor (or step motor) is a brush less, synchronous electric motor that can divide a full rotation into a large number of steps. (author)

  7. Fuel Rod Flow-Induced Vibration Overview

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kang Hee; Kang, Heung Seok; Kim, Hyung Kyu [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-10-15

    To ensure fuel design safety and structural integrity requires the response prediction of fuel rod to reactor coolant flow excitation. However, there are many obstacles in predicting the response as described. Even if the response can be predicted, the design criteria on wear failure, including correlation with the vibration, may be difficult to establish because of a variety of related parameters, such as material, surface condition and environmental factors. Thus, a prototype test for each new fuel assembly design, i.e. a long-term endurance test, is performed for design validation with respect to flow-induced vibration (FIV) and wear. There are still needs of theoretical prediction methods for the response and anticipated failure. This paper revisits the general aspect on the response prediction, mathematical description, analysis procedure and wear correlation aspect of fuel rod's FIV

  8. System and method for consolidating spent fuel rods

    International Nuclear Information System (INIS)

    Baudro, T.O.

    1987-01-01

    A system is described for consolidating spent fuel rods from spent fuel assemblies, comprising: a consolidation container in which the fuel rods may be packed; a frame capable of holding a fuel assembly and the container during consolidation, the frame permitting each of the fuel assembly and the container to be removed; tool means with gripper means for gripping and releasing a rod, the tool means including means for moving the gripper means upwardly and downwardly; a first indexing head having first guide means for guiding the gripper means while the gripper means moves downwardly; a first rail, the first indexing head being slidably mounted on the first rail; a second indexing head having second guide means for guiding the gripper means while the gripper means moves downwardly; a second rail, the second indexing head being slidably mounted on the second rail; and a third rail, the first rail and the second rail being slidably mounted on the third rail; wherein the first indexing head is slidable on the first and third rails to a first position that is above a preselected rod in the fuel assembly; and wherein the second indexing head is slidable on the second and third rails to a second position that is above a preselected location in the container

  9. Numerical solution of the transport equation describing the radon transport from subsurface soil to buildings

    International Nuclear Information System (INIS)

    Savovic, S.; Djordjevich, A.; Ristic, G.

    2012-01-01

    A theoretical evaluation of the properties and processes affecting the radon transport from subsurface soil into buildings is presented in this work. The solution of the relevant transport equation is obtained using the explicit finite difference method (EFDM). Results are compared with analytical steady-state solution reported in the literature. Good agreement is found. It is shown that EFDM is effective and accurate for solving the equation that describes radon diffusion, advection and decay during its transport from subsurface to buildings, which is especially important when arbitrary initial and boundary conditions are required. (authors)

  10. Reactivity shimming in FFTF with safety rods

    International Nuclear Information System (INIS)

    Bennett, C.L.

    1986-01-01

    The Fast Flux Test Facility (FFTF) reactor demonstrated operation at a 65% annual capacity factor in 1984 and a 70% annual capacity factor in 1985. This was achieved in part by extending the cycle lengths beyond the originally designed 100-day cycles. The advent of extended cycle lengths at the FFTF reactor led to the development of a contingency plan to shim some of the installed excess reactivity at the beginning of the cycle with the primary control rod bank (safety rods) in order to maintain the minimum required shutdown margin in the secondary shutdown system. This paper describes both the implementation of this plan in terms of the licensing aspects and the actual use of primary shim during cycles 5 and 6 operation at FFTF

  11. 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

  12. 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.)

  13. Summary of the fuel rod support system (grids) design for LWBR (LWBR development program)

    International Nuclear Information System (INIS)

    Richardson, K.D.

    1979-02-01

    Design features of the fuel rod support system (grids) for the Light Water Breeder Reactor (LWBR) installed in the Shippingport Atomic Power Station, Shippingport, Pennsylvania, are described. The grids are fabricated from AM-350 stainless steel and provide lateral support of the fuel rods in the three regions (seed, blanket, and reflector) of the reactor. A comparison is made of the LWBR grids, whose cells are arranged in triangular-pitched arrays, with rod support systems employed in commercial light water reactors

  14. A user input manual for single fuel rod behaviour analysis code FEMAXI-III

    International Nuclear Information System (INIS)

    Saito, Hiroaki; Yanagisawa, Kazuaki; Fujita, Misao.

    1983-03-01

    Principal objectives of Safety related research in connection with lighr water reactor fuel rods under normal operating condition are mainly addressed 1) to assess fuel integrity under steady state condition and 2) to generate initial condition under hypothetical accident. These assessments have to be relied principally upon steady state fuel behaviour computing code that is able to calculate fuel conditions to tbe occurred in a various manner. To achieve these objectives, efforts have been made to develope analytical computer code that calculates in-reactor fuel rod behaviour in best estimate manner. The computer code developed for the prediction of the long-term burnup response of single fuel rod under light water reactor condition is the third in a series of code versions:FEMAMI-III. The code calculates temperature, rod internal gas pressure, fission gas release and pellet-cladding interaction related rod deformation as a function of time-dependent fuel rod power and coolant boundary conditions. This document serves as a user input manual for the code FEMAMI-III which has opened to the public in year of 1982. A general description of the code input and output are included together with typical examples of input data. A detailed description of structures, analytical submodels and solution schemes in the code shall be given in the separate document to be published. (author)

  15. 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.)

  16. Magnetic switch for reactor control rod

    International Nuclear Information System (INIS)

    Germer, J.H.

    1986-01-01

    This patent describes a control rod system for a nuclear reactor utilizing an electromagnetic grapple mechanism for holding and releasing a control rod, the improvement comprising a magnetic reed switch assembly having a Curie-point magnetic shunt and responsive to reactor coolant temperature for short circuiting the electromagnetic grapple mechanism causing release of the control rod when the coolant temperature reaches the Curie-point of the magnetic shunt. The magnetic reed switch assembly includes a: a permanent magnet, a pair of magnetic pole pieces located at and in contact with opposite ends of the permanent magnet, the Curie-point magnetic shunt being positioned adjacent the permanent magnet and in contact with the pair of magnetic pole pieces, and a reed switch positioned intermediate the pole pieces and provided with a pair of ferromagnetic reeds, a nonmagnetic enclosure around the reeds, a first of the reeds being secured at one end to a first of the pair of pole pieces, a second of the reeds having one end extending into and secured to a hollow member positioned in and extending through a second of the pair of pole pieces, the one end of the second of the reeds secured to a condector adapted to be connected to the electromagnetic grapple mechanism

  17. 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

  18. Tensile Characterization of FRP Rods for Reinforced Concrete Structures

    Science.gov (United States)

    Micelli, F.; Nanni, A.

    2003-07-01

    The application of FRP rods as an internal or external reinforcement in new or damaged concrete structures is based on the development of design equations that take into account the mechanical properties of FRP material systems.The measurement of mechanical characteristics of FRP requires a special anchoring and protocol, since it is well known that these characteristics depend on the direction and content of fibers. In this study, an effective tensile test method is described for the mechanical characterization of FRP rods. Twelve types of glass and carbon FRP specimens with different sizes and surface characteristics were tested to validate the procedure proposed. In all, 79 tensile tests were performed, and the results obtained are discussed in this paper. Recommendations are given for specimen preparation and test setup in order to facilitate the further investigation and standardization of the FRP rods used in civil engineering.

  19. Simulation of fuel rod behaviour during various break LOCAs in PWRs

    International Nuclear Information System (INIS)

    Gadalla, A.A.; El-Fawal, M.M.

    1996-01-01

    During loss of coolant accident (LOCAs) course of events, attention focuses on fuel rod cladding temperature behaviour. In this study, the DRUFAN analytical model and LOBI-MOD2 experimental modeling scheme for fuel rod temperature behaviour during C L-Break LOCA in PWRs, are described and discussed. These models are applied for the investigation of fuel rod cladding temperature behaviour during LOCA blowdown phase. A spectrum of selected values representing small, intermediate and large CL- Break sizes are considered in the predictions. The results of the predictions demonstrated that calculated heater rod temperature at steady state as well as the transient period up to 1000 sec are going in good agreement with the measured values. However above 1000 sec the calculated temperatures are higher than the measured values. This indicates that code predictions in this period are conservative. The results indicated also that, in case of small CL-break LOCA (0.01 A and 0.01 and 0.03 A), the heater rod cladding temperature don't rise above saturation temperature. However, on the top of the heater rod, DNB is occurred in case of 0.03 A CL break, while for 0.01 A break, DNB didn't occur. In case of intermediate and large CL-break; (0.05 A, 0.10 A and 1 A), the results showed that, the heater rod cladding temperature exceeded the saturation temperature and DNB prevailed in upper and intermediate sections of the core. 15 figs., 2 tabs

  20. 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.)

  1. 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.)

  2. 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

  3. 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.)

  4. Growth and Morphology of Rod Eutectics

    Energy Technology Data Exchange (ETDEWEB)

    Jing Teng; Shan Liu; R. Trivedi

    2008-03-17

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

  5. Effect of local heat flux spikes on DNB in non-uniformly heated rod bundles

    International Nuclear Information System (INIS)

    Cadek, F.F.; Hill, K.W.; Motley, F.E.

    1975-02-01

    High pressure water tests were carried out to measure the DNB heat flux using an electrically heated rod bundle in which three adjacent rods had 20 percent heat flux spikes at the axial location where DNB is most likely to occur. This test series was run at the same conditions as those of two earlier test series which had unspiked rods, so that spiked and unspiked runs could be paired and spike effects could thus be isolated. Results are described. 7 references. (U.S.)

  6. 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

  7. 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.)

  8. 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.)

  9. 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.)

  10. First start-up of nuclear criticality safety experiment facility for uranyl nitrate solution

    International Nuclear Information System (INIS)

    Zhu Qingfu; Shi Yongqian; Shen Leisheng; Hu Dingsheng; Zhao Shouzhi; He Tao; Sun Zheng; Lin Shenghuo; Yao Shigui

    2005-01-01

    The uranyl nitrate solution experiment facility for the research on nuclear criticality safety is described. The nuclear fuel loading steps in the first start-up for water-reflected core are presented. During the experiments, the critical volume of uranyl nitrate solution was determined as 20479.62 mL with count rate inverse extrapolation method, reactivity interpolation method, and steady power method. By calculation, critical mass of 235 U was derived as 1579.184 g from experimental data. The worth of control rods was also calibrated in the first start-up of the facility. (authors)

  11. Development of numerical solution techniques in the KIKO3D code

    International Nuclear Information System (INIS)

    Panka, Istvan; Kereszturi, Andras; Hegedus, Csaba

    2005-01-01

    The paper describes the numerical methods applied in KIKO3D three-dimensional reactor dynamics code and present a new, more effective method (Bi-CGSTAB) for accelerating the large sparse matrix equation solution. The convergence characteristics were investigated in a given macro time step of a Control Rod Ejection transient. The results obtained by the old GMRES and new Bi-CGSTAB methods are compared. It is concluded that the real relative errors of the solutions obtained by GMRES or Bi - CGSTAB algorithms are in fact closer together than the estimated relative errors. The KIKO3D-Bi-CGSTAB method converges safely and it is 7-12 % faster than the old KIKO3D-GMRES solution (Authors)

  12. 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

  13. 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

  14. The experimental development and performance test of the pneumatic control-rod drive for the THTR

    International Nuclear Information System (INIS)

    Lange, G.; Boehlo, D.; Heim, H.; Kleine-Tebbe, A.

    1976-01-01

    Reactor control and shutdown of the THTR is accomplished by two independent systems, the first consisting of 36 absorber rods penetrating the graphite reflector region surrounding the core, the second consisting of 42 absorber rods that insert directly into the pebble bed core. This paper describes the design development and testing of the pneumatic rod drives used for movement of the 42 core control rods. The core control rods have two functions: the first, for reactor safety purposes, provides for adequate safe shutdown of the reactor under cold conditions; the second, for operational purposes, provides for compensation of slow changes in reactivity. The safety and operational functions for each absorber rod are respectively carried out by a long-stroke-piston pneumatic drive and by a stepping-piston pneumatic drive, both of these independent, helium-driven drives being incorporated in the rod drive unit for each control rod. To study the performance of the rod drive, a complete prototype control rod and rod drive unit was built and tested under simulated reactor operational conditions. Operational experience under helium temperatures and pressures was gained and the drives were tested under stress and simulated accident conditions. The reliability of this system has been demonstrated to licensing authorities and to the customer. The programme will be completed with the commissioning tests of drives for the THTR-300 reactor. (author)

  15. Method for placing fuel rods in individual cells, and device for performing this procedure

    Energy Technology Data Exchange (ETDEWEB)

    Jabsen, F S

    1972-10-16

    A lattice grid is described in which an egg box type assembly is formed by metal plates deformed to form spring loaded spacers which retain the fuel pins in their correct position. In order to be able to insert the fuel pins without causing scratches on their surfaces, which could lead to corrosion, the springs are displaced outwards by inserting and rotating a square-sectioned rod with rounded corners which when rotated acts as a cam, pressing the springs out. The springs are held in this position by inserting keys horizontally between the lattice plates, through holes for this purpose. The cam rod is then withdrawn, the fuel pins inserted, and the keys withdrawn. Hydraulic equipment for carrying out these operations for a large number of fuel rods simultaneously is also described.

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

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  17. Measurements of Burnout Conditions for Flow of Boiling Water in Vertical 3-Rod and 7-Rod Clusters

    Energy Technology Data Exchange (ETDEWEB)

    Becker, Kurt M; Hernborg, G; Flinta, J E

    1964-08-15

    The present report deals with measurements of burnout conditions for flow of boiling water in vertical 3-rod and 7-rod clusters. Data were obtained,in respect of heating the rods only, as well as for simultaneous uniform and non-uniform heating of the rods and the shroud. Totally, 520 runs were performed. In the case of equal heat fluxes on all surfaces of the channels, burnout always occurred on the rods, and the data were low by a factor of about 1.3 compared with round duct data. When only the rods were heated, the data showed very low burnout values in comparison with the results for total uniform heating and round ducts. This disagreement was explained by considering the climbing film flow model and the fact that only a fraction of the channel perimeter was heated. For simultaneous and non-uniform heating of the rods and the shroud it was found that the shroud could be overloaded up to 50 per cent without reducing the margin of safety in respect of burnout for the rod cluster. Finally, a correlation for predicting burnout conditions in round ducts, annuli and rod clusters has been presented. This correlation predicts the burnout heat fluxes for the present measurements and previously obtained annuli measurements within {+-} 5 per cent.

  18. Development of Welding and Instrumentation Technology for Nuclear Fuel Test Rod

    International Nuclear Information System (INIS)

    Joung, Chang Young; Ahn, Sung Ho; Heo, Sung Ho; Hong, Jin Tae; Kim, Ka Hye

    2013-01-01

    It is necessary to develop various types of welding, instrumentation and helium gas filling techniques that can conduct TIG spot welding exactly at a pin-hole of the end-cap on the nuclear fuel rod to fill up helium gas. The welding process is one of the most important among the instrumentation processes of the nuclear fuel test rod. To manufacture the nuclear fuel test rod, a precision welding system needs to be fabricated to develop various welding technologies of the fuel test rod jointing the various sensors and end-caps on a fuel cladding tube, which is charged with fuel pellets and component parts. We therefore designed and fabricated an orbital TIG welding system and a laser welding system. This paper describes not only some experiment results from weld tests for the parts of a nuclear fuel test rod, but also the contents for the instrumentation process of the dummy fuel test rod installed with the C-type T. C. A dummy nuclear fuel test rod was successfully fabricated with the welding and instrumentation technologies acquired with various tests. In the test results, the round welding has shown a good weldability at both the orbital TIG welding system and the fiber laser welding system. The spot welding to fill up helium gas has shown a good welding performance at a welding current of 30A, welding time of 0.4 sec and gap of 1 mm in a helium gas atmosphere. The soundness of the nuclear fuel test rod sealed by a mechanical sealing method was confirmed by helium leak tests and microstructural analyses

  19. Development of Welding and Instrumentation Technology for Nuclear Fuel Test Rod

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-10-15

    It is necessary to develop various types of welding, instrumentation and helium gas filling techniques that can conduct TIG spot welding exactly at a pin-hole of the end-cap on the nuclear fuel rod to fill up helium gas. The welding process is one of the most important among the instrumentation processes of the nuclear fuel test rod. To manufacture the nuclear fuel test rod, a precision welding system needs to be fabricated to develop various welding technologies of the fuel test rod jointing the various sensors and end-caps on a fuel cladding tube, which is charged with fuel pellets and component parts. We therefore designed and fabricated an orbital TIG welding system and a laser welding system. This paper describes not only some experiment results from weld tests for the parts of a nuclear fuel test rod, but also the contents for the instrumentation process of the dummy fuel test rod installed with the C-type T. C. A dummy nuclear fuel test rod was successfully fabricated with the welding and instrumentation technologies acquired with various tests. In the test results, the round welding has shown a good weldability at both the orbital TIG welding system and the fiber laser welding system. The spot welding to fill up helium gas has shown a good welding performance at a welding current of 30A, welding time of 0.4 sec and gap of 1 mm in a helium gas atmosphere. The soundness of the nuclear fuel test rod sealed by a mechanical sealing method was confirmed by helium leak tests and microstructural analyses.

  20. Development of a control rod drive

    International Nuclear Information System (INIS)

    1991-01-01

    In the period under review, the computer codes required for transients calculation have been completed, as well as the programs for modelling and testing the hot-gas temperature control by means of combined core rod and reflector rod operation. The specification of requirements to be fulfilled by the rod drive computer and the neutron flux measuring system has been done relying essentially on the data obtained by the transients calculations performed and the resulting informations on operating conditions. The work for optimization of the core rod drive with regard to rod driving speeds and the 'three-point switch' with hysteresis for controlled, automatic core rod operation has been concentrating on the case of specified, normal operation of the reactor. (orig./DG) [de

  1. In-Situ Growth and Characterization of Indium Tin Oxide Nanocrystal Rods

    Directory of Open Access Journals (Sweden)

    Yan Shen

    2017-11-01

    Full Text Available Indium tin oxide (ITO nanocrystal rods were synthesized in-situ by a vapor-liquid-solid (VLS method and electron beam evaporation technique. When the electron-beam gun bombarded indium oxide (In2O3 and tin oxide (SnO2 mixed sources, indium and tin droplets appeared and acted as catalysts. The nanocrystal rods were in-situ grown on the basis of the metal catalyst point. The nanorods have a single crystal structure. Its structure was confirmed by X-ray diffraction (XRD and transmission electron microscopy (TEM. The surface morphology was analyzed by scanning electron microscopy (SEM. During the evaporation, a chemical process was happened and an In2O3 and SnO2 solid solution was formed. The percentage of doped tin oxide was calculated by Vegard’s law to be 3.18%, which was in agreement with the mixture ratio of the experimental data. The single crystal rod had good semiconductor switch property and its threshold voltage of single rod was approximately 2.5 V which can be used as a micro switch device. The transmission rate of crystalline nanorods ITO film was over 90% in visible band and it was up to 95% in the blue green band as a result of the oxygen vacancy recombination luminescence.

  2. A parallel multi-domain solution methodology applied to nonlinear thermal transport problems in nuclear fuel pins

    Energy Technology Data Exchange (ETDEWEB)

    Philip, Bobby, E-mail: philipb@ornl.gov [Oak Ridge National Laboratory, One Bethel Valley Road, Oak Ridge, TN 37831 (United States); Berrill, Mark A.; Allu, Srikanth; Hamilton, Steven P.; Sampath, Rahul S.; Clarno, Kevin T. [Oak Ridge National Laboratory, One Bethel Valley Road, Oak Ridge, TN 37831 (United States); Dilts, Gary A. [Los Alamos National Laboratory, PO Box 1663, Los Alamos, NM 87545 (United States)

    2015-04-01

    This paper describes an efficient and nonlinearly consistent parallel solution methodology for solving coupled nonlinear thermal transport problems that occur in nuclear reactor applications over hundreds of individual 3D physical subdomains. Efficiency is obtained by leveraging knowledge of the physical domains, the physics on individual domains, and the couplings between them for preconditioning within a Jacobian Free Newton Krylov method. Details of the computational infrastructure that enabled this work, namely the open source Advanced Multi-Physics (AMP) package developed by the authors is described. Details of verification and validation experiments, and parallel performance analysis in weak and strong scaling studies demonstrating the achieved efficiency of the algorithm are presented. Furthermore, numerical experiments demonstrate that the preconditioner developed is independent of the number of fuel subdomains in a fuel rod, which is particularly important when simulating different types of fuel rods. Finally, we demonstrate the power of the coupling methodology by considering problems with couplings between surface and volume physics and coupling of nonlinear thermal transport in fuel rods to an external radiation transport code.

  3. 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)

  4. Improved Monte Carlo-perturbation method for estimation of control rod worths in a research reactor

    International Nuclear Information System (INIS)

    Kalcheva, Silva; Koonen, Edgar

    2009-01-01

    A hybrid method dedicated to improve the experimental technique for estimation of control rod worths in a research reactor is presented. The method uses a combination of Monte Carlo technique and perturbation theory. Perturbation method is used to obtain the equation for the relative efficiency of control rod insertion. A series of coefficients, describing the axial absorption profile are used to correct the equation for a composite rod, having a complicated burn-up irradiation history. These coefficients have to be determined - by experiment or by using some theoretical/numerical method. In the present paper they are derived from the macroscopic absorption cross-sections, obtained from detailed Monte Carlo calculations by MCNPX 2.6.F of the axial burn-up profile during control rod life. The method is validated on measurements of control rod worths at the BR2 reactor. Comparison with direct MCNPX evaluations of control rod worths is also presented

  5. 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.)

  6. 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.)

  7. Low reactivity penalty burnable poison rods

    International Nuclear Information System (INIS)

    1978-01-01

    A nuclear reactor burnable poison rod is described which consists of an elongated tubular sheath enclosing a neutron absorbing material which, at least during reactor operation, also encloses a neutron moderating material. The excess reactivity existing at the beginning of core life is compensated for by the depletion of the burnable poison throughout the life of the core, so that the life of the core is extended. (UK)

  8. 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

  9. 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

  10. 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.)

  11. 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

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

    International Nuclear Information System (INIS)

    1997-03-01

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

  13. Measurements of Burnout Conditions for Flow of Boiling Water in Vertical 3-Rod and 7-Rod Clusters

    International Nuclear Information System (INIS)

    Becker, Kurt M.; Hernborg, G.; Flinta, J.E.

    1964-08-01

    The present report deals with measurements of burnout conditions for flow of boiling water in vertical 3-rod and 7-rod clusters. Data were obtained,in respect of heating the rods only, as well as for simultaneous uniform and non-uniform heating of the rods and the shroud. Totally, 520 runs were performed. In the case of equal heat fluxes on all surfaces of the channels, burnout always occurred on the rods, and the data were low by a factor of about 1.3 compared with round duct data. When only the rods were heated, the data showed very low burnout values in comparison with the results for total uniform heating and round ducts. This disagreement was explained by considering the climbing film flow model and the fact that only a fraction of the channel perimeter was heated. For simultaneous and non-uniform heating of the rods and the shroud it was found that the shroud could be overloaded up to 50 per cent without reducing the margin of safety in respect of burnout for the rod cluster. Finally, a correlation for predicting burnout conditions in round ducts, annuli and rod clusters has been presented. This correlation predicts the burnout heat fluxes for the present measurements and previously obtained annuli measurements within ± 5 per cent

  14. 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 5 discusses the following topics: Lower Cutting System Test Results and Analysis Report; NFBC Loading System Test Results and Analysis Report; Robotic Bridge Transporter Test Results and Analysis Report; RM-10A Remotec Manipulator Test Results and Analysis Report; and Manipulator Transporter Test Results and Analysis Report

  15. Long-rod penetration: the transition zone between rigid and hydrodynamic penetration modes

    Directory of Open Access Journals (Sweden)

    Jian-feng Lou

    2014-06-01

    Full Text Available Long-rod penetration in a wide range of velocity means that the initial impact velocity varies in a range from tens of meters per second to several kilometers per second. The long rods maintain rigid state when the impact velocity is low, the nose of rod deforms and even is blunted when the velocity gets higher, and the nose erodes and fails to lead to the consumption of long projectile when the velocity is very high due to instantaneous high pressure. That is, from low velocity to high velocity, the projectile undergoes rigid rods, deforming non-erosive rods, and erosive rods. Because of the complicated changes of the projectile, no well-established theoretical model and numerical simulation have been used to study the transition zone. Based on the analysis of penetration behavior in the transition zone, a phenomenological model to describe target resistance and a formula to calculate penetration depth in transition zone are proposed, and a method to obtain the boundary velocity of transition zone is determined. A combined theoretical analysis model for three response regions is built by analyzing the characteristics in these regions. The penetration depth predicted by this combined model is in good agreement with experimental result.

  16. Improved Monte Carlo - Perturbation Method For Estimation Of Control Rod Worths In A Research Reactor

    International Nuclear Information System (INIS)

    Kalcheva, Silva; Koonen, Edgar

    2008-01-01

    A hybrid method dedicated to improve the experimental technique for estimation of control rod worths in a research reactor is presented. The method uses a combination of Monte Carlo technique and perturbation theory. The perturbation theory is used to obtain the relation between the relative rod efficiency and the buckling of the reactor with partially inserted rod. A series of coefficients, describing the axial absorption profile are used to correct the buckling for an arbitrary composite rod, having complicated burn up irradiation history. These coefficients have to be determined - by experiment or by using some theoretical/numerical method. In the present paper they are derived from the macroscopic absorption cross sections, obtained from detailed Monte Carlo calculations by MCNPX 2.6.F of the axial burn up profile during control rod life. The method is validated on measurements of control rod worths at the BR2 reactor. Comparison with direct Monte Carlo evaluations of control rod worths is also presented. The uncertainties, arising from the used approximations in the presented hybrid method are discussed. (authors)

  17. Apparatus for installing and removing a control rod drive in a nuclear reactor

    International Nuclear Information System (INIS)

    Turner, A.P.L.; Ward, R.

    1989-01-01

    This patent describes an apparatus for installing and removing a control rod drive from beneath the pressure vessel of a nuclear reactor. It consists of elevator carriage for carrying the control rod drive into and out of the region beneath the pressure vessel in a generally horizontal position, an elevator cradle mounted on the carriage for pivotal movement about an axis between horizontal and vertical positions and for vertical movement, when in the vertical position, means for securing the control rod drive to the elevator cradle, and a winch cart movable horizontally between a first position spaced from the pivot axis and a second position near the pivot axis. The cart has a winch cable supporting the lower end of the elevator carriage for moving the elevator carriage and the control rod drive between horizontal and vertical positions on the elevator carriage when the cart is spaced from the pivot axis and for raising and lowering the elevator cradle and the control rod drive when the cart is positioned near the pivot axis. The control rod drive is mounted on the elevator cradle by a bearing permitting rotational and horizontal movement of the control rod drive when the drive is in a vertical position, a swing arm, a pneumatically actuated cylinder in axial alignment with the control rod drive for raising and lowering the control rod drive, and means pivotally mounting the cylinder on the swing arm for movement about an axis spaced from and generally parallel to the vertically extending axis so that the position of the cylinder and the control rod drive can be shifted horizontally about the vertically extending axes

  18. 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)

  19. Reversal in the time order of interactive events: the collision of inclined rods

    International Nuclear Information System (INIS)

    Iyer, Chandru; Prabhu, G M

    2006-01-01

    In the rod and hole paradox as described by Rindler (1961 Am. J. Phys. 29 365-6), a rigid rod moves at high speed over a table towards a hole of the same size. Observations from the inertial frames of the rod and slot are widely different. Rindler explains these differences by the concept of differing perceptions in rigidity. Groen and Johannesen (1993 Eur. J. Phys. 14 97-100) confirmed this aspect by computer simulation where the shapes of the rod are different as observed from the co-moving frames of the rod and slot. Lintel and Gruber (2005 Eur. J. Phys. 26 19-23) presented an approach based on retardation due to speed of stress propagation. In this paper, we consider the situation when two parallel rods collide while approaching each other along a line at an inclination with their axis. The collisions of the top and bottom ends are shown to be reversed in time order as seen from the two co-moving frames. This result is explained by the concept of 'extended present' derived from the principle of relativity of simultaneity

  20. 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.

  1. 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.

  2. 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

  3. Mechanical components design for PWR - control rod drive mechanism

    International Nuclear Information System (INIS)

    Leme, Francisco Louzano; Mattar Neto, Miguel

    2002-01-01

    The Control Rod Drive Mechanism (CRDM) is usually - a high precision - equipment incorporating mechanical and electrical components designed to move the control rods. The 'control rods' refer to all rods or assemblies that are moved to assess the performance of the reactor. The CRDM here presented is the Nut and Lead Screw type. This type is basically a power screw type magnetically coupled to a slow speed reluctance electric motor that provides a means of axially positioning the movable fuel assemblies in the reactor core for purpose of controlling core reactivity. A helically threaded lead screw assembly, comprising one element of power screw, is attached to a movable fuel assemblies. The CRDM usually has closer and more consistent contact with environment peculiar to the reactor than has only other machinery component. This environment includes not only the radiation field of the reactor, but also the temperature, pressure and chemical properties associated with the material used as the coolant for reactor fuel. Specific and special materials are needed because of the above mentioned application. Due to the importance of the above described CRDM functions, this paper will also consider the nuclear functions and their safety classes as well as the CRDM nuclear design criteria. (author)

  4. The Preliminary Study for Numerical Computation of 37 Rod Bundle in CANDU Reactor

    International Nuclear Information System (INIS)

    Jeon, Yu Mi; Park, Joo Hwan

    2010-09-01

    A typical CANDU 6 fuel bundle consists of 37 fuel rods supported by two endplates and separated by spacer pads at various locations. In addition, the bearing pads are brazed to each outer fuel rod with the aim of reducing the contact area between the fuel bundle and the pressure tube. Although the recent progress of CFD methods has provided opportunities for computing the thermal-hydraulic phenomena inside of a fuel channel, it is yet impossible to reflect numerical computations on the detailed shape of rod bundle due to challenges with computing mesh and memory capacity. Hence, the previous studies conducted a numerical computation for smooth channels without considering spacers and bearing pads. But, it is well known that these components are an important factor to predict the pressure drop and heat transfer rate in a channel. In this study, the new computational method is proposed to solve complex geometry such as a fuel rod bundle. Before applying a solution to the problem of the 37 rod bundle, the validity and the accuracy of the method are tested by applying the method to simple geometry. The split channel method has been proposed with the aim of computing the fully shaped CANDU fuel channel with detailed components. The validity was tested by applying the method to the single channel problem. The average temperature have similar values for the considered two methods, while the local temperature shows a slight difference by the effect of conduction heat transfer in the solid region of a rod. Based on the present result, the calculation for the fully shaped 37-rod bundle is scheduled for future work

  5. 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

  6. 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

  7. Transient three-dimensional thermal-hydraulic analysis of nuclear reactor fuel rod arrays: general equations and numerical scheme

    International Nuclear Information System (INIS)

    Wnek, W.J.; Ramshaw, J.D.; Trapp, J.A.; Hughes, E.D.; Solbrig, C.W.

    1975-11-01

    A mathematical model and a numerical solution scheme for thermal-hydraulic analysis of fuel rod arrays are given. The model alleviates the two major deficiencies associated with existing rod array analysis models, that of a correct transverse momentum equation and the capability of handling reversing and circulatory flows. Possible applications of the model include steady state and transient subchannel calculations as well as analysis of flows in heat exchangers, other engineering equipment, and porous media

  8. The lifetime of the control rod drives

    International Nuclear Information System (INIS)

    Avet, B.; Cauquelin, C.

    1989-01-01

    The lifetime of the control rod drives is studied. Their function is to take out or to pull in the control rods. The drive and the experiments carried out, are described. The analysis of the behaviour under operation, the drive inspections and surveyance, are also considered. The results are obtained from: the investigations performed on the fatigue strength of the 900 MW and 1300 MW drives, which allowed to deduce a low of wear and to identify the important aspects to be studied, the measurements of the dynamical stresses of mobile elements and a dynamical calculation model. The study leads to the conclusion that a probabilistic approach is needed for the fatigue damage analysis of some elements. Moreover, a systematic examination is also needed, to verify the agreement betwem the drives calculated aging values and the measured ones [fr

  9. 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

  10. 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.

  11. 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

  12. 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.

  13. 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)

  14. Critical heat flux in tubes and tight hexagonal rod lattices

    International Nuclear Information System (INIS)

    Erbacher, F.J.; Cheng Xu; Zeggel, W.

    1994-01-01

    The critical heat flux (CHF) in small-diameter tubes and in tight hexagonal 7-rod and 37-rod bundles was investigated in the KRISTA test facility, using Freon 12 as the working fluid. The measurements in tubes showed that the influence of the tube diameter on CHF cannot be described as suggested by earlier publications with sufficient accuracy. CHF in bundles is lower than in tubes under comparable conditions. The influence of spacers (grid spacers, wire wraps) on CHF was found to be governed by local steam qualities. A comparison of the test results with some CHF prediction methods showed that the look-up table method reproduces the test results in circular tubes most accurately. Combined with CHF look-up tables, subchannel analysis and Ahmad's fluid-to-fluid scaling law, Freon experiments have proven to be a suitable tool for CHF prediction in water-cooled rod bundles. (orig.) [de

  15. 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.

  16. Fault tolerant, multiplexed control rod position detection and indication system for nuclear power plants

    International Nuclear Information System (INIS)

    Dufek, W.L.; Jelovich, J.J.; Neuner, J.A.

    1977-01-01

    The majority of Westinghouse nuclear plants placed in service thus far have incorporated a Rod Position Indication system based upon an analog design philosophy. This system, while meeting all functional and accuracy requirements, has proven somewhat cumbersome, particularly in the area of initial field calibration and maintenance. This paper describes a new Digital Rod Position Indication system (DRPI) developed for use with pressurized water reactors. The system is based upon a digital design philosophy and meets all previous design constraints and environmental requirements. Further, fault tolerance, improved accuracy, interference from adjacent rods and the elimination of adjustments and calibration has been provided

  17. Effect of detector size and position on measured vibration spectra of strings and rods

    International Nuclear Information System (INIS)

    Lipcsei, S.; Kiss, S.; Por, G.

    1993-04-01

    Weight functions of string and rod vibrations are described by standing and travelling wave models. The effects of detector size and position on the measured vibration spectra was investigated, and the main characteristics of the transfer function were calculated by a simple standing wave model. The theoretical results were compared with data from laboratory rod vibration experiments, and with pressure fluctuation spectra obtained at the Paks Nuclear Power Plant. In addition, some fundamental physical consequences can be made using the theory of superposition of travelling waves and their reflection on clamped rod ends. (R.P.) 5 refs.; 10 figs

  18. Analytical Solution to the Pneumatic Transient Rod System at ACRR

    Energy Technology Data Exchange (ETDEWEB)

    Fehr, Brandon Michael [Georgia Inst. of Technology, Atlanta, GA (United States)

    2016-01-08

    The ACRR pulse is pneumatically driven by nitrogen in a system of pipes, valves and hoses up to the connection of the pneumatic system and mechanical linkages of the transient rod (TR). The main components of the TR pneumatic system are the regulator, accumulator, solenoid valve and piston-cylinder assembly. The purpose of this analysis is to analyze the flow of nitrogen through the TR pneumatic system in order to develop a motion profile of the piston during the pulse and be able to predict the pressure distributions inside both the cylinder and accumulators. The predicted pressure distributions will be validated against pressure transducer data, while the motion profile will be compared to proximity switch data. By predicting the motion of the piston, pulse timing will be determined and provided to the engineers/operators for verification. The motion profile will provide an acceleration distribution to be used in Razorback to more accurately predict reactivity insertion into the system.

  19. Analytical Solution to the Pneumatic Transient Rod System at ACRR

    International Nuclear Information System (INIS)

    Fehr, Brandon Michael

    2016-01-01

    The ACRR pulse is pneumatically driven by nitrogen in a system of pipes, valves and hoses up to the connection of the pneumatic system and mechanical linkages of the transient rod (TR). The main components of the TR pneumatic system are the regulator, accumulator, solenoid valve and piston-cylinder assembly. The purpose of this analysis is to analyze the flow of nitrogen through the TR pneumatic system in order to develop a motion profile of the piston during the pulse and be able to predict the pressure distributions inside both the cylinder and accumulators. The predicted pressure distributions will be validated against pressure transducer data, while the motion profile will be compared to proximity switch data. By predicting the motion of the piston, pulse timing will be determined and provided to the engineers/operators for verification. The motion profile will provide an acceleration distribution to be used in Razorback to more accurately predict reactivity insertion into the system.

  20. 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

  1. A finite element method with contact for tensile analysis in fuel rods

    International Nuclear Information System (INIS)

    Tanajura, C.A.S.; Galeao, A.C.N.R.

    1987-01-01

    Elements for mechanical analysis of fuel rod of a PWR type reactor, are presented. The rod, consists basically in a cylindrical coating of zircalloy which contains pilling of UO 2 pellets, is submitted to strong internal and external pressures, intense temperature gradients and neutron flux. These conditions lead several phenomena in the pellet (swelling, fracture, densification, creep) and in the cladding (embrittlement, corrosion, creep) which undergo deformations leading them to contact the restriction for the interpenetration is included in the problem without restriction by Lagrange multipliers. Considering a non-linear problem, due to the surface of contact to be not known a priori, the numerical solutions were obtained using the finite element method. (M.C.K.) [pt

  2. Eddy-current inspection of high flux isotope reactor nuclear control rods

    International Nuclear Information System (INIS)

    Smith, J.H.; Chitwood, L.D.

    1981-07-01

    Inner control rods for the High Flux Isotope Reactor were nondestructively inspected for defects by eddy-current techniques. During these examinations aluminum cladding thickness and oxide thickness on the cladding were also measured. Special application techniques were required because of the high-radiation levels (approx. 10 5 R/h at 30 cm) present and the relatively large temperature gradients that occurred on the surface of the control rods. The techniques used to perform the eddy-current inspections and the methods used to reduce the associated data are described

  3. 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

  4. Nuclear fuel rod grid spring and dimple structures having chamfered edges for reduced pressure drop

    International Nuclear Information System (INIS)

    De Mario, E.E.

    1990-01-01

    This patent describes a nuclear fuel rod grid including inner and outer straps being interleaved with one another to form a matrix of hollow cells, each cell for receiving one fuel rod and being defined by pairs of opposing wall sections of the straps which wall sections are shared with adjacent cells, each cell having a central longitudinal axis defining a coolant flow direction through the cell, at least fuel rod engaging dimple structure of resiliently yieldable material being integrally formed on each wall section of the inner straps

  5. 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)

  6. Application of tube critical heat flux tables to annuli and rod bundles

    International Nuclear Information System (INIS)

    Ulrych, G.

    1985-01-01

    The purpose of this paper is to show that tables for the critical heat flux (CHF) in tubes have a much wider range of applicability than only to tubes. With the proper choice of a characteristic length replacing the tube diameter as a parameter the validity of the tables can be expanded to more complex geometries. The paper describes how the tables must be applied to annuli or rod bundles. The data base for comparisons is mainly taken from the open literature. For rod bundles the proposed methodology was checked for very different geometries including rod bundles from very tight hexagonal to extremely open square bundle arrays. It is concluded that the tables give reasonable results for a wide range of hydraulic diameters

  7. Process and device for fastening and removing fuel-absorber rods in fuel elements of nuclear reactors

    International Nuclear Information System (INIS)

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

    1980-01-01

    This is concerned with an improvement of the fixing of absorber rods in a nuclear reactor. It is important that the rod should not be damaged during removal from the reactor, and that no particles of material are shed during this process. According to the invention, the rod has a stalk which is pressed into a hole in the star shaped arms and welded in. During removal, the stalk is broken at a preferred position. Details of construction are described. (UWI) [de

  8. Methods for acquiring data in power ramping experiments with WWER fuel rods at high burnup

    Energy Technology Data Exchange (ETDEWEB)

    Bobrov, S N; Grachev, A F; Ovchinnikov, V A; Poliakov, I S; Matveev, N P [Research Inst. of Atomic Reactors, Dimitrovgrad (Russian Federation); Novikov, V V [Institute of Inorganic Materials, Moscow (Russian Federation)

    1997-08-01

    A programme on in-pile test which involve fuel burnup up to 60 MWd/kg and up to 12 fuel rods in the experimental rig is considered. Testing methods with reference to the MIR-M1 reactor are reported. Power ramping regime can be realized either by an increase of the total reactor capacity or by displacement of the nearest to the experimental cell control rods or by combination of these two ways. A total thermal capacity of the fuel rod cluster is determined by means of the thermal balance technique. The thermal capacity of each separate fuel rod can be estimated from the distribution of their relative activity within the accuracy range 5-10%. The important condition for this procedure is to keep the initial distribution of the fuel rod heating during power ramping. Means of instrumentation are described. They are standard detectors of loop facilities and transducers installed both in the irradiation rigs and fuel rods. Different ways of processing data on the fuel rod loss of integrity are reported. When the time of fuel rod loss of tightness is placed in correspondence with its capacity, processing can be made either on the maximum fuel rod heat load or on that at crack location. The information acquired in the experiments on the burnup values, heat rating distribution, kinetics of fission product gas emission, fuel rod elongation, fuel rod diameter changes, crack availability and fission products migration is used for the development and verification of calculation codes. (author). 1 ref., 4 figs, 1 tab.

  9. 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 ...

  10. Test requirement for PIE of HANARO irradiated fuel rod

    International Nuclear Information System (INIS)

    Lim, I. C.; Cho, Y. G.

    2000-06-01

    Since the first criticality of HANARO reached in Feb. of 1995, the rod type U 3 Si-A1 fuel imported from AECL has been used. From the under-water fuel inspection which has been conducted since 1997, a ballooning-rupture type abnormality was observed in several fuel rods. In order to find the root cause of this abnormality and to find the resolution, the post irradiation examination(PIE) was proposed as the best way. In this document, the information from the under-water inspection as well as the PIE requirements are described. Based on the information in this document, a detail test plan will be developed by the project team who shall conduct the PIE

  11. Characteristics of axial splits in failed BWR fuel rods

    International Nuclear Information System (INIS)

    Lysell, G.; Grigoriev, V.

    2000-01-01

    Secondary cladding defects in BWR fuel sometimes have the shape of long axial cracks or ''splits''. Due to the large open UO 2 surfaces exposed to the water, fission product and UO 2 release to the coolant can reach excessive levels leading to forced shut downs to remove the failed fuel rods. A number of such fuel rods have been examined in Studsvik over the last 10 years. The paper describes observations from the PIE of long cracks and discusses the driving force of the cracks. Details such as starting cracks, macroscopic and microscopic fracture surface appearance, cross sections of cracks, hydride precipitates, location and degree of plastic deformation are given. (author)

  12. Method of inspecting control rod drive mechanism

    International Nuclear Information System (INIS)

    Sato, Tomomi; Tatemichi, Shin-ichiro; Hasegawa, Hidenobu.

    1988-01-01

    Purpose: To conduct inspection for control rod drives and fuel handling operations in parallel without taking out the entire fuel, while maintaining the reactor in a subcritical state. Method: Control rod drives are inspected through the release of connection between control rods and control rod drives, detachment and dismantling of control rod drives, etc. In this case, structural materials having neutron absorbing power equal to or greater than the control rods are inserted into the gap after taking out fuels. Since the structural materials have neutron absorbing portion, subcriticality is maintained by the neutron absorbing effect. Accordingly, there is no requirement for taking out all of the fuels, thereby enabling to check the control rod drives and conduct handling for the fuels in parallel. As a result, the number of days required for the inspection can be shortened and it is possible to improve the working efficiency for the decomposition, inspection, etc. of the control rod drives and, thus, improve the operation efficiency of the nuclear power plant thereby attaining the predetermined purpose. (Kawakami, Y.)

  13. 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.)

  14. 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.)

  15. Grid for nuclear fuel rod assembly

    International Nuclear Information System (INIS)

    Brayman, K.W.; George, D.K.; Rawlings, J.C.; Dix, G.E.

    1976-01-01

    A grid is described for placing a least four corner fuel rods in a tubular flow channel of a nuclear reactor. It includes a bearer component composed of four side strips joined by four corner strips so as to form a rigid unit structure, each side strip having an L-shaped piece adjacent at each of its ends to a lug of each L-shaped piece extending to the adjacent end of its associated side strip [fr

  16. Experimental determination of temperature fields in sodium-cooled rod bundles with hexagonal rod arrangement and grid spacers

    International Nuclear Information System (INIS)

    Moeller, R.; Tschoeke, H.; Kolodziej, M.

    1977-01-01

    Three-dimensional temperature fields in the claddings of sodium cooled rods were determined experimentally under representative nominal operating conditions for a SNR typical 19-rod bundle model provided with spark-eroded spacers. These experiments are required to verify thermohydraulic computer programs which will provide the output data for strength calculations of the high loaded cladding tubes. In this work the essentials are reported of the measured circumferential distributions of wall temperatures of peripheral rods. In addition the sub-channel temperatures measured over the bundle cross section are indicated, they are required to sustain codes for the global thermohydraulic design of core elements. The most important results are: 1) The whole fuel element is located within the thermal entrance length. 2) High azimuthal temperature differences were measured in the claddings of peripheral rods, which are strongly influenced by the distance between the rod and the shroud, especially for the corner rod. 3) With decreasing Pe-number ( [de

  17. Stressed and strained state for cermetic-rod-type fuel element

    International Nuclear Information System (INIS)

    Kulikov, I.S.

    1987-01-01

    Calculation technique for designing the stress-strained state of a cermetic rod-type fuel element has been proposed. The technique is based on the time-dependent step-by-step method and the solution of the deformation equilibrium equation for continuous and thick-wall long cylinders at every temporal step by the finite difference method. Additional strains, caused by thermal expansion and radiation swelling, have been taken into account. The transion from the non-contact model to the stiff-contact model has been provided in the case of cladding-fuel gap dissappearing in one or a number of cross-sections along the fuel element height. The method is supplemented by the formula for fuel cans stability estimation in the case of high coolant external pressure. The example of estimation of the cermetic-rod-type fuel elements are considered as an example

  18. Elastic plastic analysis of fuel element assemblies - hexagonal claddings and fuel rods

    International Nuclear Information System (INIS)

    Mamoun, M.M.; Wu, T.S.; Chopra, P.S.; Rardin, D.C.

    1979-01-01

    Analytical studies have been conducted to investigate the structural, thermal, and mechanical behavior of fuel rods, claddings and fuel element assemblies of several designs for a conceptual Safety Test Facility (STF). One of the design objectives was to seek a geometrical configuration for a clad by maximizing the volume fraction of fuel and minimizing the resultant stresses set-up in the clad. The results of studies conducted on various geometrical configurations showed that the latter design objective can be achieved by selecting a clad of an hexagonal geometry. The analytical studies necessitated developing solutions for determining the stresses, strains, and displacements experienced by fuel rods and an hexagonal cladding subjected to thermal fuel-bowing loads acting on its internal surface, the external pressure of the coolant, and elevated temperatures. This paper presents some of the initially formulated analytical methods and results. It should be emphasized that the geometrical configuration considered in this paper may not necessarily be similar to that of the final design. Several variables have been taken into consideration including cladding thickness, the dimensions of the fuel rod, the temperature of the fuel and cladding, the external pressure of the cooling fluid, and the mechanical strength properties of fuel and cladding. A finite-element computer program, STRAW Code, has also been employed to generate several numerical results which have been compared with those predicted by employing the initially formulated solutions. The theoretically predicted results are in good agreement with those of the STRAW Code. (orig.)

  19. 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...

  20. 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.

  1. 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.)

  2. 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.)

  3. Simulation of rod drop experiments in the initial cores of Loviisa and Mochovce

    International Nuclear Information System (INIS)

    Kaloinen, E.; Kyrki-Rajamaeki, R.; Wasastjerna, F.

    1999-01-01

    Interpretation of rod drop measurements during startup tests of the Loviisa reactors has earlier been studied with two-dimensional core calculations using a spatial prompt jump approximation. In these calculations the prediction for the reactivity meter reading was lower than the measured values by 25%. Another approach to solve the problem is simulation of the rod drop experiment with dynamic core calculations coupled with out of core calculations to estimate the response of ex-core ionization chambers for the reactivity meter. This report described the calculations performed with the three-dimensional dynamic code HEXTRAN for prediction of the reactivity meter readings in rod drop experiments in initial cores of the WWER-440 reactors. (Authors)

  4. Nuclear fuel rod grip with modified diagonal spring structures

    International Nuclear Information System (INIS)

    DeMario, E.E.

    1990-01-01

    This patent describes a spring structure in a nuclear fuel rod grid including a plurality of inner and outer straps being interleaved with one another to form a matrix of hollow cells. Each of the cells is for receiving one fuel rod and being defined by pairs of opposing wall sections of the straps which wall sections are shared with adjacent cells. Each of the cells has a central longitudinal axis, a fuel rod engaging spring structure of resiliently yieldable material being integrally formed on each wall section of the inner straps. The spring structure comprising: a pair of spaced apart opposite outer portions being integrally attached at their outer ends to the respective wall section. The portions extending in alignment with one another and in generally diagonal relation to the direction of the central longitudinal axis of the one cell; and a middle portion disposed between and integrally connected at its outer ends with respective inner ends of the outer portions. The middle portion extending in generally transverse relation to the direction of the central longitudinal axis of the one cell

  5. Whole analogy between Daniel Bernoulli solution and direct kinematics solution

    Directory of Open Access Journals (Sweden)

    Filipović Mirjana

    2010-01-01

    Full Text Available In this paper, the relationship between the original Euler-Bernoulli's rod equation and contemporary knowledge is established. The solution which Daniel Bernoulli defined for the simplest conditions is essentially the solution of 'direct kinematics'. For this reason, special attention is devoted to dynamics and kinematics of elastic mechanisms configuration. The Euler-Bernoulli equation and its solution (used in literature for a long time should be expanded according to the requirements of the mechanisms motion complexity. The elastic deformation is a dynamic value that depends on the total mechanism movements dynamics. Mathematical model of the actuators comprises also elasticity forces.

  6. Development of joining techniques for fabrication of fuel rod simulators

    International Nuclear Information System (INIS)

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

    1980-10-01

    Much of the safety-related thermal-hydraulic tests on nuclear reactors are conducted not in the reactor itself, but in mockup segments of a core that uses resistance-heated fuel rod simulators (FRS) in place of the radioactive fuel rods. Laser welding and furnace brazing techniques are described for joining subassemblies for FRS that have survived up to 1000 h steady-state operation at 700 to 1100 0 C cladding temperatures and over 5000 thermal transients, ranging from 10 to 100 0 C/s. A pulsed-laser welding procedure that includes use of small-diameter filler wire is used to join one end of a resistance heating element of Pt-8 W, Fe-22 Cr-5.5 Al-0.5 Co, or 80 Ni-20 Cr (wt %) to a tubular conductor of an appropriate intermediate material. The other end of the heating element is laser welded to an end plug, which in turn is welded to a central conductor rod

  7. 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.)

  8. Intraspinal canal rod migration causing late-onset paraparesis 8 years after scoliosis surgery.

    Science.gov (United States)

    Obeid, Ibrahim; Vital, Jean-Marc; Aurouer, Nicolas; Hansen, Steve; Gangnet, Nicolas; Pointillart, Vincent; Gille, Olivier; Boissiere, Louis; Quraishi, Nasir A

    2016-07-01

    Complete intraspinal canal rod migration with posterior bone reconstitution has never been described in the adolescent idiopathic scoliosis (AIS) population. We present an unusual but significant delayed neurological complication after spinal instrumentation surgery. A 24-year-old woman presented with lower limb weakness (ASIA D) 8 years after posterior instrumentation from T2 to L4 for AIS. CT scan and MRI demonstrated intra-canal rod migration with complete laminar reconstitution. The C-reactive protein was slightly elevated (fluctuated between 10 and 20 mg/l). Radiographs showed the convex rod had entered the spinal canal. The patient was taken into the operating room for thoracic spinal decompression and removal of the convex rod. This Cotrel-Dubousset rod, which had been placed on the convexity of the thoracic curve had completely entered the canal from T5 to T10 and was totally covered by bone with the eroded laminae entirely healed and closed. There was no pseudarthrosis. Intra-operatively, the fusion mass was opened along the whole length of this rod and the rod carefully removed and the spinal cord decompressed. The bacteriological cultures returned positive for Propionibacterium acnes. The patient recovered fully within 2 months post-operatively. We opine that the progressive laminar erosion with intra-canal rod migration resulted from mechanical and infectious-related factors. The very low virulence of the strain of Propionibacterium acnes is probably involved in this particular presentation where the rod was trapped in the canal, owing to the quite extensive laminar reconstitution.

  9. 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.

  10. 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

  11. 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

  12. 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...

  13. 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.)

  14. Long-rod projectiles against oblique targets: analysis and design recommendations

    International Nuclear Information System (INIS)

    Norris, D.M.; McMaster, W.H.; Wilkins, M.L.

    1976-01-01

    Computer calculations provide an understanding of impact phenomena associated with long rod penetrators striking targets at oblique angles. The rod and target material behaviors are described by elastic-plastic work-hardening constitutive models. Ductile fracture is simulated by setting all tensile and shear stresses to zero when the calculated plastic deformation reaches a critical value. The important penetration material properties to defeat a given target are identified. There is no single material property of overriding importance; a combination of properties is required for an efficient penetrator. Experimental results are presented to demonstrate the effects of ductility and toughness on penetration performance. Recommendations for possible improvement are suggested

  15. 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.)

  16. Neutron diffraction analysis of the structure of rod photoreceptor membranes in intact retinas

    Energy Technology Data Exchange (ETDEWEB)

    Yaeger, M; Schoenborn, B; Engelman, D; Moore, P; Stryer, L

    1980-01-01

    Neutron diffraction data have been collected from samples containing ten dark-adapted Rana catesbiana bullfrog retinas in 100, 80, 60, 40, and 30% D/sub 2/O Ringer's solution using a step-scanning Soller slit diffractometer. Diffraction was also recorded from retinas equilibrated in D/sub 2/O solutions with varying osmolarity. Rocking curve experiments demonstrated that the rods are disoriented in a cylindrically symmetrical fashion. Structure factor amplitudes were obtained using semi-automated curve-fitting procedures, and phases were obtained by interpreting the D/sub 2/O-H/sub 2/O and osmotic Patterson maps. In D/sub 2/O Ringer's solution the first four structure factors are -353 +- 25, 246 +- 19, 434 +- 13 and 383 +- 19. Neutron scattering density profiles were calculated to 75 A resolution using these structure factors. These neutron diffraction data are consistent with the view that the lipid bilayer is a major structural motif of the rod outer segment disc membrane. Neutron Fourier syntheses in different mixtures of D/sub 2/O and H/sub 2/O indicate that the intradisc and extradisc spaces are predominantly aqueous, consistent with the increase in the intradisc and extradisc volumes as the Ringer's solution is made more hypotomic. In isotonic Ringer's solution, the thicknesses of the intradisc and extradisc spaces are about 36 A and 160 A, respectively, and the center-to-center separation between the 50 A thick lipid bilayer is 88 A.

  17. 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

  18. Kraus operator solutions to a fermionic master equation describing a thermal bath and their matrix representation

    Science.gov (United States)

    Xiang-Guo, Meng; Ji-Suo, Wang; Hong-Yi, Fan; Cheng-Wei, Xia

    2016-04-01

    We solve the fermionic master equation for a thermal bath to obtain its explicit Kraus operator solutions via the fermionic state approach. The normalization condition of the Kraus operators is proved. The matrix representation for these solutions is obtained, which is incongruous with the result in the book completed by Nielsen and Chuang [Quantum Computation and Quantum Information, Cambridge University Press, 2000]. As especial cases, we also present the Kraus operator solutions to master equations for describing the amplitude-decay model and the diffusion process at finite temperature. Project supported by the National Natural Science Foundation of China (Grant No. 11347026), the Natural Science Foundation of Shandong Province, China (Grant Nos. ZR2013AM012 and ZR2012AM004), and the Research Fund for the Doctoral Program and Scientific Research Project of Liaocheng University, Shandong Province, China.

  19. Quasitravelling waves

    International Nuclear Information System (INIS)

    Beklaryan, Leva A

    2011-01-01

    A finite difference analogue of the wave equation with potential perturbation is investigated, which simulates the behaviour of an infinite rod under the action of an external longitudinal force field. For a homogeneous rod, describing solutions of travelling wave type is equivalent to describing the full space of classical solutions to an induced one-parameter family of functional differential equations of point type, with the characteristic of the travelling wave as parameter. For an inhomogeneous rod, the space of solutions of travelling wave type is trivial, and their 'proper' extension is defined as solutions of 'quasitravelling' wave type. By contrast to the case of a homogeneous rod, describing the solutions of quasitravelling wave type is equivalent to describing the quotient of the full space of impulsive solutions to an induced one-parameter family of point-type functional differential equations by an equivalence relation connected with the definition of solutions of quasitravelling wave type. Stability of stationary solutions is analyzed. Bibliography: 9 titles.

  20. The modeling of fuel rod behaviour under RIA conditions in the code DYN3D

    International Nuclear Information System (INIS)

    Rohde, U.

    1998-01-01

    A description of the fuel rod behaviour and heat transfer model used in the code DYN3D for nuclear reactor core dynamic simulations is given. Besides the solution of heat conduction equations in fuel and cladding, the model comprises detailed description of heat transfer in the gas gap by conduction, radiation and fuel-cladding contact. The gas gap behaviour is modeled in a mechanistic way taking into account transient changes of the gas gap parameters based on given conditions for the initial state. Thermal, elastic and plastic deformations of fuel and cladding are taken into account within 1D approximation. Numerical studies concerning the fuel rod behaviour under RIA conditions in power reactors are reported. Fuel rod behaviour at high pressures and flow rates in power reactors is different from the behaviour under atmospheric pressure and stagnant flow conditions in the experiments. The mechanisms of fuel rod failure for fresh and burned fuel reported from the literature can be qualitatively reproduced by the DYN3D model. (author)

  1. Estimation of the core-wide fuel rod damage during a LWR LOCA

    International Nuclear Information System (INIS)

    Mattila, L.; Sairanen, R.; Stengaard, J.-O.

    1975-01-01

    The number of fuel rods puncturing during a LWR LOCA must be estimated as a part of the plant radioactivity release analysis. Due to the great number of fuel rods in the core and the great number of contributing parameters, many of them associated with wide uncertainty and/or truly random variability limits, probabilistic methods are well applicable. A succession of computer models developed for this purpose is described together with applications to WWER-440 PWR. Deterministic models are shown to be seriously inadequate and even misleading under certain circumstances. A simple analytical probabilistic model appears to be suitable for many applications. Monte Carlo techniques allow the development of such sophisticated models that errors in the input data presently available probably become dominant in the residual uncertainty of the corewide fuel rod puncture analysis. (author)

  2. 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.)

  3. Wave motion in a thick cylindrical rod undergoing longitudinal impact

    Czech Academy of Sciences Publication Activity Database

    Červ, Jan; Adámek, V.; Valeš, František; Gabriel, Dušan; Plešek, Jiří

    2016-01-01

    Roč. 66, November (2016), s. 88-105 ISSN 0165-2125 R&D Projects: GA ČR(CZ) GAP101/12/2315; GA TA ČR(CZ) TH01010772 Institutional support: RVO:61388998 Keywords : elastic waves * impact * thick cylindrical rod * analytical solution * semi-analytical solution Subject RIV: BI - Acoustics Impact factor: 1.575, year: 2016 http://ac.els-cdn.com/S0165212516300427/1-s2.0-S0165212516300427-main.pdf?_tid=d91eee02-7a55-11e6-8c02-00000aab0f6c&acdnat=1473842161_c56543aaec31b7e091ab47d3fb38f361

  4. Analysis of heat transfer from fuel rods with externally attached thermocouples

    International Nuclear Information System (INIS)

    Gill, C.R.; Coddington, P.

    1988-05-01

    This paper describes the development of 2 and 3 dimensional finite element heat conduction models to simulate the behaviour of the external thermocouples attached to the LOFT fuel rods during the blowdown phase of a large break loss-of-coolant accident. To establish the model and determine the thermal coupling between the thermocouple and the fuel rod extensive use was made of two series of experiments performed at INEL in the LOFT Test Support Facility (LTSF). These experiments were high pressure reflood experiments with fluid conditions 'typical' of those seen during the bottom-up flow period of the LOFT experiments. (author)

  5. Model investigation of fuel rod behaviour

    International Nuclear Information System (INIS)

    Girgis, M.M.; Wiesenack, W.; Stegemann, D.

    1985-06-01

    Thermal and mechanical behaviour of fuel rods can be explained but unsatisfactorily by models based of an axial symmetry concept. Recently developed models include, with respect to their thermal components, a simple method for the computation of the temperature distribution within the fuel, and they also take into account the influence of excentrically placed pellets for the computation of heat transfer in the cold gap. Additionally, a finite-element model is used to evaluate the effects of cracking and fragmentation on the thermal behaviour of pellets. The reaction of fuel and fuel cladding to external and internal loadings and the axial interaction between fuel and cladding are described in the mechanical portion of the model. A special case of axial coupling is the so-called random stacking interaction caused by fuel pellets placed excentrically at the cladding and sliding radially and axially. In the comparison of measurement results, both thermal and mechanical behaviour of different rods from the OECD Halden Reactor Project are subject to investigations. (RF) [de

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

    International Nuclear Information System (INIS)

    Rosa, J.M.

    1986-01-01

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

  7. Fabrication of BaCO3 Nano-Rods in Water/poly Vinyl Alcohol Mixed Solvents

    Science.gov (United States)

    Asgari, Motahare; Ataie, Abolghasem; Mokhtabad Amrei, Diba

    Barium carbonate nano-particles have been synthesized by a simple PVA-assisted precipitation method using hydrated barium chloride and sodium carbonate as the starting materials. Crystal structure and morphology of the products were evaluated by XRD and SEM techniques, respectively. XRD results indicated that under the identical experimental conditions, single phase BaCO3 was produced. SEM results showed that in the presence of PVA as guide reagent in the solution, nano-rods with regular (facet) surfaces were obtained. The results also showed that the concentration of PVA in the solution, reaction time and temperature have obvious influence on the agglomeration nature and mean particle size of the products. At high concentration of PVA less agglomerated nano size particles of BaCO3 were synthesized. Increasing the synthesis time and temperature resulted in increasing the average diameter of BaCO3 nano-rods. The performance of produced barium carbonate in the processing of barium hexaferrite ceramic magnet via conventional route has been approved.

  8. 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

  9. 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

  10. 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.

  11. PROCOPE, Collision Probability in Pin Clusters and Infinite Rod Lattices

    International Nuclear Information System (INIS)

    Amyot, L.; Daolio, C.; Benoist, P.

    1984-01-01

    1 - Nature of physical problem solved: Calculation of directional collision probabilities in pin clusters and infinite rod lattices. 2 - Method of solution: a) Gauss integration of analytical expressions for collision probabilities. b) alternately, an approximate closed expression (not involving integrals) may be used for pin-to-pin interactions. 3 - Restrictions on the complexity of the problem: number of fuel pins must be smaller than 62; maximum number of groups of symmetry is 300

  12. 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.

  13. The use of eddy current testing for nuclear fuel rods cladding evaluation

    International Nuclear Information System (INIS)

    Silva Junior, Silverio F. da; Alencar, Donizete A.; Brito, Mucio Jose D. de

    2007-01-01

    Nuclear fuel rods cladding must be tested after their manufacture and during their operational life. This paper describes a study about the use of eddy current test method as a nondestructive tool for nuclear fuel rods cladding evaluation. The experiments were carried out using two different probes: an external probe and an internal probe. The main goal was to verify the sensitivity of the eddy current test system, to develop calibration and reference standards and to establish the main capabilities and limitations presented by this test method for this application. (author)

  14. 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

  15. 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

  16. Towards Viscoplastic Constitutive Models for Cosserat Rods

    Directory of Open Access Journals (Sweden)

    Dörlich Vanessa

    2016-06-01

    Full Text Available Flexible, slender structures like cables, hoses or wires can be described by the geometrically exact Cosserat rod theory. Due to their complex multilayer structure, consisting of various materials, viscoplastic behavior has to be expected for cables under load. Classical experiments like uniaxial tension, torsion or three-point bending already show that the behavior of e.g. electric cables is viscoplastic. A suitable constitutive law for the observed load case is crucial for a realistic simulation of the deformation of a component. Consequently, this contribution aims at a viscoplastic constitutive law formulated in the terms of sectional quantities of Cosserat rods. Since the loading of cables in applications is in most cases not represented by these mostly uniaxial classical experiments, but rather multiaxial, new experiments for cables have to be designed. They have to illustrate viscoplastic effects, enable access to (viscoplastic material parameters and account for coupling effects between different deformation modes. This work focuses on the design of such experiments.

  17. 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.)

  18. Towards Viscoplastic Constitutive Models for Cosserat Rods

    OpenAIRE

    Dörlich Vanessa; Linn Joachim; Scheffer Tobias; Diebels Stefan

    2016-01-01

    Flexible, slender structures like cables, hoses or wires can be described by the geometrically exact Cosserat rod theory. Due to their complex multilayer structure, consisting of various materials, viscoplastic behavior has to be expected for cables under load. Classical experiments like uniaxial tension, torsion or three-point bending already show that the behavior of e.g. electric cables is viscoplastic. A suitable constitutive law for the observed load case is crucial for a realistic simul...

  19. A cold mass support system based on the use of oriented fiberglass epoxy rods in bending

    International Nuclear Information System (INIS)

    Green, Michael A.; Corradi, Carol A.; LaMantia, Roberto F.; Zbasnik, Jon P.

    2002-01-01

    This report describes a cold mass support system that uses oriented fiberglass epoxy (other low heat leak oriented fiber material can also be used) rods. In the direction of the rods, where forces are carried in tension or compression, the support system is very stiff. In the other directions, the rods are subjected to bending stresses. When the support rods are put in bending the cold mass support is quite compliant. This type of support system can be used in situation where space for a cold mass support system is limited and where compliance can be tolerated in at least one direction. Break test data for 15.9-mm and 19.1-mm diameter oriented fiberglass rods is presented in this report. The cold mass supports for the DFBX distribution boxes are presented as an example of this type of cold mass support system

  20. 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.)

  1. 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

  2. MIL-68 (In) nano-rods for the removal of Congo red dye from aqueous solution.

    Science.gov (United States)

    Jin, Li-Na; Qian, Xin-Ye; Wang, Jian-Guo; Aslan, Hüsnü; Dong, Mingdong

    2015-09-01

    MIL-68 (In) nano-rods were prepared by a facile solvothermal synthesis using NaOAc as modulator agent at 100°C for 30 min. The BET test showed that the specific surface area and pore volume of MIL-68 (In) nanorods were 1252 m(2) g(-1) and 0.80 cm(3) g(-1), respectively. The as-prepared MIL-68 (In) nanorods showed excellent adsorption capacity and rapid adsorption rate for removal of Congo red (CR) dye from water. The maximum adsorption capacity of MIL-68 (In) nanorods toward CR reached 1204 mg g(-1), much higher than MIL-68 (In) microrods and most of the previously reported adsorbents. The adsorption process of CR by MIL-68 (In) nano-rods was investigated and found to be obeying the Langmuir adsorption model in addition to pseudo-second-order rate equation. Moreover, the MIL-68 (In) nanorods showed an acceptable reusability after regeneration with ethanol. All information gives an indication that the as-prepared MIL-68 (In) nanorods show their potential as the adsorbent for highly efficient removal of CR in wastewater. Copyright © 2015 Elsevier Inc. All rights reserved.

  3. Vernotte-Cattaneo approximation for heat conduction in fuel rod

    International Nuclear Information System (INIS)

    Espinosa P, G.; Espinosa M, E. G.

    2009-10-01

    In this paper we explore the applicability of a fuel rod mathematical model based on the Vernotte-Cattaneo transient heat conduction as constitutive law (Non-Fourier approach) for light water reactors transient analysis. In the classical theory of diffusion, the Fourier law of heat conduction is used to describe the relation between the heat conduction is used to describe the relation between the heat flux vector and the temperature gradient assuming that the heat propagation speeds are infinite. The motivation for this research was to eliminate the paradox of an infinite. The motivation for this research was to eliminate the paradox of an infinite thermal wave speed. The time-dependent heat sources were considered in the fuel rod heat transfer model. The close of the main steam isolated valves transient in a boiling water reactor was analyzed for different relaxation times. The results show that for long-times the heat fluxes on the clad surface under Vernotte-Cattaneo approach can be important, while for short-times and from the engineering point of view the changes are very small. (Author)

  4. Development of cutting device for irradiated fuel rod

    International Nuclear Information System (INIS)

    Lee, E. P.; Jun, Y. B.; Hong, K. P.; Min, D. K.; Lee, H. K.; Su, H. S.; Kim, K. S.; Kwon, H. M.; Joo, Y. S.; Yoo, K. S.; Joo, J. S.; Kim, E. K.

    2004-01-01

    Post Irradiation Examination(PIE) on irradiated fuel rods is essential for the evaluation of integrity and irradiation performance of fuel rods of commercial reactor fuel. For PIE, fuel rods should be cut very precisely. The cutting positions selected from NDT data are very important for further destructive examination and analysis. A fuel rod cutting device was developed witch can cut fuel rods longitudinal very precisely and can also cut the fuels into the same length rod cuts repeatedly. It is also easy to remove the fuel cutting powder after cutting works and it can extend the life time of cutting device and lower the contamination level of hot cell

  5. 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

  6. Ultrasonic Structural Health Monitoring to Assess the Integrity of Spinal Growing Rods In Vitro.

    Science.gov (United States)

    Oetgen, Matthew E; Goodley, Addison; Yoo, Byungseok; Pines, Darryll J; Hsieh, Adam H

    2016-01-01

    Rod fracture is a common complication of growing rods and can result in loss of correction, patient discomfort, and unplanned revision surgery. The ability to quantitate rod integrity at each lengthening would be advantageous to avoid this complication. We investigate the feasibility of applying structural health monitoring to evaluate the integrity of growing rods in vitro. Single-rod titanium 4.5-mm growing rod constructs (n = 9), one screw proximally and one distally connected by in-line connectors, were assembled with pedicle screws fixed in polyethylene blocks. Proximal and distal ends were loaded and constructs subjected to cyclic axial compression (0-100 N at 1 Hz), with incrementally increasing maximum compressive loads of 10 N every 9k cycles until failure. Four piezoceramic transducers (PZTs) were mounted along the length the constructs to interrogate the integrity of the rods with an ultrasonic, guided lamb wave approach. Every 9k cycles, an 80 V excitatory voltage was applied to a PZT to generate high-frequency vibrations, which, after propagating through the construct, was detected by the remaining PZTs. Amplitude differences between pre- and postload waveform signals were calculated until rod failure. Average construct lifetime was 88,991 ± 13,398 cycles. All constructs failed due to rod fracture within 21 mm (mean = 15 ± 4.5 mm) of a screw or connector. Amplitude differences between pre- and postload increased in a stepwise fashion as constructs were cycled. Compared to baseline, we found a 1.8 ± 0.6-fold increase in amplitude 18k cycles before failure, a 2.2 ± 1.0-fold increase in amplitude 9k cycles before failure, and a 2.75 ± 1.5-fold increase in amplitude immediately before rod fracture. We describe a potential method for assessing the structural integrity of growing rods using ultrasonic structural health monitoring. These preliminary data demonstrate the ability of periodic rod assessment to detect structural changes in cycled growing

  7. 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

  8. Control rod computer code IAMCOS: general theory and numerical methods

    International Nuclear Information System (INIS)

    West, G.

    1982-11-01

    IAMCOS is a computer code for the description of mechanical and thermal behavior of cylindrical control rods for fast breeders. This code version was applied, tested and modified from 1979 to 1981. In this report are described the basic model (02 version), theoretical definitions and computation methods [fr

  9. A numerical solution model of the rewetting of a nuclear fuel rod

    International Nuclear Information System (INIS)

    Braz Filho, F.A.

    1984-01-01

    The study of thermal behaviour of a nuclear reactor fuel rod during the reflooding phase of the loss-of-coolant accident (LOCA) is presented. A mathematical model and a numerical scheme were proposed in order to solve the bidimensional heat conduction equation in cylindrical coordinates. The phenomenon of reflooding is not completely understood. One of the main difficulties is to estimate the heat transfer coefficient (h). For this reason two different models were elaborated: in the first three regions are considered and in each region h is considered constant; in the second the h profile is adjusted according to the boiling curve. The three region model yields satisfactory results at high and low mass flows while the 'boiling curve' model yields reasonable at low flows. (Author) [pt

  10. 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

  11. 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

  12. 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)

  13. 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.)

  14. Adsorption of asymmetric rigid rods or heteronuclear diatomic moleculeson homogeneous surfaces

    Science.gov (United States)

    Engl, W.; Courbin, L.; Panizza, P.

    2004-10-01

    We treat the adsorption on homogeneous surfaces of asymmetric rigid rods (like for instance heteronuclear diatomic molecules). We show that the n→0 vector spin formalism is well suited to describe such a problem. We establish an isomorphism between the coupling constants of the magnetic Hamiltonian and the adsorption parameters of the rigid rods. By solving this Hamiltonian within a mean-field approximation, we obtain analytical expressions for the densities of the different rod’s configurations, both isotherm and isobar adsorptions curves. The most probable configurations of the molecules (normal or parallel to the surface) which depends on temperature and energy parameters are summarized in a diagram. We derive that the variation of Qv , the heat of adsorption at constant volume, with the temperature is a direct signature of the adsorbed molecules configuration change. We show that this formalism can be generalized to more complicated problems such as for instance the adsorption of symmetric and asymmetric rigid rods mixtures in the presence or not of interactions.

  15. On the neutron noise diagnostics of pressurized water reactor control rod vibrations II. Stochastic vibrations

    International Nuclear Information System (INIS)

    Pazsit, I.; Glockler, O.

    1984-01-01

    In an earlier publication, using the theory of neutron fluctuations induced by a vibrating control rod, a complete formal solution of rod vibration diagnostics based on neutron noise measurements was given in terms of Fourier-transformed neutron detector time signals. The suggested procedure was checked in numerical simulation tests where only periodic vibrations could be considered. The procedure and its numerical testing are elaborated for stochastic two-dimensional vibrations. A simple stochastic theory of two-dimensional flow-induced vibrations is given; then the diagnostic method is formulated in the stochastic case, that is, in terms of neutron detector auto- and crosspower spectra. A previously suggested approximate rod localization technique is also formulated in the stochastic case. Applicability of the methods is then investigated in numerical simulation tests, using the proposed model of stochastic two-dimensional vibrations when generating neutron detector spectra that simulate measured data

  16. 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.…

  17. 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.)

  18. Heat transfer and friction on smooth and rough test rods

    International Nuclear Information System (INIS)

    Franken, W.M.P.; Hoogland, H.; Deijman, P.

    1977-06-01

    Results are reported on heat transfer and pressure drop tests on one smooth and nine rough test rods in an annular geometry. The wall roughness consisted of transversal ribs with various roughness pitches, rib heights and rib widths. The tests were performed with air as coolant under a wide range of experimental conditions: 10 5 5 , 1.1 2. Special attention has been given to the effect of variation of the physical coolant properties over the flow cross section. This effect could be described by the power function (Tsub(w)/Tsub(b))sup(-0.3l) in additional systematic variation of the heat transfer could be recognized, dependent on the coolant temperature level. The experimental results were correlated by the equation St = C(Tsub(in)) Resup(-0.2) Prsup(-0.6) (Tsub(w)/Tsub(b)sup(-0.31). Values of C(Tsub(in)) are given in tabular form. The thermal entrance effect has been measured on various test rods. A substantial reduction of the heat transfer coefficient was almost constant along the rough test rods

  19. The physical and chemical degradation of PWR fuel rods in severe accident conditions

    International Nuclear Information System (INIS)

    Parsons, P.D.; Mowat, J.A.S.; Dewhurst, D.W.F.; Hughes, T.E.

    1983-01-01

    An experimental study of the interaction between Zircaloy-4 cladding and UO 2 in PWR fuel rods heated to high temperatures with a negligible differential pressure across the cladding wall is described. The fuel rods were of dimensions appropriate to the 17x17 PWR fuel sub-assembly and were heated in a non-oxidising environment (vacuum) up to approx. 1850 deg. C either isothermally or through heating ramps. Observations were made concerning the extent and nature of the reaction zone between Zircaloy-4 and UO 2 over the temperature range 1500-1850 deg. C for times ranging from 1 min to 125 min. The location, morphology and the chemical composition of the phases formed are described along with the kinetics of their formation. (author)

  20. Closed form analytic solutions describing glow discharge plasma

    International Nuclear Information System (INIS)

    Pai, S.T.; Guo, X.M.; Zhou, T.D.

    1996-01-01

    On the basis of an analytic model developed previously [S. T. Pai, J. Appl. Phys. 71, 5820 (1992)], an improved version of the model for the description of dc glow discharge plasma was successfully developed. A set of closed form solutions was obtained from the governing equations. The two-dimensional, analytic solutions are functional and completely satisfy the governing equations, the actual boundary conditions, and Maxwell equations. They can be readily used to carry out numerical calculations without the necessity of employing any assumed boundary conditions. Results obtained from the model reveal that as the discharge gap spacing or pressure increases the maximum value in the electron density distribution moves toward the cathode. At a sufficiently large value of gap spacing, the positive column phenomenon begins to appear in the discharge region. The model has the capability of treating the positive column and negative glow as a continuous system without the necessity of studying them separately. The model also predicts a sharp rise of the positive ion density near the cathode and field reversal in the anode region. Variation of the electrode radius produces little effect on the axial spatial distribution of physical quantities studied. copyright 1996 American Institute of Physics

  1. Control Rod Reactivity Measurements in the Aagesta Reactor with the Pulsed Neutron Method

    Energy Technology Data Exchange (ETDEWEB)

    Bjoereus, K

    1969-07-01

    An extensive series of control rod measurements was made in the Aagesta reactor during the low power experimental period following the first criticality. This report describes the part of these investigations made with the pulsed neutron method, comprising nearly 300 measurements. The main objective was the determination of control rod reactivity worths for different rods and groups of rods, but some supplementary measurements were also made, e.g. a determination of the prompt neutron decay constant for the delayed critical condition and four different cores. The cores consisted of 20, 32, 68, and 140 fuel elements respectively, and measurements were made at room temperature and with the moderator level close to critical for each core, and for the 140-element core also with full moderator height and at the temperatures 140 deg C and 215 deg C. Both fully and partly inserted control rod groups were investigated. The measurements at critical water level give directly the control rod reactivity worths, whereas those with full water height give the shut-down reactivity. A comparison was made between measured reactivity worths for a number of rod groups and those calculated with the HETERO code. The prompt neutron decay constant at delayed criticality {alpha}{sub 0}={beta}/l, for the full core at 215 deg C was found to be 9.60 {+-} 0.30/sec, corresponding to l = 0.76 {+-} 0.02 msec. The shut-down reactivity with 16 coarse control rods in pos. A-D 22, 40-04, 44, 26 is -5% at 25 deg C and -13% at 215 deg C. The relative error is usually around 8% in the reactivity worths, originating mainly from the higher harmonics content in the measured curves.

  2. Features of electric drive sucker rod pumps for oil production

    Science.gov (United States)

    Gizatullin, F. A.; Khakimyanov, M. I.; Khusainov, F. F.

    2018-01-01

    This article is about modes of operation of electric drives of downhole sucker rod pumps. Downhole oil production processes are very energy intensive. Oil fields contain many oil wells; many of them operate in inefficient modes with significant additional losses. Authors propose technical solutions to improve energy performance of a pump unit drives: counterweight balancing, reducing of electric motor power, replacing induction motors with permanent magnet motors, replacing balancer drives with chain drives, using of variable frequency drives.

  3. Numerical experiment designs. Study of vibratory behaviour of PWR'S control rod clusters

    International Nuclear Information System (INIS)

    Bosselut, D.; Soulier, B.; Regnier, G.

    1997-01-01

    The application of Experiment Design method to Finite Element Model (FEM) calculations is an original way of performing parametric studies. It has been used at EDF to simulate on a large parametric domain the vibrations of PWR's control rod cluster and to analyse the rod wear process. In the first part the FEM and the location of excitation sources are described. The calculated values are: rod displacement in the guiding cards, shock forces on the guiding cards and the wear power produced. In the second part, the computed Experiment Domain is described. This method approaches the response surface by a second degree polynomial. The retained model is composed for every parameters of all linear, quadratic and interaction terms (26 coefficients). In all, 34 polynomials have been built to approach the effective shock forces and the mean wear power at each of the 17 guiding points. In the third part the building of the computer Experiment Design is detailed: by Doehlert design adaptation to take into account a qualitative parameter, design optimization by adding four well chosen experiments and finally, design extension by passing from 4 to 6 parameters. In the last part, all the information deduced from application of this method are presented. The influence of parameters on calculated effective shock forces has been determined along rods and response surface have been easily approximated. The systematism and closeness of Experiment Design technique is underlined. Easy simulation of all the response domain by polynomial approach allows comparison with experimental results. (authors)

  4. The Verification of Coupled Neutronics Thermal-Hydraulics Code NODAL3 in the PWR Rod Ejection Benchmark

    Directory of Open Access Journals (Sweden)

    Surian Pinem

    2014-01-01

    Full Text Available A coupled neutronics thermal-hydraulics code NODAL3 has been developed based on the few-group neutron diffusion equation in 3-dimensional geometry for typical PWR static and transient analyses. The spatial variables are treated by using a polynomial nodal method while for the neutron dynamic solver the adiabatic and improved quasistatic methods are adopted. In this paper we report the benchmark calculation results of the code against the OECD/NEA CRP PWR rod ejection cases. The objective of this work is to determine the accuracy of NODAL3 code in analysing the reactivity initiated accident due to the control rod ejection. The NEACRP PWR rod ejection cases are chosen since many organizations participated in the NEA project using various methods as well as approximations, so that, in addition to the reference solutions, the calculation results of NODAL3 code can also be compared to other codes’ results. The transient parameters to be verified are time of power peak, power peak, final power, final average Doppler temperature, maximum fuel temperature, and final coolant temperature. The results of NODAL3 code agree well with the PHANTHER reference solutions in 1993 and 1997 (revised. Comparison with other validated codes, DYN3D/R and ANCK, shows also a satisfactory agreement.

  5. Application of COMSOL in the solution of the neutron diffusion equations for fast nuclear reactors in stationary state; Aplicacion de COMSOL en la solucion de las ecuaciones de difusion de neutrones para reactores nucleares rapidos en estado estacionario

    Energy Technology Data Exchange (ETDEWEB)

    Silva A, L.; Del Valle G, E., E-mail: evalle@ipn.mx [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN s/n, Col. San Pedro Zacatenco, 07738 Mexico D. F. (Mexico)

    2012-10-15

    This work shows an application of the program COMSOL Multi physics Ver. 4.2a in the solution of the neutron diffusion equations for several energy groups in nuclear reactors whose core is formed by assemblies of hexagonal transversal cut as is the cas of fast reactors. A reference problem of 4 energy groups is described of which takes the cross sections which are processed by means of a program that prepares the definition of the constants utilized in COMSOL for the generic partial differential equations that this uses. The considered solution domain is the sixth part of the core which is applied frontier conditions of reflection and incoming flux zero. The discretization mesh is elaborated in automatic way by COMSOL and the solution method is one of finite elements of Lagrange grade two. The reference problem is known as the Knk with and without control rod which led to propose the calculation of the effective multiplication factor in function of the control rod fraction from a value 0 (completely inserted control rod) until the value 1 (completely extracted control rod). Besides this the reactivity was determined as well as the change of this in function of control rod fraction. The neutrons scalar flux for each energy group with and without control rod is proportioned. The reported results show a behavior similar to the one reported in other works but using the discreet ordinates S{sub 2} approximation. (Author)

  6. 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)

  7. Post test investigation of the single rod tests ESSI 1-11 on temperature escalation in PWR fuel rod simulators due to the Zircaloy/steam reaction

    International Nuclear Information System (INIS)

    Hagen, S.; Kapulla, H.; Malauschek, H.; Katanishi, S.

    1987-03-01

    This KfK-report describes the posttest investigation of the single rod tests ESSI-1 to ESSI-11. The objective of these tests was to investigate the temperature escalation behaviour of Zircaloy clad PWR-fuel rods in steam. The investigation of the temperature escalation is part of the program of out-of-pile experiments (CORA) performed within the frame work of the PNS Severe Fuel Damage Program. The experimental arrangement consisted of fuel rod simulator (central tungsten heater, UO 2 ring pellets and Zircaloy cladding), Zircaloy shroud and fiber ceramic insulation. The introductory test ESSI-1 to ESSI-3 were scoping tests designed to obtain information on the temperature escalation of zircaloy in steam. ESSI-4 to ESSI-8 were run with increasing heating rates to investigate the influence of the oxide layer thickness at the start of the escalation. ESSI-9 to ESSI-11 were performed to investigate the influence of the insulation thickness on the escalation behaviour. In these tests we also learned that the gap between removed shroud and insulation has a remarkable influence due to heat removal by convection in the gap. After the test the fuel rod simulator was embedded into epoxy and cut by a diamond saw. The cross sections were photographed and investigated by metalograph microscope, SEM and EMP examinations. (orig./GL) [de

  8. Methods for reactor physics calculations for control rods in fast reactors

    International Nuclear Information System (INIS)

    Grimstone, M.J.; Rowlands, J.L.

    1988-12-01

    The IAEA Specialists' Meeting on ''Methods for Reactor Physics Calculations for Control Rods in Fast Reactors'' was held in Winfrith, United Kingdom, on 6-8 December, 1988. The meeting was attended by 23 participants from nine countries. The purpose of the meeting was to review the current calculational methods and their accuracy as assessed by theoretical studies and comparisons with measurements, and then to identify the requirements for improved methods or additional studies and comparisons. The control rod properties or effects to be considered were their reactivity worths, their effect on the power distribution through the core, and the reaction rates and energy deposition both within and adjacent to the rods. The meeting was divided into five sessions, in the first of which each national delegation presented a brief overview of their programme of work on calculational methods for fast reactor control rods. In the next three sessions a total of seventeen papers were presented describing calculational methods and assessments of their accuracy. The final session was a discussion to draw conclusions regarding the current status of methods and the further developments and validation work required. A separate abstract was prepared for each of the 23 papers presented at the meeting. Refs, figs and tabs

  9. 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.

  10. Eliminating the human element and the drudgery from control-rod calibrations

    Energy Technology Data Exchange (ETDEWEB)

    Ruby, L; Wang, H -K [Univ. of California, Berkeley (United States)

    1974-07-01

    , and a RODCALN provides both a table and a plot of the results. From an input of t{sub 50}'s obtained over the whole rod length, the program determines the reactivity for each increment of rod distance by a continuous function which approximates the table of Inhour solutions prepared by GA. The plots are of sufficient detail to be used in routine reactor operation for reactivity determinations. The results obtained with this program have been compared with Randall's contention that a consine-squared function is an excellent fit to the data for the differential worth of a control rod. The close agreement confirms Randall's hypothesis. (author)

  11. 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.)

  12. 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

  13. 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

  14. Pushing the limits of photoreception in twilight conditions: The rod-like cone retina of the deep-sea pearlsides

    KAUST Repository

    Busserolles, Fanny de

    2017-11-09

    Most vertebrates have a duplex retina comprising two photoreceptor types, rods for dim-light (scotopic) vision and cones for bright-light (photopic) and color vision. However, deep-sea fishes are only active in dim-light conditions; hence, most species have lost their cones in favor of a simplex retina composed exclusively of rods. Although the pearlsides, Maurolicus spp., have such a pure rod retina, their behavior is at odds with this simplex visual system. Contrary to other deep-sea fishes, pearlsides are mostly active during dusk and dawn close to the surface, where light levels are intermediate (twilight or mesopic) and require the use of both rod and cone photoreceptors. This study elucidates this paradox by demonstrating that the pearlside retina does not have rod photoreceptors only; instead, it is composed almost exclusively of transmuted cone photoreceptors. These transmuted cells combine the morphological characteristics of a rod photoreceptor with a cone opsin and a cone phototransduction cascade to form a unique photoreceptor type, a rod-like cone, specifically tuned to the light conditions of the pearlsides\\' habitat (blue-shifted light at mesopic intensities). Combining properties of both rods and cones into a single cell type, instead of using two photoreceptor types that do not function at their full potential under mesopic conditions, is likely to be the most efficient and economical solution to optimize visual performance. These results challenge the standing paradigm of the function and evolution of the vertebrate duplex retina and emphasize the need for a more comprehensive evaluation of visual systems in general.

  15. Mechanical reliability of falling systems. Case of control rods of nuclear reactors

    International Nuclear Information System (INIS)

    Rezkallah, B.

    1991-05-01

    After having briefly recalled the main principles adopted for nuclear reactor safety (barrier based method, emergency stop system, emergency cooling system) and how control rod operation anomalies are described, this research thesis first gives an overview of knowledge in the field of reliability (principles, qualitative and quantitative methods, failure trees, performance function, shock modelling and mode analysis). Then, the author presents the modelling of the drop and rotation of a control rod. The computation technique is the modal synthesis. The author highlights the influence of physical (friction, initial conditions) and numerical (modal truncation) parameters. He reports a probabilistic analysis of the problem and proposes a stochastic model for the different shock forces (notably friction). He compares the results with those obtained with the DEMT software. The last part reports the investigation of a rare case: the jamming or untimely stop of the control rod device

  16. Activity determination of the Am-241 sources from radioactive lightning rods

    International Nuclear Information System (INIS)

    Minematsu, Denise; Dellamano, Jose Claudio; Ferreira, Robson de Jesus

    2009-01-01

    The authorization for manufacture commerce and installation of radioactive lightning rods, in Brazil, was lifted in 1989 by the National Nuclear Energy Commission - CNEN (Resolution no 4/89). Since this date, these devices have been replaced and have been sent to the Institutes subordinated to the CNEN, amongst them the Nuclear and Energy Research Institute - IPEN-CNEN/SP. Radioactive Waste Management Laboratory - RWML of the IPEN - CNEN/SP had received, approximately, 16,000 units up to the end of 2008. The radioactive lightning rod is constituted in its majority, for a central metallic rod, where two or three metallic plates are mounted. In these plates, on average, six Am-241 sources are fixed. The process used for the radioactive lightning rods treatment is the dismantling of the device and the withdrawal of the sources from the metallic plates. The activity values of the lightning rods sources, supplied by the manufacturers, vary from two to three orders of magnitude and therefore it is necessary to characterize these sources. This paper describes the methodology used to measure the actual activity of each Am-241 sources extracted from the radioactive lightning rods. The first step was to sample tens of Am-241 sources and carry out the activity measurements for further use in the system calibration. The equipment used in this first stage was a gamma spectrometer, previously calibrated with an Am-241 standard source, in agreement with the same arrangement and same geometry in the measures of the sources. Results show that there are sources with similar activity values of those supplied by the manufacturers, but there are also sources with no activity - or also activity very low compared with the expected value -, as well as sources contend other radionuclides. (author)

  17. Numerical experiment designs: study of the vibrational behaviour of the control rod cluster of a pressurized water reactor

    International Nuclear Information System (INIS)

    Soulier, B.; Bosselut, D.; Regnier, G.

    1997-01-01

    A finite element model has been performed at EDF to simulate the vibrations of control rod cluster assembly and to analyse the wear phenomenon of control rods. A parametrical study bas been performed for a given computer experiment domain with an experimental design method. The building of the computer experiment design is described. The influence of parameters on calculated mean wear power has been determined along rods and responses surfaces have been easily approximated. Systematism and closeness of experiment design technique is underlined. (authors)

  18. 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...

  19. 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.)

  20. 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.).

  1. Analytical Solutions To Describe Juxtaposed Sands | Adeniji ...

    African Journals Online (AJOL)

    Mathematical (linear diffusion) equations are presented for two pseudoreservoir regions intersected by fault that describe the effects of partial communicating fault on pressure transient behaviour for each fault block. Green's and source function technique solve these equations. A two-well system is considered for the ...

  2. 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....

  3. FREC-4A: a computer program to predict fuel rod performance under normal reactor operation

    International Nuclear Information System (INIS)

    Harayama, Yasuo; Izumi, Fumio

    1981-10-01

    The program FREC-4A (Fuel Reliability Evaluation Code-version 4A) is used for predicting fuel rod performance in normal reactor operation. The performance is calculated in accordance with the irradiation history of fuel rods. Emphasis is placed on the prediction of the axial elongation of claddings induced by pellet-cladding mechanical interaction, including the influence of initially preloaded springs inserted in fuel rod lower plenums. In the FREC-4A, an fuel rod is divided into axial segments. In each segment, it is assumed that the temperature, stress and strain are axi-symmetrical, and the axial strain in constant in fuel pellets and in a cladding, though the values in the pellets and in the cladding are different. The calculation of the contact load and the clearance along the length of a fuel rod and the stress and strain in each segment is explained. The method adopted in the FREC-4A is simple, and suitable to predict the deformation of fuel rods over their full length. This report is described on the outline of the program, the method of solving the stiffness equations, the calculation models, the input data such as irradiation history, output distribution, material properties and pores, the printing-out of input data and calculated results. (Kako, I.)

  4. 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.

  5. Lightning rod ionizing natural ionca - Ionic electrode active trimetallictriac of grounding - Definitive and total solution against 'blackouts' and electrical faults generated by atmospheric charges (lightning)

    Energy Technology Data Exchange (ETDEWEB)

    Cabareda, Luis

    2010-09-15

    The Natural Ionizing System of Electrical Protection conformed by: Lightning Rod Ionizing Natural Ionca and Ionic Electrode Active Trimetallic Triac of Grounding offers Total Protection, Maximum Security and Zero Risk to Clinics, Hospitals, Integral Diagnostic Center, avoiding ''the burning'' of Electronics Cards; Refineries, Tanks and Stations of Fuel Provision; Electrical Substations, Towers and Transmission Lines with transformer protection, motors, elevators, A/C, mechanicals stairs, portable and cooling equipment, electrical plants, others. This New High Technology is the solution to the paradigm of Benjamin Franklin and it's the mechanism to end the 'Blackouts' that produces so many damages and losses throughout the world.

  6. 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.)

  7. Comparison of the worth of control and protection system rods of different design on the basis of the measurements in BN-600 reactor

    International Nuclear Information System (INIS)

    Vasilyev, B.A.; Roslyakov, V.F.; Farakshin, M.R.

    1988-01-01

    The results of the worth measurements of the basic and experimental absorbing rods of BN-600 reactor are presented. The procedure used for the rods worth comparison on the basis of calculated and experimental data interpretation is described here. Basic and experimental rods relative worth is also presented. (author). 5 refs, 3 figs, 2 tabs

  8. 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.

  9. Interactions between halloysite nanotubes and poly(styrene sulfonate) in solution

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jae Heon; Ryu, Jung Ju; Shin, Joo Huei; Lee, Hoik; Sohn, Dae Won [Dept. of Chemistry and Research Institute for Convergence of Basic Science, Hanyang University, Seoul (Korea, Republic of); Kim, Ick Soo [Nano Fusion Technology Research Lab, Division of Frontier Fibers, Institute for Fiber Engineering (IFES), Interdisciplinary Cluster for Cutting Edge Research (ICCER), Shinshu University, Nagano (Japan)

    2017-01-15

    The interaction between halloysite nanotubes (HNT) and poly(styrene sulfonate) (PSS) in aqueous solution was investigated by dynamic light scattering. Dynamic behavior of HNT/PSS was observed with different salt, HNT, and PSS concentrations. The HNT colloids were stabilized by PSS over a wide range of HNT concentrations, and HNT suspension in dilute solution formed stable HNT/PSS particles. On the other hand, HNT particles aggregated as sediments at higher concentrations due to strong attraction among HNT rods, and HNT aggregates were stabilized by additional PSS. The interactions between HNT and PSS are described by the van der Waals–London force (VDWL). The stabilization process of HNT/PSS particles in salt solution was proposed by comparing the hydrodynamic radii and apparent intensities of samples. The results demonstrate that electrostatic, steric, and depletion stabilization processes are responsible for the stable dispersion of HNT even at high concentration.

  10. 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)

  11. A Secondary Flow Effect on the Heat and Mass Transfer Processes in the Finned Rod Bundles of Gas-cooled Reactors

    Directory of Open Access Journals (Sweden)

    A. A. Dunaitsev

    2017-01-01

    Full Text Available In nuclear power engineering a need to justify an operability of products and their components is of great importance. In high-temperature gas reactors, the critical element affecting the facility reliability is the fuel rod cladding, which in turn leads to the need to gain knowledge in the field of gas dynamics and heat transfer in the reactor core and to increase the detail of the calculation results. For the time being, calculations of reactor core are performed using the proven techniques of per-channel calculations, which show good representativeness and count rate. However, these techniques require additional experimental studies to describe correctly the inter-channel exchange, which, being taken into account, largely affects the pattern of the temperature fields in the region under consideration. Increasingly more relevant and demandable are numerical simulation methods of fluid and gas dynamics, as well as of heat exchange, which consist in the direct solution of the system of differential equations of mass balance, kinetic moment, and energy. Calculation of reactor cores or rod bundles according these techniques does not require additional experimental studies and allows us to obtain the local distributions of flow characteristics in the bundle and the flow characteristics that are hard to measure in the physical experiment.The article shows the calculation results and their analysis for an infinite rod lattice of the reactor core. The results were obtained by the technique of modelling one rod of a regular lattice using the periodic boundary conditions, followed by translating the results to the neighbouring rods. In channels of complex shape, there are secondary flows caused by changes in the channel geometry along the flow and directed across the main front of the flow. These secondary flows in the reactor cores with rods spaced by the winding wire lead to a redistribution of the coolant along the channel section, which in turn

  12. Status and results of the theoretical and experimental investigations on the LWR fuel rod behavior under accident conditions

    International Nuclear Information System (INIS)

    Bocek, M.; Hofmann, P.; Leistikow, S.; Class, G.; Meyder, R.; Raff, S.; Erbacher, F.; Hofmann, G.; Ihle, P.; Karb, E.; Fiege, A.

    1978-09-01

    In this report the status of knowledge is described which has been gathered up to the end of 1977 of the LWR fuel rod behavior in loss-of-coolant accidents. The majority of results indicated have been derived from studies on the fuel rod behavior performed within the framework of the Nuclear Safety Project (PNS); partly, also the results of cooperating research establishments and fm international exchange of experience are referred to. The report has been subdivided into two complete parts: Part I provides a survey of the most significant results of the theoretical and experimental research projects on fuel rod behavior. Part II describes by detailed individual presentations the status as well as the results with respect to the major central subjects. (orig.) 891 RW 892 AP [de

  13. Modelling of Rod No 8 in IFA-597:3

    International Nuclear Information System (INIS)

    Malen, K.

    2002-06-01

    A Westinghouse Atom 8x8 fuel rod irradiated in the Ringhals 1 BWR for 12 years to a local burnup of about 67 MWd/kgU was refabricated, instrumented with centreline thermocouple and pressure transducer, and irradiated in IFA-597.2 for about 20 days and in IFA-597.3 for about four months. The rod was then sent to Kjeller for puncturing and then to the Studsvik hot cells for detailed post-irradiation examinations. The peak centreline, temperature was close to 1350 deg C. The total fission gas release (FGR) determined from the puncturing was approximately 20 %. Electron probe microanalysis on a fuel section from the central part of the rod showed that virtually 100 % Xe release had occurred in the central part of the pellet out to about half the pellet radius, and this thermal release from the central part of the fuel accounted for the measured total FGR. Optical and scanning electron microscopy of the fuel cross-section showed complete pellet-clad bonding as well as an extensive high burnup 'rim' structure extending at least 0,15 mm in from the fuel surface. The fuel microstructure was characterised at different radial positions in the pellet. This report describes modelling of the rod behaviour using the code SKIROD, in particular fuel temperature and fission gas release. The transient response of the fuel centre line temperature after a scram is also modelled using the code TOODEE2. The modelling results are compared to the experimental results

  14. 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

  15. 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

  16. 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

  17. 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.

  18. 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)

  19. 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.)

  20. 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.

  1. The treatment of absorber rod heterogeneity effects using homogeneous equivalent cross-sections and their application in large fast reactors

    International Nuclear Information System (INIS)

    Newton, T.D.

    1988-01-01

    This paper examines the application of homogeneous equivalent absorber rod cross-sections to the calculation of control rod anti-reactivities in large fast reactors. The method used to obtain the equivalent cross-sections is described and their validity in simple whole core geometry calculations is verified. Finally, they are employed in the calculation of control rod anti-reactivity worths in the Super Phenix 1 fast reactor and the results are compared with measured values. (author). 5 refs, 5 figs, 9 tabs

  2. 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.

  3. 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

  4. 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.

  5. 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.

  6. Seeded growth of InP and InAs quantum rods using indium acetate and myristic acid

    International Nuclear Information System (INIS)

    Shweky, Itzhak; Aharoni, Assaf; Mokari, Taleb; Rothenberg, Eli; Nadler, Moshe; Popov, Inna; Banin, Uri

    2006-01-01

    A synthesis of soluble III-V semiconductor quantum rods using gold nanoparticles to direct and catalyze one-dimensional growth is developed. The growth takes place via the solution-liquid-solid (SLS) mechanism where proper precursors are injected into a coordinating solvent. We report the synthesis of InP nanorods using indium acetate and myristic acid with gold nanoparticles as the catalysts in the SLS growth mode. A similar route was successfully developed for the growth of InAs nanorods. We find that the amount of Au catalyst in the reaction is an important parameter to achieve shape control. Transmission electron microscope (TEM) images of InP and InAs nanocrystals revealed that the crystals are mostly rod-shaped, and provide strong evidence for Au presence in one edge. The rods were characterized structurally using X-ray diffraction and high-resolution TEM and optically by absorption and photoluminescence

  7. 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

  8. 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)

  9. 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

  10. LWR fuel rod testing facilities in high flux reactor (HFT) Petten for investigation of power cycling and ramping behaviour

    Energy Technology Data Exchange (ETDEWEB)

    Markgraf, J; Perry, D; Oudaert, J [Commission of the European Communities, Joint Reserach Centre, Petten Establishment, Petten (Netherlands)

    1983-06-01

    LWR fuel rod irradiation experiments are being performed in HFR's Pool Side Facility (PSF) by means of pressurized boiling water capsules (BWFC). For more than 6 years the major application of these devices has been in performing irradiation programs to investigate the power ramp behaviour of PWR and BWR rods which have been pre-irradiated in power reactors. Irradiation devices with various types of monitoring instrumentation are available, e.g. for fuel rod length, fuel stack length, fuel rod internal pressure and fuel rod central temperature measurements. The application scope covers PWR and BWR fuel rods, with burn-up levels up to 45 MWd/kg(U), max. linear heat generation rates up to 700 W/cm and simulation of constant power change rates between 0.05 and 1000 W/cm.min. The paper describes the different designs of LWR fuel rod testing facilities and associated non-destructive testing devices in use at the HFR Petten. It also addresses the new test capabilities that will become available after exchange of the HFR vessel in 1983. Furthermore it shows some typical results. (author)

  11. Geometric Nonlinear Computation of Thin Rods and Shells

    Science.gov (United States)

    Grinspun, Eitan

    2011-03-01

    We develop simple, fast numerical codes for the dynamics of thin elastic rods and shells, by exploiting the connection between physics, geometry, and computation. By building a discrete mechanical picture from the ground up, mimicking the axioms, structures, and symmetries of the smooth setting, we produce numerical codes that not only are consistent in a classical sense, but also reproduce qualitative, characteristic behavior of a physical system----such as exact preservation of conservation laws----even for very coarse discretizations. As two recent examples, we present discrete computational models of elastic rods and shells, with straightforward extensions to the viscous setting. Even at coarse discretizations, the resulting simulations capture characteristic geometric instabilities. The numerical codes we describe are used in experimental mechanics, cinema, and consumer software products. This is joint work with Miklós Bergou, Basile Audoly, Max Wardetzky, and Etienne Vouga. This research is supported in part by the Sloan Foundation, the NSF, Adobe, Autodesk, Intel, the Walt Disney Company, and Weta Digital.

  12. 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)

  13. 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)

  14. 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

  15. 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.)

  16. Experiments on the quench behavior of fuel rods

    International Nuclear Information System (INIS)

    Hofmann, P.; Noack, V.; Burbach, J.; Metzger, H.

    1995-01-01

    Because of the importance of the observed reflood phenomena for safety of current and future LWRs, the Forschungszentrum Karlsruhe (FZKA) started a program to investigate the mechanisms of quench-induced oxidation of Zircaloy. A small scale test-rig was designed and built in which it is possible to quench single Zircaloy rods by water and steam. The report describes the status of this work in May 1995. Some experimental results are presented. (orig./HP)

  17. Experiments on the quench behavior of fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Hofmann, P.; Noack, V.; Burbach, J.; Metzger, H.

    1995-08-01

    Because of the importance of the observed reflood phenomena for safety of current and future LWRs, the Forschungszentrum Karlsruhe (FZKA) started a program to investigate the mechanisms of quench-induced oxidation of Zircaloy. A small scale test-rig was designed and built in which it is possible to quench single Zircaloy rods by water and steam. The report describes the status of this work in May 1995. Some experimental results are presented. (orig./HP)

  18. The fractional and nonlinear magneto-flexible rod

    International Nuclear Information System (INIS)

    David, S A; Balthazar, J M; Julio, B H S; Oliveira, C

    2012-01-01

    This paper deals with a system involving a flexible rod subjected to magnetic forces that can bend it while simultaneously subjected to external excitations produces complex and nonlinear dynamic behavior, which may present different types of solutions for its different movement-related responses. This fact motivated us to analyze such a mechanical system based on modeling and numerical simulation involving both, integer order calculus (IOC) and fractional order calculus (FOC) approaches. The time responses, pseudo phase portraits and Fourier spectra have been presented. The results obtained can be used as a source for conduct experiments in order to obtain more realistic and more accurate results about fractional-order models when compared to the integer-order models.

  19. 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)

  20. 3-D modeling and motion simulation of fuel rod-replacing equipment

    International Nuclear Information System (INIS)

    Ding Jie; Zhu Libing

    2010-01-01

    In this paper, the process of 3-D modeling and motion simulation of fuel rod-replacing equipment using SolidWorks is described, and the application of SolidWorks in manufacturing and design improvement is discussed. Complexity of the manufacturing is reduced and reliability of the design is improved. (authors)

  1. 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

  2. A survey of blockage measurement methods used in PWR multi-rod experiments

    Energy Technology Data Exchange (ETDEWEB)

    Hindle, E.D.; Jones, C.; Whitty, S. (AEA Reactor Services, Springfield (UK))

    1986-05-01

    The deformation characteristics of Zircaloy multi-rod arrays are being investigated in laboratory and in-reactor tests, and heat transfer experiments are being carried out on pre-deformed arrays. The primary objective is to demonstrate that cladding distension occurring under hypothetical loss-of-coolant accident (LOCA) conditions will not impede the PWR emergency coolant flow during the reflood stage to the extent that unacceptably high cladding temperatures are reached, i.e. that a coolable geometry is maintained. This Report critically reviews the current methods for measuring blockage in multi-rod arrays and discusses their application. A new definition which overcomes the deficiencies of the previous methods is proposed even though it still has drawbacks in the case of overall blockage measurement. A method for automatically measuring the individual rod strain, general cluster blockage sub-channel blockage and sub-channel perimeter changes is described and the results from a deformed array presented. (author).

  3. A survey of blockage measurement methods used in PWR multi-rod experiments

    International Nuclear Information System (INIS)

    Hindle, E.D.; Jones, C.; Whitty, S.

    1986-05-01

    The deformation characteristics of Zircaloy multi-rod arrays are being investigated in laboratory and in-reactor tests, and heat transfer experiments are being carried out on pre-deformed arrays. The primary objective is to demonstrate that cladding distension occurring under hypothetical loss-of-coolant accident (LOCA) conditions will not impede the PWR emergency coolant flow during the reflood stage to the extent that unacceptably high cladding temperatures are reached, i.e. that a coolable geometry is maintained. This Report critically reviews the current methods for measuring blockage in multi-rod arrays and discusses their application. A new definition which overcomes the deficiencies of the previous methods is proposed even though it still has drawbacks in the case of overall blockage measurement. A method for automatically measuring the individual rod strain, general cluster blockage sub-channel blockage and sub-channel perimeter changes is described and the results from a deformed array presented. (author)

  4. National supply of reactivity control rods for Embalse nuclear power plant (CNE)

    International Nuclear Information System (INIS)

    Biondo, C.D.; Carloni, J.G.; Aba, J.A.

    1987-01-01

    The manufacture and supply on industrial scale of reactivity control rods for CNE (Embalse nuclear power plant) were developed by the National Atomic Energy Commission (CNEA) together with the private industry, as part of a program aimed to the substitution of imported supplies used in the operation of power plants by materials manufactured in Argentina. So far, the control rods were imported from Canada. In this work, the different development stages performed by CNEA and CONUAR S.A. are described, leading to the supply of a set of 21 cobalt rods to be included in a reactor of CNE in order to qualify this component. Among the main activities performed, the following stand out: specifications development, particularly those concerning to cobalt cores, evaluation of design documentation and elaboration of bidding conditions and a plan of manufacture and control. According to the results obtained during the service and the post-irradiation measurements, the design will be reviewed in order to undertake new manufacturing plans. (Author)

  5. 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.)

  6. 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

  7. Experimental benchmark data for PWR rod bundle with spacer-grids

    International Nuclear Information System (INIS)

    Dominguez-Ontiveros, Elvis E.; Hassan, Yassin A.; Conner, Michael E.; Karoutas, Zeses

    2012-01-01

    In numerical simulations of fuel rod bundle flow fields, the unsteady Navier–Stokes equations have to be solved in order to determine the time (phase) dependent characteristics of the flow. In order to validate the simulations results, detailed comparison with experimental data must be done. Experiments investigating complex flows in rod bundles with spacer grids that have mixing devices (such as flow mixing vanes) have mostly been performed using single-point measurements. In order to obtain more details and insight on the discrepancies between experimental and numerical data as well as to obtain a global understanding of the causes of these discrepancies, comparisons of the distributions of complete phase-averaged velocity and turbulence fields for various locations near spacer-grids should be performed. The experimental technique Particle Image Velocimetry (PIV) is capable of providing such benchmark data. This paper describes an experimental database obtained using two-dimensional Time Resolved Particle Image Velocimetry (TR-PIV) measurements within a 5 × 5 PWR rod bundle with spacer-grids that have flow mixing vanes. One of the unique characteristic of this set-up is the use of the Matched Index of Refraction technique employed in this investigation to allow complete optical access to the rod bundle. This unique feature allows flow visualization and measurement within the bundle without rod obstruction. This approach also allows the use of high temporal and spatial non-intrusive dynamic measurement techniques namely TR-PIV to investigate the flow evolution below and immediately above the spacer. The experimental data presented in this paper includes explanation of the various cases tested such as test rig dimensions, measurement zones, the test equipment and the boundary conditions in order to provide appropriate data for comparison with Computational Fluid Dynamics (CFD) simulations. Turbulence parameters of the obtained data are presented in order to gain

  8. 2D model for melt progression through rods and debris

    International Nuclear Information System (INIS)

    Fichot, F.

    2001-01-01

    During the degradation of a nuclear core in a severe accident scenario, the high temperatures reached lead to the melting of materials. The formation of liquid mixtures at various elevations is followed by the flow of molten materials through the core. Liquid mixture may flow under several configurations: axial relocation along the rods, horizontal motion over a plane surface such as the core support plate or a blockage of material, 2D relocation through a debris bed, etc.. The two-dimensional relocation of molten material through a porous debris bed, implemented for the simulation of late degradation phases, has opened a new way to the elaboration of the relocation model for the flow of liquid mixture along the rods. It is based on a volume averaging method, where wall friction and capillary effects are taken into account by introducing effective coefficients to characterize the solid matrix (rods, grids, debris, etc.). A local description of the liquid flow is necessary to derive the effective coefficients. Heat transfers are modelled in a similar way. The derivation of the conservation equations for the liquid mixture falling flow (momentum) in two directions (axial and radial-horizontal) and for the heat exchanges (energy) are the main points of this new model for simulating melt progression. In this presentation, the full model for the relocation and solidification of liquid materials through a rod bundle or a debris bed is described. It is implemented in the ICARE/CATHARE code, developed by IPSN in Cadarache. The main improvements and advantages of the new model are: A single formulation for liquid mixture relocation, in 2D, either through a rod bundle or a porous debris bed, Extensions to complex structures (grids, by-pass, etc..), The modeling of relocation of a liquid mixture over plane surfaces. (author)

  9. Dynamic surface-pressure instrumentation for rods in parallel flow

    International Nuclear Information System (INIS)

    Mulcahy, T.M.; Lawrence, W.

    1979-01-01

    Methods employed and experience gained in measuring random fluid boundary layer pressures on the surface of a small diameter cylindrical rod subject to dense, nonhomogeneous, turbulent, parallel flow in a relatively noise-contaminated flow loop are described. Emphasis is placed on identification of instrumentation problems; description of transducer construction, mounting, and waterproofing; and the pretest calibration required to achieve instrumentation capable of reliable data acquisition

  10. Pellet clad interaction analysis of AFA 3G fuel rod

    International Nuclear Information System (INIS)

    Liu Tong; Shen Caifen; Jiao Yongjun; Lu Huaquan; Zhou Zhou

    2002-01-01

    The author described Pellet Clad Interaction (PCI) analysis of AFA 3G fuel rod during condition II transients for GNPS 18-months alternating equilibrium cycles. It provided PCI technical limit, analytical methods and computer code used in the analyses of condition II transients and thermal-mechanical. Finally, given main calculation results and the conclusion for GNPS 18-months cycles

  11. 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.

  12. Rodding Surgery

    Science.gov (United States)

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

  13. Thermal performance of annular-coated and sphere-pac LWR fuel rod designs

    International Nuclear Information System (INIS)

    Guenther, R.J.; Hsieh, K.A.; Barner, J.O.; Freshley, M.D.

    1980-01-01

    Two FCI-resistant UO 2 fuel rod designs are being compared to a reference design in irradiation tests in the Halden Boiling Water Reactor (HBWR) as part of the DOE-sponsored Fuel Performance Improvement Program (FPIP). The primary fuel design (annular-coated-pressurized) incorporates annular pellets, a graphite coating on the inner surface of the Zircaloy cladding, and pressurized helium fill gas. Also being investigated is an 87% smear density sphere-pac design with pressurized helium fill gas. The solid pellet (reference) and annular-coated designs described had helium fill gas at approx. 100 kPa and the sphere-pac rods were pressurized at approx. 455 kPa

  14. Thermal phenomenae in nuclear fuel rods

    International Nuclear Information System (INIS)

    Baigorria, Carlos.

    1983-12-01

    Thermal phenomenae occurring in a nuclear fuel rod under irradiation are studied. The most important parameters of either steady or transient thermal states are determined. The validity of applying the Fourier's approximation equations to these problems is also studied. A computer program TRANS is developed in order to study the transient cases. This program solves a system of coupled, non-linear partial differential equations, of parabolic type, in cylindrical coordinates with various boundary conditions. The benchmarking of the TRANS program is done by comparing its predictions with the analytical solution of some simplified transient cases. Complex transient cases such as those corresponding to characteristic reactor accidents are studied, in particular for typical pressurized heavy water reactor (PHWR) fuel rods, such as those of Atucha I. The Stefan problem emerging in the case of melting of the fuel element is solved. Qualitative differences between the classical Stefan problem, without inner sources, and that one, which includes sources are discussed. The MSA program, for solving the Stefan problem with inner sources is presented; and furthermore, it serves to predict thermal evolution, when the fuel element melts. Finally a model for fuel phase change under irradiation is developed. The model is based on the dimensional invariants of the percolation theory when applied to the connectivity of liquid spires nucleated around each fission fragment track. Suggestions for future research into the subject are also presented. (autor) [es

  15. The pellet-cladding contact in a fuel rod and its simulation by finite elements

    International Nuclear Information System (INIS)

    Tanajura, C.A.S.

    1988-01-01

    A model to analyse the mechanical behavior of a fuel rod of a PWR is presented. We drew our attention to the phenomenon of pellet-pellet and pellet-cladding contact by taking advantage of a model which assumes the hypotheses of axisymmetry, elastic behavior with infinitesimal deformations and changes of the material properties with temperature. It also includes the effects of swelling and initial relocation. The problem of contact gives rise to a variational formulation which employs Lagrangian multipliers. With this approach an iterative scheme is constructed to obtain the solution. The finite element method is applied to space discretization. The model sensibility to some parameters and its performance concerning fuel rod behavior is discussed by means of numerical simulations. (author) [pt

  16. 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.

  17. 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.)

  18. 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.

  19. 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.)

  20. 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