WorldWideScience

Sample records for resistance formed-ferrite rod

  1. Flow resistance in rod assemblies

    International Nuclear Information System (INIS)

    Korsun, A.S.; Sokolova, M.S.

    2000-01-01

    The general form of relation between the resistance force and the velocity vector, resistance tensor structure and possible types of anisotropy in the flow thorough such structures as rod or tube assemblies are under discussion. Some questions of experimental determination of volumetric resistance force tensor are also under consideration. (author)

  2. Antimicrobial Resistance in Gram-Negative Rods Causing Bacteremia in Hematopoietic Stem Cell Transplant Recipients

    DEFF Research Database (Denmark)

    Averbuch, Diana; Tridello, Gloria; Hoek, Jennifer

    2017-01-01

    for resistance to fluoroquinolones, noncarbapenem anti-Pseudomonas β-lactams (noncarbapenems), carbapenems, and multidrug resistance. Results: Sixty-five HSCT centers from 25 countries in Europe, Australia, and Asia reported data on 655 GNR episodes and 704 pathogens in 591 patients (Enterobacteriaceae, 73......- and β-lactam/β-lactamase inhibitor resistance rates in allo-HSCT adults. Non-Klebsiella Enterobacteriaceae were rarely carbapenem resistant. Multivariable analysis revealed resistance risk factors in allo-HSCT patients: fluoroquinolone resistance: adult, prolonged neutropenia, breakthrough......%; nonfermentative rods, 24%; and 3% others). Half of GNRs were fluoroquinolone and noncarbapenem resistant; 18.5% carbapenem resistant; 35.2% multidrug resistant. The total resistance rates were higher in allogeneic HSCT (allo-HSCT) vs autologous HSCT (auto-HSCT) patients (P

  3. Specification for corrosion-resisting chromium and chromium-nickel steel welding rods and bare electrodes - approved 1969

    International Nuclear Information System (INIS)

    Anon.

    1975-01-01

    This specification covers corrosion-resisting chromium and chromium-nickel steel welding rods for use with the atomic hydrogen and gas-tungsten-arc welding processes and bare electrodes for use with the submerged arc and gas metal-arc welding processes. These welding rods and electrodes include those alloy steels designated as corrosion- or heat-resisting chromium and chromium-nickel steels, in which chromium exceeds 4% and nickel does not exceed 50%

  4. Control rod

    International Nuclear Information System (INIS)

    Takahashi, Akio.

    1982-01-01

    Purpose: To prevent distortion in control rod elements such as cladding tubes by decreasing the temperature difference between them. Constitution: In the case of housing a plurality of control rod elements in a protection pipe, flow rate control members are disposed in the protection pipe to equalize the flow resistance in each of coolant flow passages formed between the control rod elements and between the control rod elements and the inner surface of the protection pipe, to thereby unify the flow rate of the coolants flowing through these coolant flowing passages. Accordingly, each of the control rod elements can be cooled uniformly to thereby unify the temperature distribution and avoid the distortion in the cladding tubes, which may be resulted from bending due to the difference in thermal expansion and ununiform swelling due to the temperature difference. (Aizawa, K.)

  5. Experimental study on local resistance of two-phase flow through spacer grid with rod bundle

    International Nuclear Information System (INIS)

    Yan Chaoxing; Yan Changqi; Sun Licheng; Tian Qiwei

    2015-01-01

    The experimental study on local resistance of single-phase and two-phase flows through a spacer grid in a vertical channel with 3 × 3 rod bundle was carried out under the normal temperature and pressure. For the case of single-phase flow, the liquid Reynolds number covered the range of 290-18 007. For the case of two-phase flow, the ranges of gas and liquid superficial velocities were 0.013-3.763 m/s and 0.076-1.792 m/s, respectively. A correlation for predicting local resistance of single-phase flow was given based on experimental results. Eight classical two-phase viscosity formulae for homogeneous model were evaluated against the experimental data of two-phase flow. The results show that Dukler model predicts the experimental data well in the range of Re 1 < 9000 while McAdams correlation is the best one for Re 1 ≥ 9000. For all experimental data, Dukler model provides the best prediction with the mean relative error of 29.03%. A new correlation is fitted for the range of Re 1 < 9000 by considering mass quality, two- phase Reynolds number and liquid and gas densities, resulting in a good agreement with the experimental data. (authors)

  6. Upset Resistance Welding of Carbon Steel to Austenitic Stainless Steel Narrow Rods

    Science.gov (United States)

    Ozlati, Ashkaan; Movahedi, Mojtaba; Mohammadkamal, Helia

    2016-11-01

    Effects of welding current (at the range of 2-4 kA) on the microstructure and mechanical properties of upset resistance welds of AISI-1035 carbon steel to AISI-304L austenitic stainless steel rods were investigated. The results showed that the joint strength first increased by raising the welding current up to 3 kA and then decreased beyond it. Increasing trend was related to more plastic deformation, accelerated diffusion, reduction of defects and formation of mechanical locks at the joint interface. For currents more than 3 kA, decrease in the joint strength was mainly caused by formation of hot spots. Using the optimum welding current of 3 kA, tensile strength of the joint reached to 76% of the carbon steel base metal strength. Microstructural observations and microhardness results confirmed that there was no hard phase, i.e., martensite or bainite, at the weld zone. Moreover, a fully austenitic transition layer related to carbon diffusion from carbon steel was observed at the weld interface.

  7. Advanced KSNP fuel, plus7 : grid-to-rod fretting wear resistance of the plus7 spacer grids

    International Nuclear Information System (INIS)

    Kim, Kyu Tae; Kim, Yong Hwan; Jang, Young Ki; Choi, Joon Hyung

    2003-01-01

    Vibration-induced grid-to-rod fretting wear initiates at a certain critical gap correlated with a critical work rate. A critical gap between grid and rod forms due to in-reactor performance of fuel, thermal relaxation of grid spring and irradiation growth of grid strap, etc. A critical work rate may be generated by three vibration mechanisms proposed in this paper. Three vibration mechanisms have been derived with various fretting wear experience in commercial reactors as well as various out-of-pile hydraulic test results. The first active vibration mechanism is high turbulence-induced excessive fuel rod vibration with the combination of excessive grid-to-rod gap. The second active vibration mechanism is self-excited fuel assembly vibration in a low frequency range caused by hydraulically unbalanced mixing vanes of the spacer grid assembly. The third active vibration mechanism is self-excited spacer grid strap vibration in quite a high frequency range caused by some spacer grid designs. In this study, each vibration mechanism on the grid-to-rod fretting wear damage is discussed. On the other hand, the effects of various grid designs on the fretting wear damage in the commercial reactors are predicted using the long-term fretting wear test results. It is found that the larger grid-to-rod initial contact area generates the less fretting wear damage. Consequently the conformal spring of PLUS7 is superior to typical convex shaped spring with regard to fretting wear resistance since the former generates relatively larger contact area than the latter

  8. Specification for corrosion-resisting chromium and chromium-nickel steel bare and composite metal cored and stranded arc welding electrodes and welding rods

    International Nuclear Information System (INIS)

    Anon.

    1983-01-01

    This specification prescribes requirements for corrosion or heat resisting chromium and chromium-nickel steel electrodes and welding rods. These electrodes and welding rods are normally used for arc welding and include those alloy steels designated as corrosion or heat-resisting chromium and chromium-nickel steels, in which chromium exceeds 4.0 percent and nickel does not exceed 50.0 percent

  9. Study on Ground Resistance for Different Design Ground Rod(Asia-Pacific Symposium on Applied Electromagnetics and Mechanics(APSAEM10))

    OpenAIRE

    H., Farhan; S., Saifulnizam; A., Nazmin; A., Khamis; Faculty of Electrical Engineering, Universiti Teknikal Malaysia; Faculty of Electrical Engineering, Universiti Teknikal Malaysia; Faculty of Electrical Engineering, Universiti Teknikal Malaysia; Faculty of Electrical Engineering, Universiti Teknikal Malaysia

    2011-01-01

    Main purpose of grounding system are to carry dissipate electrical current into the earth and guarantee safety of people or equipment if any failure occur on electrical system. Lowest Ground resistance value is important elements to get the effective of the grounding system. As grounding standards evolved to require lower resistance levels, better design of electrodes are main contributions of this research. In this work, five different designs of ground rod which are solid rod or conventiona...

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

    International Nuclear Information System (INIS)

    Koscielniak, S.

    1994-06-01

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

  11. Development of Grounding Resistance Analysis Model of Rod Electrode Considering the Effect of Large-Current Characteristic for Distribution Lines

    Science.gov (United States)

    Asaoka, Yoshinobu; Motoyama, Hideki; Matsubara, Hiroji; Sugimoto, Hitoshi

    Grounding resistance is one of the important parameters in the lightning-protection design of electric power systems. The grounding resistance of electrode decreases as large currents are injected to the electrode by electric discharges in soil. This characteristic is not considered in lightning protection design. Therefore, the design level is kept on the exceeding level in actual phenomena. From viewpoint of the rational lightning protection design for the electric power systems, this characteristic should be considered in practical design. In this study, experiments were conducted using rod electrodes, and the physical phenomena of the electrical discharge in soil were considered by assuming a certain electrical discharge model in soil. Based on these results, a grounding resistance large-current characteristic analysis model that could be easily used in EMTP was developed.

  12. Morganella sp. rods – characteristics, infections, mechanisms of resistance to antibiotics 

    Directory of Open Access Journals (Sweden)

    Patrycja Zalas-Więcek

    2012-04-01

    Full Text Available The Morganella genus is one member of the tribe Proteae, which also includes the genera Proteus and Providencia. These bacteria are commonly present in the environment.Morganella sp. rods are known to be a causative agent of opportunistic hospital infections, mainly urinary tract, wound and blood infections of severe and high mortality, even in cases of an appropriate antibiotic.These bacteria may produce many virulence factors, for example urease, hemolysins, LPS, adhesins and enzymes hydrolyzing and modifying antibiotics commonly used to treat infections.Understanding the diverse biological properties of these rods may be of importance in the development of effective methods of prevention and control of infections with their participation. 

  13. Why Rods and Cocci

    Indian Academy of Sciences (India)

    experience greater frictional resistance. This hypothesis is supported by the fact that among the flagellated motile bacteria almost all are rod shaped. Only exceptionally few cocci are motile. This hypothesis, however, is not adequate since a large number of species of bacteria are non-motile. A rod shape can confer another ...

  14. 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.)

  15. Effect of mechanical pre-loadings on corrosion resistance of chromium-electroplated steel rods in marine environment

    Science.gov (United States)

    Shubina Helbert, Varvara; Dhondt, Matthieu; Homette, Remi; Arbab Chirani, Shabnam; Calloch, Sylvain

    2018-03-01

    Providing high hardness, low friction coefficient, as well as, relatively good corrosion resistance, chromium-plated coatings (∼20 μm) are widely used for steel cylinder rods in marine environment. However, the standardized corrosion test method (ISO 9227, NSS) used to evaluate efficiency of this type of coatings does not take into account in-service mechanical loadings on cylinder rods. Nevertheless, the uniform initial network of microcracks in chromium coating is changing under mechanical loadings. Propagation of these microcracks explains premature corrosion of the steel substrate. The aim of the study was to evaluate relationship between mechanical loadings, propagation of microcracks network and corrosion resistance of chromium coatings. After monotonic pre-loading tests, it was demonstrated by microscopic observations that the microcracks propagation started at stress levels higher than the substrate yield stress (520 MPa). The microcracks become effective, i.e. they have instantly undergone through the whole coating thickness to reach the steel substrate. The density of effective microcracks increases with the total macroscopic level, i.e. the intercrack distance goes from 60 ± 5 μm at 1% of total strain to approximately 27 ± 2 μm at 10%. Electrochemical measurements have shown that the higher the plastic strain level applied during mechanical loading, the more the corrosion potential of the sample decreased until reaching the steel substrate value of approximately ‑0.65 V/SCE after 2 h of immersion. The polarization curves have also highligthed an increase in the corrosion current density with the strain level. Therefore, electrochemical measurements could be used to realize quick and comprehensive assesment of the effect of monotonic pre-loadings on corrosion properties of the chromium coating.

  16. Activity of levofloxacin alone and in combination with a DnaK inhibitor against gram-negative rods, including levofloxacin-resistant strains.

    Science.gov (United States)

    Credito, Kim; Lin, Gengrong; Koeth, Laura; Sturgess, Michael A; Appelbaum, Peter C

    2009-02-01

    Synergy time-kill testing of levofloxacin alone and in combination with CHP-105, a representative DnaK inhibitor, against 50 gram-negative rods demonstrated that 34 of the 50 strains tested showed significant synergy between levofloxacin and CHP-105 after 12 h and 24 h. Fourteen of these 34 organisms were quinolone resistant (levofloxacin MICs of > or =4 microg/ml).

  17. Carbapenem disks on MacConkey agar in screening methods for detection of carbapenem-resistant Gram-negative rods in stools.

    Science.gov (United States)

    Blackburn, Julie; Tsimiklis, Catherine; Lavergne, Valéry; Pilotte, Josée; Grenier, Sophie; Gilbert, Andrée; Lefebvre, Brigitte; Domingo, Marc-Christian; Tremblay, Cécile; Bourgault, Anne-Marie

    2013-01-01

    Direct plating of simulated stool specimens on MacConkey agar (MCA) with 10-μg ertapenem, meropenem, and imipenem disks allowed the establishment of optimal zone diameters for the screening of carbapenem-resistant Gram-negative rods (CRGNR) of ≤ 24 mm (ertapenem), ≤ 34 mm (meropenem), and ≤ 32 mm (imipenem).

  18. Carbapenem Disks on MacConkey Agar in Screening Methods for Detection of Carbapenem-Resistant Gram-Negative Rods in Stools

    OpenAIRE

    Blackburn, Julie; Tsimiklis, Catherine; Lavergne, Valéry; Pilotte, Josée; Grenier, Sophie; Gilbert, Andrée; Lefebvre, Brigitte; Domingo, Marc-Christian; Tremblay, Cécile; Bourgault, Anne-Marie

    2013-01-01

    Direct plating of simulated stool specimens on MacConkey agar (MCA) with 10-μg ertapenem, meropenem, and imipenem disks allowed the establishment of optimal zone diameters for the screening of carbapenem-resistant Gram-negative rods (CRGNR) of ≤24 mm (ertapenem), ≤34 mm (meropenem), and ≤32 mm (imipenem).

  19. Rodding Surgery

    Science.gov (United States)

    ... usually undertaken as a scheduled elective procedure. An optimal age for a first rodding surgery has not ... which may prevent or postpone the need for replacement. The smallest diameter expanding rods are still too ...

  20. 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.)

  1. Activity of Levofloxacin Alone and in Combination with a DnaK Inhibitor against Gram-Negative Rods, Including Levofloxacin-Resistant Strains▿

    Science.gov (United States)

    Credito, Kim; Lin, Gengrong; Koeth, Laura; Sturgess, Michael A.; Appelbaum, Peter C.

    2009-01-01

    Synergy time-kill testing of levofloxacin alone and in combination with CHP-105, a representative DnaK inhibitor, against 50 gram-negative rods demonstrated that 34 of the 50 strains tested showed significant synergy between levofloxacin and CHP-105 after 12 h and 24 h. Fourteen of these 34 organisms were quinolone resistant (levofloxacin MICs of ≥4 μg/ml). PMID:19015359

  2. Development study of concrete reinforcement made of aramid fiber-reinforced plastic rods with high radiation resistance. 1. Epoxy resin compounds with a handling at room temperature impregnation

    International Nuclear Information System (INIS)

    Udagawa, Akira; Seguchi, Tadao; Moriya, Toshio; Matsubara, Sumiyuki; Hongou, Yoshihiko

    1999-03-01

    Aramid fiber-reinforced plastic (ArFRP) rods were developed in order to avoid from conduction current and/or magnetization of the metallic reinforcement using concrete constructions. For the polymer matrix, new epoxy resin compounds consist of tetraglycidyl diaminodiphenylmethane (30%), diglycidyl ether of bisphenol-A (60%), styrene oxide (10%) and aromatic diamine as a hardner were found to be the best formulation, and which were easily impregnated to the aramid fiber braiding yarn at room temperature. The ArFRP rods has a high radiation resistance, and the tensile strength was maintained to 98% (1.45 GPa) after irradiation dose of 100 MGy (absorbed energy MJ/kg), which is available for the reinforcement of concrete construction for the house of fusion reactor with super conducting magnets. (author)

  3. 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

  4. 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.)

  5. 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.)

  6. 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

  7. Nuclear reactor remote disconnect control rod coupling indicator

    International Nuclear Information System (INIS)

    Vuckovich, M.

    1977-01-01

    A coupling indicator for use with nuclear reactor control rod assemblies which have remotely disengageable couplings between the control rod and the control rod drive shaft is described. The coupling indicator indicates whether the control rod and the control rod drive shaft are engaged or disengaged. A resistive network, utilizing magnetic reed switches, senses the position of the control rod drive mechanism lead screw and the control rod position indicating tube, and the relative position of these two elements with respect to each other is compared to determine whether the coupling is engaged or disengaged

  8. Fuel rods

    International Nuclear Information System (INIS)

    Fukushima, Kimichika.

    1984-01-01

    Purpose: To reduce the size of the reactor core upper mechanisms and the reactor container, as well as decrease the nuclear power plant construction costs in reactors using liquid metals as the coolants. Constitution: Isotope capturing devices comprising a plurality of pipes are disposed to the gas plenum portion of a nuclear fuel rod main body at the most downstream end in the flowing direction of the coolants. Each of the capturing devices is made of nickel, nickel alloys, stainless steel applied with nickel plating on the surface, nickel alloys applied with nickel plating on the surface or the like. Thus, radioactive nuclides incorporated in the coolants are surely captured by the capturing devices disposed at the most downstream end of the nuclear fuel main body as the coolants flow along the nuclear fuel main body. Accordingly, since discharging of radioactive nuclides to the intermediate fuel exchange system can be prevented, the maintenance or reparing work for the system can be facilitated. (Moriyama, K.)

  9. Rod examination gauge

    Energy Technology Data Exchange (ETDEWEB)

    Bacvinskas, W.S.; Bayer, J.E.; Davis, W.W.; Fodor, G.; Kikta, T.J.; Matchett, R.L.; Nilsen, R.J.; Wilczynski, R.

    1991-12-31

    The present invention is directed to a semi-automatic rod examination gauge for performing a large number of exacting measurements on radioactive fuel rods. The rod examination gauge performs various measurements underwater with remote controlled machinery of high reliability. The rod examination gauge includes instruments and a closed circuit television camera for measuring fuel rod length, free hanging bow measurement, diameter measurement, oxide thickness measurement, cladding defect examination, rod ovality measurement, wear mark depth and volume measurement, as well as visual examination. A control system is provided including a programmable logic controller and a computer for providing a programmed sequence of operations for the rod examination and collection of data.

  10. 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.)

  11. FUEL ROD ASSEMBLY

    Science.gov (United States)

    Hutter, E.

    1959-09-01

    A cluster of nuclear fuel rods aod a tubular casing through which a coolant flows in heat-change contact with the ruel rods are described. The casting is of trefoil section and carries the fuel rods, each of which has two fin engaging the serrated fins of the other two fuel rods, whereby the fuel rods are held in the casing and are interlocked against relative longitudinal movement.

  12. 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

  13. Control rod blocking monitor

    International Nuclear Information System (INIS)

    Suzuki, Shigeru.

    1993-01-01

    The number of times for setting up a control rod blocking monitor of a BWR type power plant is remarkably reduced to mitigate operator's burden. In the control rod blocking monitor, trip levels, as a judging standard upon outputting control rod blocking inhibition signals, are set up stepwise depending on the power level around control rods put to blocking control. The present invention comprises an allowance judging means capable of setting up trip levels for each of power levels corresponding to a plurality of control rods at once if the power levels are within the set up allowable range. With such a constitution, the set up allowable range is determined previously in the allowance judging means. Accordingly, when a gang blocking is conducted to control rods, if power levels around the control rods are increased at once into the set up allowable range, the trip levels for each of the control rods are set up at once. (I.S.)

  14. 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.

  15. Tie rod insertion test

    CERN Multimedia

    B. LEVESY

    2002-01-01

    The superconducting coil is inserted in the outer vaccum tank and supported by a set of tie rods. These tie rods are made of titanium alloy. This test reproduce the final insertion of the tie rods inside the outer vacuum tank.

  16. 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)

  17. 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

  18. Compacting spent fuel rods

    International Nuclear Information System (INIS)

    Wachter, W.J.

    1988-01-01

    A method and apparatus for compacting spent fuel rods comprises transferring the rods from a nuclear fuel rod assembly into a different nuclear fuel rod container having a smaller cross section than the assembly. The individual rods are moved from a fuel assembly and through a transition funnel by movable grippers at opposite ends of the funnel. One movable gripper reciprocates between gripping and release positions in a gap between the fuel assembly and the transition funnel. All of the fuel rods are withdrawn concurrently and are merged towards one another into a tighter array within the transition funnel and emerge as a bundle. A movable and a stationary bundle gripper are provided between the funnel and the storage container to advance the bundle of fuel rods into the container. (author)

  19. Control rod drives

    International Nuclear Information System (INIS)

    Shimano, Kunio; Nakamura, Akira; Mizuguchi, Koji; Sakai, Kazuhito; Mitsui, Hisayasu.

    1994-01-01

    The present invention concerns upper-built-in type control rod drives of a BWR type reactor. Namely, high temperature linear motor driving type control rod drives are disposed in an upper space of the reactor pressure vessel, which generates electromagnetic power. In usual driving of control rods, driving shafts connected with control rods by a high temperature linear motor driving system comprising a driving shaft having an iron core inserted therein and electromagnetic coils is vertically moved to insert/withdraw the control rods to and from the reactor core. Upon occurrence of reactor scram, electric power source is interrupted, and the control rods are rapidly inserted to the reactor core. According to the present invention, since the control rod drives are disposed in the space above the reactor pressure vessel, pipelines or equipments passing through the bottom of the reactor pressure vessel can be saved. As a result, operation for maintenance and inspection is facilitated. (I.S.)

  20. Development of electrically heated rods with resistive element of graphite or carbon/carbon composites for simulating transients in nuclear reactors

    International Nuclear Information System (INIS)

    Polidoro, H.A.

    1987-01-01

    Thermo-hydraulic problems, in nuclear plants are normally analysed by the use of electrically heated rods. The direct or indirect heater rods are limited in their use because, for high temperatures and high heat flux, the heating element temperature approach its melting point. The use of platinum or tantalum is not economically viable. Graphite and carbon/carbon composites are alternative materials because they are good electrical conductors and have good mechanical properties at high temperatures. Graphite and carbon/carbon composites were used to make heating elements for testing by indirect heating. The swaging process used to reduce the cladding diameter prevented the fabrication of graphite heater rods. Carbon/carbon composite used to make heating elements gave good results up to a heat flux of 100 W/cm 2 . It is easy to verify that this value can be exceeded if the choice of the complementary materials for insulator and cladding improved. (author) [pt

  1. CONTROL ROD DRIVE

    Science.gov (United States)

    Chapellier, R.A.

    1960-05-24

    BS>A drive mechanism was invented for the control rod of a nuclear reactor. Power is provided by an electric motor and an outside source of fluid pressure is utilized in conjunction with the fluid pressure within the reactor to balance the loadings on the motor. The force exerted on the drive mechanism in the direction of scramming the rod is derived from the reactor fluid pressure so that failure of the outside pressure source will cause prompt scramming of the rod.

  2. Dynamic Rod Worth Measurement

    International Nuclear Information System (INIS)

    Chao, Y.A.; Chapman, D.M.; Hill, D.J.; Grobmyer, L.R.

    2000-01-01

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

  3. Correlation for cross-flow resistance coefficient using STAR-CCM+ simulation data for flow of water through rod bundle supported by spacer grid with split-type mixing vane

    Energy Technology Data Exchange (ETDEWEB)

    Agbodemegbe, V.Y., E-mail: vincevalt@gmail.com [Karlsruhe Institute of Technology, Institute of Fusion and Reactor Technique, Kaiserstrasse 12, Karlsruhe (Germany); Cheng, Xu, E-mail: xu.cheng@kit.edu [Karlsruhe Institute of Technology, Institute of Fusion and Reactor Technique, Kaiserstrasse 12, Karlsruhe (Germany); Akaho, E.H.K, E-mail: akahoed@yahoo.com [School of Nuclear and Allied Sciences, University of Ghana, PO Box AE 1, Kwabenya, Accra (Ghana); Allotey, F.K.A, E-mail: fkallotey@gmail.com [Institute of Mathematical Sciences, PO Box LG 197, Legon, Accra (Ghana)

    2015-04-15

    Highlights: • Investigate spacer grid with split-type mixing vanes. • Extent of predictability of experimental data by STAR-CCM+. • Reliability of two equation turbulence models. • Resistance to cross-flow through gaps. - Abstract: Mass transfer by diversion cross-flow through gaps is an important inter-subchannel interaction in fuel bundle of power reactors. It is normally due to the lateral pressure difference between adjacent sub-channels. This phenomenon is augmented in the presence of flow deflectors and is referred to as, directed cross-flow. Diversion cross-flow carries the momentum and energy of flow and hence affects the velocity and temperature profile in the rod bundle. The resistance to cross-flow in the transverse momentum equations is specified by the cross-flow resistant coefficient which is the subject of concern in the present study. In order to obtain data to correlate cross-flow resistance coefficient, computational fluid dynamic simulation using STAR-CCM+ was performed for flow of water at the bundle Reynolds number of Re1 = 3.4×10{sup 4} through a 5 × 5 rod bundle geometry supported by spacer grid with split mixing vanes for which the rod to rod pitch to diameter ratio was 1.33 and the rod to wall pitch to diameter ratio was 0.74. The two layer k-epsilon turbulence model with an all y+ automatic wall treatment function in STAR-CCM+ were adopted for an isothermal single phase (water) flow through the geometry. The objectives were to primarily investigate the extent of predictability of the experimental data by the computational fluid dynamic (CFD) simulation as a measure of reliability on the CFD code employed and also apply the simulation data to develop correlations for determining resistance coefficient to cross-flow. Validation of simulation results with experimental data showed good correlation of mean flow parameters with experimental data whiles turbulent fluctuations deviated largely from experimental trends. Generally, the

  4. 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.

  5. 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.)

  6. 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.)

  7. Control rod driveline and grapple

    International Nuclear Information System (INIS)

    Germer, J.H.

    1987-01-01

    A control rod driveline and grapple for engaging and releasing a control rod from a control rod drive is described comprising an enlarged control rod handle including an upwardly flaring frustum and a rod extending from the control rod handle; a relatively moving outer member; a tension rod connected to the relatively moving outer member at the upper end and provided with a lower annular flange at the lower end, the tension rod including a female cavity for receiving the upwardly extending rod from the enlarged control rod handle; a discrete and independent grapple segments for surrounding and grappling the control rod handle, each grapple segment including a first indentation for engaging and gripping the flange of the tension rod at an upper and interior annulus

  8. Rod Photoreceptors Detect Rapid Flicker

    Science.gov (United States)

    Conner, J. D.; MacLeod, Donald I. A.

    1977-01-01

    Rod-isolation techniques show that light-adapted human rods detect flicker frequencies as high as 28 hertz, and that the function relating rod critical flicker frequency to stimulus intensity contains two distinct branches. (MLH)

  9. 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.)

  10. Control rod testing apparatus

    Energy Technology Data Exchange (ETDEWEB)

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

    1987-06-02

    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.

  11. 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

  12. Distribution of impurities in monocrystalline Ge rods growth in a rectangular working zone by a group method

    CERN Document Server

    Okun, L S; Zatulovskii, L M; Levinzon, L I; Sachkov, G V; Chaikin, P M

    1973-01-01

    Seven rods were grown together, without rotation of former or crucible. The rods of diameter 8mm were grown at 2mm/min. Specific resistance and Hall effect measurements along the length of the rods revealed that there was a periodicity in these properties, having a 900 mu cycle. Specific resistance varied by 10 189184n a cycle. Single probe resistance measurements revealed a concentration of impurities at the surface of the rods. (3 refs).

  13. Heterologous expression of hydrophobins RodA and RodB from Aspergillus fumigatus in host Pichia pastoris

    DEFF Research Database (Denmark)

    Pedersen, Mona Højgaard; Borodina, Irina; Frisvad, Jens Christian

    Introduction: Hydrophobins are small amphipatic proteins present on the spore surface of filamentous fungi. They most likely play an important role in the attachment of spores to a solid phase. The pathogenic fungus Aspergillus fumigatus expresses the hydrophobins RodA and RodB on the surface...... of its conidia and these may be of importance to the pathogenesis of the fungus. Although heterologous expression of hydrophobins has proven to be a challenge by past investigators, we made it the aim of this project to produce pure hydrophobins in sufficient quantities for further characterication...... and transformants were selected by zeocin resistance. The presence of the RodA and RodB genes in the transformants was confirmed by colony PCR. The expression of RodA and RodB genes was induced by growing cells in culture flasks for three days in buffered complex methanol medium as protein production was dependent...

  14. HIGH STRENGTH CONTROL RODS FOR NEUTRONIC REACTORS

    Science.gov (United States)

    Lustman, B.; Losco, E.F.; Cohen, I.

    1961-07-11

    Nuclear reactor control rods comprised of highly compressed and sintered finely divided metal alloy panticles and fine metal oxide panticles substantially uniformly distributed theretbrough are described. The metal alloy consists essentially of silver, indium, cadmium, tin, and aluminum, the amount of each being present in centain percentages by weight. The oxide particles are metal oxides of the metal alloy composition, the amount of oxygen being present in certain percentages by weight and all the oxygen present being substantially in the form of metal oxide. This control rod is characterized by its high strength and resistance to creep at elevated temperatures.

  15. Control rod assemblies

    International Nuclear Information System (INIS)

    Yamanaka, Toshikatsu.

    1986-01-01

    Purpose: To obtain simple and practical control rod assemblies by bringing the exit temperature of the guide tube of a control rod main body closer to that of an adjacent fuel assembly and thereby suppressing the wasteful flow of coolants. Constitution: A flow control member comprises an annular flow control plate disposed above the control rod main body and bellows having a plurality of small paertures capable of passing coolants therethrough formed at the circumferencial surface. The bellows are to cause the flow control plate to resiliently abut on the upper surface of the control rod main body. Coolants flowing from below to above in the guide tube remove heat from the neutron absorbers and are discharged externally at an elevated temperature, while coolants at a lower temperature are entered and mixed through the apertures formed in the bellows. By the way, upon insertion of the control rod main body, flow of the coolants to the inside of the bellows is substantially interrupted by the extension contraction of the bellows, by which the flow rate is adjusted depending on the withdrawing stroke to suppress the occurrence of thermal problems. (Kamimura, M.)

  16. 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

  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. 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.)

  19. Flow in rod bundles

    International Nuclear Information System (INIS)

    Hazi, G.; Mayer, G.

    2005-01-01

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

  20. Attracting electromagnet for control rod

    International Nuclear Information System (INIS)

    Kato, Kazuo; Sasaki, Kotaro.

    1989-01-01

    Non-magnetic material plates with inherent resistivity of greater than 20 μΩ-cm and thickness of less than 3 mm are used for the end plates of attracting electromagnets for closed type control rods. By using such control rod attracting electromagnets, the scram releasing time can be shortened than usual. Since the armature attracting side of the electromagnet has to be sealed by a non-magnetic plate, a bronze plate of about 5 mm thickness has been used so far. Accordingly, non-magnetic plate is inserted to the electromagnet attracting face to increase air source length for improving to shorten the scram releasing time. This method, however, worsens the attracting property on one hand to require a great magnetomotive force. For overcoming these drawbacks, in the present invention, the material for tightly closing end plates in an electromagnet is changed from bronze plate to non-magnetic stainless steel SUS 303 or non-magnetic Monel metal and, in addition, the plate thickness is reduced to less than 5 mm thereby maintaining the attracting property and shortening the scram releasing time. (K.M.)

  1. 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.)

  2. Control rod driving mechanisms

    International Nuclear Information System (INIS)

    Maejima, Yoshinori.

    1986-01-01

    Purpose: To conduct reactor scram by an external signal and, also by a signal for the abnormal temperature from a temperature detector in the nuclear reactor. Constitution: Control rod driving mechanisms magnetically coupling the extension pipe with the elevating mechanism above the reactor core and the holding magnet, and retains a control rod to the lower portion of the extension pipe by way of a latch mechanism. The temperature detector is immersed in reactor coolants. If the temperature of the coolants rises abnormally, bimetal contacts of the temperature detector are opened to interrupt the current supply to the holding electromagnet. Then, the extension pipe released from the magnetic coupling is lowered and the control rod free from latch is rapidly dropped and inserted into the reactor core. Since this procedure is carried out for all of the control rods, the reactor scram can be attained. The feature of this invention resides in that the reactor scram can be attained also by the signal of the reactor core itself even if the signal system for the external signals should be failed. (Horiuchi, T.)

  3. 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.

  4. 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)

  5. Fuel rod attachment system

    International Nuclear Information System (INIS)

    Christiansen, D.W.

    1982-01-01

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

  6. Fuel rod fixing system

    International Nuclear Information System (INIS)

    Christiansen, D.W.

    1982-01-01

    This is a reusable system for fixing a nuclear reactor fuel rod to a support. An interlock cap is fixed to the fuel rod and an interlock strip is fixed to the support. The interlock cap has two opposed fingers, which are shaped so that a base is formed with a body part. The interlock strip has an extension, which is shaped so that this is rigidly fixed to the body part of the base. The fingers of the interlock cap are elastic in bending. To fix it, the interlock cap is pushed longitudinally on to the interlock strip, which causes the extension to bend the fingers open in order to engage with the body part of the base. To remove it, the procedure is reversed. (orig.) [de

  7. 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.)

  8. Control rod drives

    International Nuclear Information System (INIS)

    Kimura, Koichi.

    1994-01-01

    In control rod drives, differential pressure sensors are disposed at the inlet and the exit of a driving water pressure control valve disposed in a driving water supply device and, when deviation of fluctuation of the differential pressure from a set value is detected, a pressure control valve for driving water is controlled so as to make the differential pressure constant. The differential pressure sensors detect the differential pressure between the pressure of the control rod drives at the inlet and the exit of the driving water pressure control valve and a pressure in a reactor dome. A judging circuit judges whether the differential pressure between both sides of the driving water pressure control valve is deviated from a set value or not and, if it deviates from the set value, outputs of judging signal to the control device. In the control device, the opening degree of the driving water pressure control valve is controlled, so that the differential pressure between both sides of the driving water pressure control value is constant and does not deviate from the set value. There are provided advantageous effects of preventing abnormal control rod withdrawing phenomenon to improve safety and reliability for the control of the reactor operation. (N.H.)

  9. 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)

  10. Electronic circuit for control rod attracting electromagnet

    International Nuclear Information System (INIS)

    Ito, Koji.

    1991-01-01

    The present invention provides a discharging circuit for control rod attracting electromagnet used for a reactor which is highly reliable and has high performance. The resistor of the circuit comprises a non-linear resistor element and a blocking rectification element connected in series. The discharging circuit can be prevented from short-circuit by selecting a resistor having a resistance value about ten times as great as the coil resistance, even in a case where the blocking rectification element and the non-linear resistor element are failed. Accordingly, reduction of attracting force and the increase of scream releasing time can be minimized. (I.S.)

  11. Cone rod dystrophies

    Directory of Open Access Journals (Sweden)

    Hamel Christian P

    2007-02-01

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

  12. 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

  13. 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.)

  14. Nuclear reactor with control rods

    International Nuclear Information System (INIS)

    Obermeyer, F.D.; Berringer, R.T.

    1979-01-01

    A liquid-cooled nuclear reactor including fuel assemblies mounted within a reactor vessel having linearly movable control rods passing through control rod guide tubes into respective aligned fuel assemblies is described. Reactor coolant circulates through the assemblies. Guide tubes and other vessel internals structures located above the assemblies and is discharged through an outlet nozzle positioned above the elevation of primary flow openings in the guide tube walls. The guide tube includes internal horizontal supports and a length limited continuous control rod guide which, in conjunction with the flow openings, alleviate detrimental coolant cross flows and frictional restraints imposed upon the control rods

  15. Nuclear fuel rods

    International Nuclear Information System (INIS)

    Wada, Toyoji.

    1979-01-01

    Purpose: To remove failures caused from combination of fuel-cladding interactions, hydrogen absorptions, stress corrosions or the likes by setting the quantity ratio of uranium or uranium and plutonium relative to oxygen to a specific range in fuel pellets and forming a specific size of a through hole at the center of the pellets. Constitution: In a fuel rods of a structure wherein fuel pellets prepared by compacting and sintering uranium dioxide, or oxide mixture consisting of oxides of plutonium and uranium are sealed with a zirconium metal can, the ratio of uranium or uranium and plutonium to oxygen is specified as 1 : 2.01 - 1 : 2.05 in the can and a passing hole of a size in the range of 15 - 30% of the outer diameter of the fuel pellet is formed at the center of the pellet. This increases the oxygen partial pressure in the fuel rod, oxidizes and forms a protection layer on the inner surface of the can to control the hydrogen absorption and stress corrosion. Locallized stress due to fuel cladding interaction (PCMI) can also be moderated. (Horiuchi, T.)

  16. Comprehensive Evaluation of the MBT STAR-BL Module for Simultaneous Bacterial Identification and β-Lactamase-Mediated Resistance Detection in Gram-Negative Rods from Cultured Isolates and Positive Blood Cultures.

    Science.gov (United States)

    Lee, Annie W T; Lam, Johnson K S; Lam, Ricky K W; Ng, Wan H; Lee, Ella N L; Lee, Vicky T Y; Sze, Po P; Rajwani, Rahim; Fung, Kitty S C; To, Wing K; Lee, Rodney A; Tsang, Dominic N C; Siu, Gilman K H

    2018-01-01

    Objective: This study evaluated the capability of a MALDI Biotyper system equipped with the newly introduced MBT STAR-BL module to simultaneously perform species identification and β-lactamase-mediated resistance detection in bacteremia -causing bacteria isolated from cultured isolates and patient-derived blood cultures (BCs). Methods: Two hundred retrospective cultured isolates and 153 prospective BCs containing Gram-negative rods (GNR) were collected and subjected to direct bacterial identification, followed by the measurement of β-lactamase activities against ampicillin, piperacillin, cefotaxime, ceftazidime, and meropenem using the MBT STAR-BL module. The results and turnaround times were compared with those of routine microbiological processing. All strains were also characterized by beta-lactamase PCR and sequencing. Results: Using the saponin-based extraction method, MALDI-TOF MS correctly identified bacteria in 116/134 (86.6%) monomicrobial BCs. The detection sensitivities for β-lactamase activities against ampicillin, piperacillin, third-generation cephalosporin and meropenem were 91.3, 100, 97.9, and 100% for cultured isolates, and 80.4, 100, 68.8, and 40% for monomicrobial BCs ( n = 134) respectively. The overall specificities ranged from 91.5 to 100%. Furthermore, the MBT STAR-BL and conventional drug susceptibility test results were concordant in 14/19 (73.7%) polymicrobial cultures. Reducing the logRQ cut-off value from 0.4 to 0.2 increased the direct detection sensitivities for β-lactamase activities against ampicillin, cefotaxime and meropenem in BCs to 85.7, 87.5, and 100% respectively. The MBT STAR-BL test enabled the reporting of β-lactamase-producing GNR at 14.16 and 47.64 h before the interim and final reports of routine BCs processing, respectively, were available. Conclusion: The MALDI Biotyper system equipped with the MBT STAR-BL module enables the simultaneous rapid identification of bacterial species and

  17. Rod Microglia: A Morphological Definition

    Science.gov (United States)

    Taylor, Samuel E.; Morganti-Kossmann, Cristina

    2014-01-01

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

  18. Reactor fuel rod

    International Nuclear Information System (INIS)

    Inui, Mitsuhiro; Mori, Kazuma.

    1990-01-01

    In a high burnup degree reactor core, a problem of fuel can corrosion caused by coolants occurs due to long stay in a reactor. Then, the use of fuel cladding tubes with improved corrosion resistance is now undertaken and use of corrosion resistant alloys is attempted. However, since the conventional TIG welding melts the entire portion, the welded portion does not remain only in the corrosive resistant alloy but it forms new alloys of the corrosion resistant alloy and zircaloy as the matrix material or inter-metallic compounds, which degrades the corrosion resistance. In the present invention, a cladding tube comprising a dual layer structure using a corrosion resistant alloy only for a required thickness and an end plug made of the same material as the corrosion resistant alloy are welded at the junction portion by using resistance welding. Then, they are joined under welding by the heat generated to the junction surfaces between both of them, to provide corrosion resistant alloys substantially at the outside of the welded portion as well. Accordingly, the corrosion resistance is not degradated. (T.M.)

  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. 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

  1. Control rod drives

    International Nuclear Information System (INIS)

    Kono, Nobuaki.

    1987-01-01

    Purpose: To remove movable portion and improve the reliability by the direct control to coil. Constitution: Coils are disposed vertically at a predetermined interval to the outside of a control rod drive guide tube and each of the coils is adapted to be directly controlled. The coils are arranged at such an interval that a plunger laps over the vertically adjacent coils. In the case of moving the plunger upwardly, a coil just above the coil that attract the plunger is energized while the coil attracting the plunger so far is denergized. Then, the plunger is pulled up to an aimed position by repeating the procedures. In the case of moving the plunger downwardly, the procedures are conducted in the manner opposite to the above. (Kawakami, Y.)

  2. Fuel rod for a reactor

    International Nuclear Information System (INIS)

    Tsuboi, Yoshiaki.

    1976-01-01

    Object: To accurately and simply measure gas pressure within a cladding tube of a fuel rod. Structure: The fuel rod is closed by an end plug with pellets made of uranium dioxide and a pressure detector element sealed into the cladding tube. When the fuel rod is manufactured, helium gases are introduced under pressure into the cladding tube, and the pressure detector element is contracted proportionally by the aforesaid pressure and therefore the amount of contract may be measured to thereby measure the inside pressure of the cladding tube. The amount of contract of the pressure detector element may be measured exteriorly of the fuel rod by arranging a fluorescent screen or film for X-rays or other radiations on one side of the fuel rod. (Yoshino, Y.)

  3. Freon 7-rod cluster subchannel mixing experiments

    International Nuclear Information System (INIS)

    Bowring, R.W.; Levy, L.

    1969-12-01

    The subchannel thermal-hydraulic characteristics of a 7-rod cluster were measured using boiling freon-12, modelling water at 1,000 psia. The heated length was 36 inches, the rod diameter 0.625 inches and rod gap 0.060 inches. The following were determined from the experiments: lateral resistance to cross-flow, the diversion length, cross-flow enthalpy, the fraction of heat retained in the highly rated subchannel and the turbulent mixing factor, F m , used in HAMBO. The mixing factor was found to decrease when boiling started but was then independent of vapour quality, within the experimental error. It also decreased more rapidly with mass velocity in the boiling than in the non-boiling regime. The cross-flow enthalpy was found to be a continuous function with cross-flow and not that of the donor subchannel; the enthalpy appeared to be intermediate between that of the two subchannels but the possibility of steam enhancement could not be discounted. (author)

  4. 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.)

  5. 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

  6. Control rod drives

    International Nuclear Information System (INIS)

    Sato, Takao; Arita, Setsuo; Mizuno, Katsuhiro.

    1986-01-01

    Purpose: To enable fine positioning by using an induction motor of a simple structure as a driving source and thereby improve the reliability of control rod drives. Constitution: A step actuator is directly coupled with an induction motor, in which the induction motor is connected by way of a pulse driving control circuit to an AC power source, while the step actuator is connected to a DC power source. When a thyristor is turned ON, the motor outputs a positive torque and rotates and starts to rotate in the forward direction. When the other thyristor is turned ON, the motor is applied with braking by a reverse excitation in a manner equivalent to the change for the exciting phase sequence. When the speed is lowered to a predetermined value, braking is actuated by the torque of the step actuator and the motor stops at a zero position or balanced position. In this way, braking is actuated from the decelerating step to the stopping with no abrasion and a highly accurate positioning is possible due to the characteristics of the step actuator. (Horiuchi, T.)

  7. Control rod drive

    International Nuclear Information System (INIS)

    Watando, Kosaku; Tanaka, Yuzo; Mizumura, Yasuhiro; Hosono, Kazuya.

    1975-01-01

    Object: To provide a simple and compact construction of an apparatus for driving a drive shaft inside with a magnetic force from the outside of the primary system water side. Structure: The weight of a plunger provided with an attraction plate is supported by a plunger lift spring means so as to provide a buffer action at the time of momentary movement while also permitting the load on lift coil to be constituted solely by the load on the drive shaft. In addition, by arranging the attraction plate and lift coil so that they face each other with a small gap there-between, it is made possible to reduce the size and permit efficient utilization of the attracting force. Because of the small size, cooling can be simply carried out. Further, since there is no mechanical penetration portion, there is no possibility of leakage of the primary system water. Furthermore, concentration of load on a latch pin is prevented by arranging so that with a structure the load of the control rod to be directly beared through the scrum latch. (Kamimura, M.)

  8. Nuclear reactor control rod

    International Nuclear Information System (INIS)

    Cearley, J.E.; Izzo, K.R.

    1987-01-01

    This patent describes a vertically oriented bottom entry control rod from a nuclear reactor: a frame including an elongated central spine of cruciform cross section connected between an upper support member and a lower support member both of cruciform shape having four laterally extending arms. The arms are in alignment with the arms of the lower support member and each aligned upper and lower support members has a sheath extending between; absorber plates of neutron absorber material, different from the material of the frame, one of the absorber plates is positioned within a sheath beneath each of the arms; attachment means suspends the absorber plates from the arms of the upper support member within a sheath; elongated absorber members positioned within a sheath between each of the suspended absorber plates and an arm of the lower support member; and joint means between the upper ends of the absorber members and the lower ends of the suspended absorber plates for minimizing gaps; the sheath means encloses the suspended absorber plates and the absorber members extending between aligned arms of the upper and lower support members and secured

  9. Hydraulic testing of accelerator-production-of-tritium rod bundles

    International Nuclear Information System (INIS)

    Spatz, T.L.; Siebe, D.A.

    1999-01-01

    Hydraulic tests have been performed on small pitch-to-diameter-ratio rod bundles using light water (1.7 Tr < 13,000). Also presented is the comparison of the overall rung pressure drop to a solution based on hydraulic-resistance handbook calculations

  10. Means for driving control rod

    International Nuclear Information System (INIS)

    Sato, Haruo; Sasaki, Masayoshi.

    1974-01-01

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

  11. 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

  12. 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...

  13. 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.)

  14. Advanced gray rod control assembly

    Energy Technology Data Exchange (ETDEWEB)

    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.

  15. 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

  16. Lateral Penetration of a Rod

    Science.gov (United States)

    Alston, James; Bless, Stephan; Subramanian, Ravi

    2002-03-01

    Penetration of yawed rods remains one of the outstanding problems in terminal ballistics. An essential feature of yawed rod penetration is the interaction of the shank of the projectile with the side of the penetration cavity. A two-dimensional finite difference code was used to solve this problem for the case of a projectile with a circular cross section penetrating armor steel. This case is particularly relevant for the problem of a high velocity high density rod penetrating a finite plate. The force exerted by the target on the projectile was determined as a function of embedment depth and lateral velocity. The solution was verified by checking the centerline pressure against the closed form solution for cylindrical cavity expansion

  17. 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

  18. Control rod housing alignment 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 paper discusses an alignment device for precisely locating the position of the top of a control rod drive housing from an overlying and corresponding hole and alignment pin in a core plate within a boiling water nuclear reactor. It includes a shaft, the shaft having a length sufficient to extend from the vicinity of the top of the control rod drive housing up to and through the hole in the core plate; means for registering the top of the shaft to the hole in the core plate, the registering means including means for registering with an alignment pin in the core plate adjacent the hole

  19. Thermal-hydraulic stability tests for newly designed BWR rod bundle (step-III fuel type A)

    International Nuclear Information System (INIS)

    Mitsutake, T.; Chuman, K.; Miura, S.; Morooka, S.; Moriya, K.; Kitamura, H.; Toba, A.; Omoto, A.

    2004-01-01

    Thermal-hydraulic stability tests have been performed on electrically heated bundles to simulate the newly designed Boiling Water Reactor (BWR) fuels in a parallel channel test loop. The objective of the current experimental program is to investigate how the newly designed bundle could improve the thermal-hydraulic stability. Measurements of the thermal-hydraulic instability thresholds in two vertical rod bundles have been conducted in steam-water two-phase flow conditions at the TOSHIBA test loop. Fluid conditions were BWR operating conditions of 7 MPa system pressure, 1.0-2.0x10 6 kg/m 2 /h inlet mass flux and 28-108 kJ/kg inlet subcooling. The parallel channel test loop consists of a main bundle of 3x3 indirectly heated rods of 1/9 symmetry of 9x9 full lattice and a bypass bundle of 8x8. These are both simulated BWR rod bundles in respect of rod diameter, heated length, rod configuration, fuel rod spacer, core inlet hydraulic resistance and upper tie plate. There are three types of the 3x3 test bundles with different configurations of a part length rod of two-thirds the length of the other rods and an axial power shape. The design innovation of the part length rod for a 9x9 lattice development, though addition of more fuel rods increases bundle pressure drop, reduces pressure drop in the two-phase portion of the bundle, and enhances the thermal hydraulic stability. Through the experiments, the parameter dependency on the channel stability threshold is obtained for inlet subcooling, inlet mass flux, inlet flow resistance, axial power shape and part length rod. The main conclusion is that the stability threshold is about 10% greater with the part length rod than without the part length rod. The new BWR bundle consisting of the part length rod has been verified in respect of thermal hydraulic stability performance. (author)

  20. Control rod removal blocking device

    International Nuclear Information System (INIS)

    Yoshioka, Ritsuo.

    1981-01-01

    Purpose: To prevent excess power increase resulted from erroneous control rod removal during high power operation in BWR type reactor by decreasing the continuous removal enabling distance for the control rods along with increase in the reactor power where the reactor power is greater than a predetermined level. Constitution: When control rod selection signals are supplied from a control unit to a control rod removal blocking device, the blocking device judges whether the reactor core power is greater than a predetermined value A or not, using reactor core power signals outputted from an average power monitor. Where the reactor core power exceeds the predetermined value A and if the reactor core power is relatively low, a large continuous removal enabling distance N 1 is calculated in the blocking device to allow the continuous removal as far as the notch N 1 . The continuous removal enabling distance is shortened as the reactor core power increases and the removal is blocked, for example, at notch N 2 . While on the other hand, if the reactor core power is below the predetermined value A, both of the notchwise removal and the continuous removal are enabled. (Seki, T.)

  1. Control rod behaviour in earthquakes

    International Nuclear Information System (INIS)

    Kawakami, S.; Akiyama, H.; Shibata, H.; Watabe, M.; Ichikawa, T.; Fujita, K.

    1990-01-01

    For some years the Japanese have been working on a major research programme to determine the likely effects of an earthquake on nuclear plant internals. One aspect of this was a study of the behaviour of Pressurized Water Reactor control rods as they are being inserted in the core, which is reported here. (author)

  2. 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.)

  3. 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

  4. 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

  5. Solid-state-laser-rod holder

    Science.gov (United States)

    Gettemy, D.J.; Barnes, N.P.; Griggs, J.E.

    1981-08-11

    The disclosure relates to a solid state laser rod holder comprising Invar, copper tubing, and epoxy joints. Materials and coefficients of expansion of the components of the holder combine with the rod to produce a joint which will give before the rod itself will. The rod may be lased at about 70 to 80/sup 0/K and returned from such a temperature to room temperature repeatedly without its or the holder's destruction.

  6. 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

  7. 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 ...

  8. 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.)

  9. 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.)

  10. 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...

  11. Rod and lamellar growth of eutectic

    Directory of Open Access Journals (Sweden)

    M. Trepczyńska-Łent

    2010-04-01

    Full Text Available The paper presents adaptation problem of lamellar growth of eutectic. The formation of rod eutectic microstructure was investigated systematically. A new rod eutectic configuration was observed in which the rods form with elliptical cylindrical shape. A new interpretation of the eutectic growth theory was proposed.

  12. 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.

  13. 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

  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 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

  15. Fuel rod assembly to manifold attachment

    Science.gov (United States)

    Donck, Harry A.; Veca, Anthony R.; Snyder, Jr., Harold J.

    1980-01-01

    A fuel element is formed with a plurality of fuel rod assemblies detachably connected to an overhead support with each of the fuel rod assemblies having a gas tight seal with the support to allow internal fission gaseous products to flow without leakage from the fuel rod assemblies into a vent manifold passageway system on the support. The upper ends of the fuel rod assemblies are located at vertically extending openings in the support and upper threaded members are threaded to the fuel rod assemblies to connect the latter to the support. The preferred threaded members are cap nuts having a dome wall encircling an upper threaded end on the fuel rod assembly and having an upper sealing surface for sealing contact with the support. Another and lower seal is achieved by abutting a sealing surface on each fuel rod assembly with the support. A deformable portion on the cap nut locks the latter against inadvertent turning off the fuel rod assembly. Orienting means on the fuel rod and support primarily locates the fuel rods azimuthally for reception of a deforming tool for the cap nut. A cross port in the fuel rod end plug discharges into a sealed annulus within the support, which serves as a circumferential chamber, connecting the manifold gas passageways in the support.

  16. Control-rod scram device

    International Nuclear Information System (INIS)

    Matsui, Yoshiro; Saito, Koji.

    1986-01-01

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

  17. Fuel rod and fuel assembly

    International Nuclear Information System (INIS)

    Takekawa, Tetsuya.

    1993-01-01

    Burnable poisons are contained in a portion of a pellet constituting a fuel rod. A distribution density of the burnable poison-containing pellets and a concentration of the burnable poisons in the pellet are varied depending on the axial position of the fuel rod. That is, the distribution density of the burnable poison containing-pellets is increased at the central portion of the fuel rod and it is decreased at both ends thereof, and a concentration of the burnable poisons of the burnable poison containing-pellet disposed at the end portions thereof is decreased to less than a concentration of the burnable poison-containing pellet at the central portion. With such a constitution, a central peaking at an early stage of the combustion cycle is decreased. Accordingly, power at the central portion is increased than that in the end portions at the latter half of the cycle, to flatten the power distribution. Further, a burnable poison concentration of the pellets at the end portions is decreased to promote burning of burnable poisons at the end portions which are less burnable relatively, thereby enabling to prevent worsening of neutron economy. (T.M.)

  18. Comparative Study Between Cobalt Chrome and Titanium Alloy Rods for Multilevel Spinal Fusion: Proximal Junctional Kyphosis More Frequently Occurred in Patients Having Cobalt Chrome Rods.

    Science.gov (United States)

    Han, Sanghyun; Hyun, Seung-Jae; Kim, Ki-Jeong; Jahng, Tae-Ahn; Kim, Hyun-Jib

    2017-07-01

    The use of titanium alloy (Ti) rods is frequently associated with rod fracture after spinal fixation. To address this issue, cobalt chrome (CoCr) rods, which are advantageous because of their greater strength and resistance to fatigue relative to Ti rods, have been introduced. The purpose of the present study was to compare radiographic outcomes after the use of Ti versus CoCr rods in a matched cohort of patients undergoing posterior spinal fusion for treatment of spinal instability. We retrospectively reviewed data from patients who had undergone spinal fusion involving more than 3 levels at a single institution between 2004 and 2015. Patients were matched for age, diagnosis, 3-column osteotomy, levels fused, and T score. Fifty patients with Ti rods were identified and appropriately matched to 50 consecutive patients with CoCr rods. The distributions of age at surgery, sex, diagnosis, 3-column osteotomy, levels fused, number of patients with previous surgical procedures, and T score did not significantly differ between the 2 groups. However, there were significant differences in length of follow-up (CoCr, 25.0 vs. Ti, 28.5 months; P < 0.001), fusion rate (CoCr, 45 [90%] vs. Ti, 33 [66%]; P = 0.004), occurrence of rod breakage (CoCr, 0 vs. T, 8 [16%]; P = 0.006), and junctional kyphosis (CoCr, 24 [46%] vs. Ti, 9 [18%]; P = 0.003). Our findings indicate that the use of CoCr rods is effective in ensuring stability of the posterior spinal construct and accomplishment of spinal fusion. Furthermore, our results indicate that junctional kyphosis may occur more frequently in CoCr systems than in Ti systems. Copyright © 2017 Elsevier Inc. All rights reserved.

  19. 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

  20. High temperature control rod assembly

    Science.gov (United States)

    Vollman, Russell E.

    1991-01-01

    A high temperature nuclear control rod assembly comprises a plurality of substantially cylindrical segments flexibly joined together in succession by ball joints. The segments are made of a high temperature graphite or carbon-carbon composite. The segment includes a hollow cylindrical sleeve which has an opening for receiving neutron-absorbing material in the form of pellets or compacted rings. The sleeve has a threaded sleeve bore and outer threaded surface. A cylindrical support post has a threaded shaft at one end which is threadably engaged with the sleeve bore to rigidly couple the support post to the sleeve. The other end of the post is formed with a ball portion. A hollow cylindrical collar has an inner threaded surface engageable with the outer threaded surface of the sleeve to rigidly couple the collar to the sleeve. the collar also has a socket portion which cooperates with the ball portion to flexibly connect segments together to form a ball and socket-type joint. In another embodiment, the segment comprises a support member which has a threaded shaft portion and a ball surface portion. The threaded shaft portion is engageable with an inner threaded surface of a ring for rigidly coupling the support member to the ring. The ring in turn has an outer surface at one end which is threadably engageably with a hollow cylindrical sleeve. The other end of the sleeve is formed with a socket portion for engagement with a ball portion of the support member. In yet another embodiment, a secondary rod is slidably inserted in a hollow channel through the center of the segment to provide additional strength. A method for controlling a nuclear reactor utilizing the control rod assembly is also included.

  1. Control rod for a reactor

    International Nuclear Information System (INIS)

    Natori, Hisahide.

    1975-01-01

    Object: To change arrangement and density of each layer of neutron absorber in the control rod and to render rotation by each layer possible, whereby the neutron absorber may be rotated to readily flatten power distribution. Structure: Neutron absorbers such as boron and carbide are filled into stainless steel pipes, which are peripherally arranged in a multi-layer fashion. Arrangement and density of the neutron absorber by each layer are changed and rotation by each layer is made possible, whereby surface area of the absorber or the like is changed to flatten power distribution. (Furukawa, Y.)

  2. Laboratory experiments with impacting fuel rods

    International Nuclear Information System (INIS)

    Kiss, S.; Lipcsei, S.

    1994-10-01

    Vibration surveillance and diagnostics of fuel rods and fuel assemblies are important tasks in NPPs. Thus accurate knowledge of vibration phenomena and measurability is very important. Experimental results on models without limiter give good coincidence with theoretical calculations. Spectra measured on impacting rod become smoother with increasing impacting level. Spectra of fuel rods have a wider range in impacting rate and higher level of smoothing than spectra of model rod have. The impacting rate strongly depends on mechanical properties of the rod. By the experiments, one can state that as for Fourier spectra the only thing caused by the impacts is the smoothening. However, there is a chance to give faulty diagnosis by Fourier spectra only. Consequently, investigation of fuel rod vibration requires increased caution. (author) 4 refs.; 12 figs.; 1 tab

  3. 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

  4. Automatic safety rod for reactors. [LMFBR

    Science.gov (United States)

    Germer, J.H.

    1982-03-23

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

  5. 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.

  6. 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.)

  7. 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

  8. 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.

  9. 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

  10. Dry rod consolidation technology development

    International Nuclear Information System (INIS)

    Rasmussen, T.L.; Schoonen, D.H.; Fisher, M.W.

    1986-01-01

    The Department of Energy's (DOE) Office of Civilian Radioactive Waste Management (OCRWM) is funding a Program to consolidate commercial spent fuel for testing in dry storage casks and to develop technology that will be fed into other OCRWM Programs, e.g., Prototypical Consolidation Demonstration Program. The Program is being conducted at the Idaho National Engineering Laboratory (INEL) by the Operating Contractor, EGandG Idaho, Inc. Hardware and software have been designed and fabricated for installation in a hot cell adjacent to the Test Area North (TAN) Hot Shop Facility. This equipment will be used to perform dry consolidation of commercial spent fuel from the Virginia Power (VP) Cooperative Agreement Spent Fuel Storage Cask (SPSC) Demonstration Program and assemblies that had previously been stored at the Engine Maintenance and Disassembly (EMAD) facility in Nevada. Consolidation will be accomplished by individual, horizontal rod pulling. A computerized semi-automatic control system with operator involvement will be utilized to conduct consolidation operations. Special features have been incorporated in the design to allow crud collection and measurement of rod pulling forces. During consolidation operations, data will be taken to characterize this technology. Still photo, video tape, and other documentation will be generated to make developed information available to interested parties. Cold checkout of the hardware and software will complete in September of 1986. Following installation in the hot cell, consolidation operations will begin in January 1987. Resulting consolidated fuel will be utilized in the VP Cooperative Agreement SFSC Program

  11. Wall pressure fluctuations in rod bundles

    International Nuclear Information System (INIS)

    Moeller, S.V.

    1990-01-01

    Microphones and hot wires were applied for the measurement of wall pressure fluctuations and velocity fluctuations in rod bundles with several aspect ratios. By means of auto and cross spectral density functions their interdependence was investigated. Results show that the pressure fluctuations in rod bundles are mainly associated with the phenomenon of quasi-periodic flow pulsations between subchannels. (author)

  12. Tipping Time of a Quantum Rod

    Science.gov (United States)

    Parrikar, Onkar

    2010-01-01

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

  13. Study of the rod style SFRFQ structure

    CERN Document Server

    Yan Xue Qing; Chen J

    2002-01-01

    There is a problem about upper limit of energy in the RFQ structure, although it is a wonderful low-energy-suited high current accelerating structure. After proposing an improved rod style SFRFQ structure without reversed field, the author studies its energy gain and transverse motion. The rod style SFRFQ structure is roughly compared with diaphragm SFRFQ structure

  14. 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.)

  15. Control rod controlling device of nuclear reactor

    International Nuclear Information System (INIS)

    Arita, Setsuo; Okido, Fumiyasu.

    1997-01-01

    The present invention concerns a control rod drive mechanism for use in a BWR, which does not apply undesired effects on monitoring of neutron instrumentation systems. Control rods are operated using an induction electric motor equipped with an electromagnetic brake as a driving source. The induction electric motor and the electromagnetic brake are driven by ON/OFF control. Since a switching element for driving the induction electric motor and the electromagnetic brake can be kept ON or OFF during control rod operation, electromagnetic noises are not generated during the operation of the control rods. Accordingly, the neutron instrumentation systems do not undergoing effects of electromagnetic noises during operation of control rods, and the neutron instrumentation system can accurately be monitored. (N.H.)

  16. 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

  17. Comparison of the Resistance to Bending Forces of the 4.5 LCP Plate-rod Construct and of 4.5 LCP Alone Applied to Segmental Femoral Defects in Miniature Pigs

    Directory of Open Access Journals (Sweden)

    Lucie Urbanová

    2010-01-01

    Full Text Available The study deals with the determination of mechanical properties, namely resistance to bending forces, of flexible buttress osteosynthesis using two different bone-implant constructs stabilizing experimental segmental femoral bone defects (segmental ostectomy in a miniature pig ex vivo model using 4.5 mm titanium LCP and a 3 mm intramedullary pin (“plate and rod” construct (PR-LCP, versus the 4.5 mm titanium LCP alone (A-LCP. The “plate and rod” fixation (PR-LCP of the segmental femoral defect is significantly more resistant (p in vivo experiments in the miniature pig to investigate bone defect healing after transplantation of mesenchymal stem cells in combination with biocompatible scaffolds.

  18. 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.

  19. 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.)

  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. 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

  2. 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

  3. Control rod drive hydraulic device

    International Nuclear Information System (INIS)

    Takekawa, Toru.

    1994-01-01

    The device of the present invention can reliably prevent a possible erroneous withdrawal of control rod driving mechanism when the pressure of a coolant line is increased by isolation operation of hydraulic control units upon periodical inspection for a BWR type reactor. That is, a coolant line is connected to the downstream of a hydraulic supply device. The coolant line is connected to a hydraulic control unit. A coolant hydraulic detection device and a pressure setting device are disposed to the coolant line. A closing signal line and a returning signal line are disposed, which connect the hydraulic supply device and a flow rate control valve for the hydraulic setting device. In the device of the present invention, even if pressure of supplied coolants is elevated due to isolation of hydraulic control units, the elevation of the hydraulic pressure can be prevented. Accordingly, reliability upon periodical reactor inspection can be improved. Further, the facility is simplified and the installation to an existent facility is easy. (I.S.)

  4. Taylor impact of glass rods

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  5. Dry rod consolidation technology development

    International Nuclear Information System (INIS)

    Rasmussen, T.L.; Schoonen, D.H.; Feldman, E.M.; Fisher, M.W.

    1987-01-01

    The Department of Energy's (DOE) Office of Civilian Radioactive Waste Management (OCRWM) is funding a program to consolidate commercial spent fuel for testing in dry storage casks and to develop technology that will be fed into other OCRWM programs, e.g., Prototypical Consolidation Demonstration Program (PCDP). The program is being conducted at the Idaho National Engineering Laboratory (INEL) by the INEL Operating Contractor EG and G Idaho, Inc. Hardware and software have been designed and fabricated for installation in a hot cell adjacent to the Test Area North (TAN) Hot Shop Facility. This equipment is used to perform dry consolidation of commercial spent fuel from the Virginia Power (VP) Cooperative Agreement Spent Fuel Storage Cask (SFSC) Demonstration Program and assemblies that had previously been stored at the Engine Maintenance and Disassembly (EMAD) facility in Nevada. Consolidation is accomplished by individual, horizontal rod pulling. A computerized semiautomatic control system with operator involvement is utilized to conduct consolidation operations. During consolidation operations, data is taken to characterize this technology. Still photo, video tape, and other documentation will be generated to make developed information available to interested parties. Cold checkout of the hardware and software was completed in September of 1986. Following installation in the hot cell, consolidation operations begins in May 1987. Resulting consolidated fuel will be utilized in the VP Cooperative Agreement SFSC Program

  6. Wavelength-selective thermal emitters using Si-rods on MgO

    Science.gov (United States)

    Suemitsu, Masahiro; Asano, Takashi; De Zoysa, Menaka; Noda, Susumu

    2018-01-01

    Supporting substrates for Si rod-type photonic crystals (PCs) are investigated for realizing highly wavelength-selective near-infrared thermal emitters. Three materials—SiO2, Al2O3, and MgO—are considered for their low infrared emission (transparency) and remarkable heat resistance. Theoretical calculations of the emissivity spectra of Si-rod PCs (rod height = 500 nm, rod diameter = 300 nm, and lattice constant = 600 nm) on 50 μm-thick supporting substrates at 1400 K indicate that the long-wavelength (>3 μm) emission power from the emitter using MgO is less than 1/10 of that of the other two materials. Fabrication of the Si-rod PCs on the 50 μm-thick MgO substrate requires the insertion of a thin (30 nm) HfO2 film between MgO and Si to improve the stability at high temperatures (>1400 K). Experimental results of the fabricated structure show that at 1400 K, the ratio of emissive power at wavelengths <1.8 μm to the total emissive power is 34% and that this can be increased to over 53% in an optimized rod-array structure with a 10 μm-thick MgO substrate.

  7. Hydrodynamic prediction of multidimensional single- and two-phase flow in rod arrays. Progress report, January 1-December 31, 1983

    International Nuclear Information System (INIS)

    Ebeling-Koning, D.B.; Robinson, J.T.; Todreas, N.E.

    1984-01-01

    The objective of this research is to develop comprehensive constitutive models for multidimensional two-phase flow in rod arrays. The constitutive parameters are the solid-fluid flow resistance and the gas-liquid interfacial momentum exchange force. This report covers work in four areas: (1) a correlation for flow resistance across banks of tubes which is independent of rod arrangement has been developed. The correlation was developed from data from three rod arrangements covering a Reynolds number range (based on superficial velocity) of 1 to 40,000; (2) complete pressure drop data for water flows in the laminar region in crossflow and 45 0 inclined rod arrays were taken; (3) the development of a model for the interfacial momentum exchange force in bubbly flows has been completed. This model has been validated against single bubble velocity data in inclined rod arrays. The model has been cast in a form suitable for implementation to two-fluid computer codes; and (4) rise velocities of bubbles in 0 0 , 45 0 , and 90 0 inclined rod arrays have been measured. This data should prove useful for the development of a bubble drag coefficient model for rod arrays

  8. Development of a suitable weld geometry for pressure resistance welding of the leader test assembly (LTA's) 16NGF fuel assembly fuel rod at Angra-1 Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Junqueira, Fabio da Silva; Silva, Josue Ribeiro, E-mail: fabiojunqueira@inb.gov.br, E-mail: josueribeiro@inb.gov.br [Industrias Nucleares do Brasil (GEPRDN/INB), Rio de Janeiro, RJ (Brazil). Gerencia do Produto

    2013-07-01

    The purpose of this work is to develop suitable weld geometry for pressure resistance welding of the zircaloy-4 end plug to the special zirconium alloy cladding tube, Ø 9,14mm, for demonstration at Angra-1 Nuclear Plant. Weld geometry development was carried out in two steps: at the first one, the influence caused by the variation of the welding process key parameters, the axial compression strength of the end plug against the cladding tube, projection of the cladding tube into the welding chamber and the welding current have been evaluated; at the second step, the influence of the variation of end-plug weld geometry area was checked. For the combination of welding parameters, the technique of factorial design was used. Results from mechanical and metallographic tests have indicated a strong and direct influence of weld geometry dimensional variation on the weld mechanical resistance, and a modest influence in relation to the range of key parameters used to carry out tests. (author)

  9. Evaluation of the TIG welding mechanical behavior in AISI 316 tubes for fuel rods

    International Nuclear Information System (INIS)

    Bittencourt, M.S.Q.; Carvalho Perdigao, S. de

    1985-10-01

    The effect of service temperature, the mechanical resistance and the creep behaviour of a steel which is intendend to be used as fuel rods in Nuclear Reactors was investigated. The tests were performed in seamless tubes of austenitic stainless steel, AISI 316, 20% cold worked, TIG welded. (Author) [pt

  10. 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

  11. 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

  12. Rod bundle burnout data and correlation comparisons

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  13. Genetics Home Reference: cone-rod dystrophy

    Science.gov (United States)

    ... common cause of autosomal recessive cone-rod dystrophy , accounting for 30 to 60 percent of cases. At ... Patient Support and Advocacy Resources (4 links) American Foundation for the Blind Foundation Fighting Blindness Retina International ...

  14. Control rod drive of nuclear reactor

    International Nuclear Information System (INIS)

    Zhuchkov, I.I.; Gorjunov, V.S.; Zaitsev, B.I.

    1980-01-01

    This invention relates to nuclear reactors and, more particularly, to a drive of a control rod of a nuclear reactor and allows power control, excess reactivity compensation, and emergency shut-down of a reactor. (author)

  15. Strain Measurement Using Embedded Fiber Bragg Grating Sensors Inside an Anchored Carbon Fiber Polymer Reinforcement Prestressing Rod for Structural Monitoring

    DEFF Research Database (Denmark)

    Kerrouche, Abdelfateh; Boyle, William J.O.; Sun, Tong

    2009-01-01

    Results are reported from a study carried out using a series of Bragg grating based optical fiber sensors written into a very short length (60mm) optical fiber net work and integrated into carbon fiber polymer reinforcement (CFPR) rod. Such rods are used as reinforcements in concrete structures...... from the calibrated force applied by the pulling machine and from a conventional resistive strain gauge mounted on the rod itself is obtained. Calculations from strain to shear stress show a relatively uniform stress distribution along the bar anchor used. The results give confidence to results from...

  16. 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)

  17. 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.)

  18. Crippling Strength of Axially Loaded Rods

    Science.gov (United States)

    Natalis, FR

    1921-01-01

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

  19. Epigenomic landscapes of retinal rods and cones

    Science.gov (United States)

    Mo, Alisa; Luo, Chongyuan; Davis, Fred P; Mukamel, Eran A; Henry, Gilbert L; Nery, Joseph R; Urich, Mark A; Picard, Serge; Lister, Ryan; Eddy, Sean R; Beer, Michael A; Ecker, Joseph R; Nathans, Jeremy

    2016-01-01

    Rod and cone photoreceptors are highly similar in many respects but they have important functional and molecular differences. Here, we investigate genome-wide patterns of DNA methylation and chromatin accessibility in mouse rods and cones and correlate differences in these features with gene expression, histone marks, transcription factor binding, and DNA sequence motifs. Loss of NR2E3 in rods shifts their epigenomes to a more cone-like state. The data further reveal wide differences in DNA methylation between retinal photoreceptors and brain neurons. Surprisingly, we also find a substantial fraction of DNA hypo-methylated regions in adult rods that are not in active chromatin. Many of these regions exhibit hallmarks of regulatory regions that were active earlier in neuronal development, suggesting that these regions could remain undermethylated due to the highly compact chromatin in mature rods. This work defines the epigenomic landscapes of rods and cones, revealing features relevant to photoreceptor development and function. DOI: http://dx.doi.org/10.7554/eLife.11613.001 PMID:26949250

  20. 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.

  1. 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

  2. Performance criteria of control rod absorbers of new VVER reactors and a possibility to increase their lifetime

    International Nuclear Information System (INIS)

    Zakharov, A.V.; Risovany, V.D.; Vasilchenko, I.N.; Kurakin, K.Y.; Kushmanov, S.A.; Makhin, V.M.

    2015-01-01

    At present, the lifetime of control rod absorbers of the existing VVER-100 and constructed VVER-1200 is designed to be ten years; they are operated as control rods for three years and as scram rods for seven years. The examination of spent absorbers did not show any degradation in their performance so their lifetime specification might be too conservative. The criteria of the control rod absorbers performance, together with an adequate computational model, can be a methodological basis to develop a system to manage the VVER-1000 control rod absorbers lifetime to fully use their design and material capacities. The criteria for control rod absorbers performance can be split into 5 groups: 1) criteria for physical efficiency, 2) criteria for dynamic characteristics, 3) criteria for radiation resistance, 4) criteria for thermo-mechanical resistance and 5) criteria for corrosion resistance. There is a need in an analytical model to define the neutron-physical and thermo-hydraulic operating conditions as well as thermo-mechanical state of an absorbing element. At present, the key components of the model have been developed (neutron physics and thermo-hydraulics) and the thermo-mechanical model is under development. The analytical models have been used for the justification of a design of absorbing elements combining dysprosium hafnate pellets at the bottom and boron carbide at the top

  3. Evaluation of the ability of electrical rods to simulate nuclear rod behavior during a loss-of-coolant accident

    International Nuclear Information System (INIS)

    Tolman, E.L.; Gottula, R.C.

    1979-01-01

    The purpose of this report is twofold: (1) to assess the adequacy of electrical rods to simulate nuclear rod behavior during loss-of-coolant-accident (LOCA) conditions, and (2) to identify ongoing tests specifically designed to directly compare electric and nuclear rod response during a LOCA. The capability and limitations of electric rods to simulate LOCA fuel rod response are reviewed. Two tests, the Halden IFA-511 test and the German PNS 4237 test, are being conducted to directly compare electric rod and nuclear rod response under LOCA conditions. The objectives and usefulness of these tests are reviewed

  4. Determination of safety rod reactivity - time function from digitally measured safety rod trajectory

    International Nuclear Information System (INIS)

    Pesic, M.; Milovanovic, S.; Milovanovic, T.

    1996-01-01

    Using rod position - time function z(t), obtained using new method and reactivity - position function rho(z), the complex dependence of the reactivity - time function rho(t) for HERBE safety rods is determined. Some improvements, comparing to the previously proposed method, are obtained and analyzed. (author)

  5. Evaluation and planning for lightning rod grounding of PSTA cyclotron building

    International Nuclear Information System (INIS)

    Suyamto; Taufik; Idrus Abdul Kudus

    2015-01-01

    Lightning rod connected with the ground resistance is an equipment protection against hazards of lightning strikes building. Lightning strike to the building may result in damage to the building and destroy all the equipment inside it. The need for a lightning rod of a building is regulated in PUIPP expressed with risk factors (FR). The amount of FR is the sum of the value of the index of five (5 ) components of the building i.e building functions, construction, the height and the situation of the building and and the number of yearly lightning days in that places. At this time 05 PSTA building has undergone changes in the function of the building's mechanical workshop into a cyclotron building so that safety criteria also change into vital building with lightning rods resistance have to < 1 Ω. From measurements of grounding resistant which exist at present known that average Rp is 1.26 Ω so it is necessary to install new additional grounding resistance to reduce being less than 1 Ω. To fulfil this and taking into consideration the cost and ease of installation, planned addition of a grounding using electrodes solid rods of copper, a diameter of 16 mm and a length of 4 m , planted the soil water depth of 12 m, as well as clay covering, with a water content of about 30 %. Under these conditions and taking into the cost and ease of installation are expected to obtain optimal results i.e. soil resistivity 18.35 Ω-m and its resistance of Rx 4.82 Ω. When coupled with existing grounding final resistant Rp 0.99 Ω obtained is thus fulfilling the requirements of PUIPP that is less than 1 Ω. (author)

  6. 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

  7. Bent Telescopic Rods in Patients With Osteogenesis Imperfecta.

    Science.gov (United States)

    Lee, R Jay; Paloski, Michael D; Sponseller, Paul D; Leet, Arabella I

    2016-09-01

    Telescopic rods require alignment of 2 rods to enable lengthening. A telescopic rod converts functionally into a solid rod if either rod bends, preventing proper engagement. Our goal was to characterize implant bending as a mode of failure of telescopic rods used in the treatment of osteogenesis imperfecta in children. We conducted a retrospective review of our osteogenesis imperfecta database for patients treated with intramedullary telescopic rods at our institution from 1992 through 2010 and identified 12 patients with bent rods. The 6 boys and 6 girls had an average age at the time of initial surgery of 3.1 years (range, 1.8 to 8.3 y) and a total of 51 telescoping rods. Clinic notes, operative reports, and radiographs were reviewed. The rods were analyzed for amount of lengthening, characteristics of bending, presence of cut out, or disengagement from an anchor point. Bends in the rods were characterized by their location on the implant component. The bent and straight rods were compared. Data were analyzed with the Mann-Whitney test (statistical significance set at P≤0.05). Of the 51 telescoping rods, 17 constructs (33%) bent. The average interval between surgery and rod bending was 4.0 years (range, 0.9 to 8.2 y). Before bending, 11 of 17 telescoping rods had routine follow-up radiographs for review. In 10 of the rods, bending was present when early signs of rod failure were first detected. Rod bending did not seem to be related to rod size. There was no area on the rod itself that seemed more susceptible to bending. Rod bending can be an early sign of impending rod failure. When rod bending is first noted, it may predispose the rod to other subsequent failures such as loss of proximal and distal fixation and cut out. Rod bending should be viewed as an indicator for closer monitoring of the patient and discussions regarding future need for rod exchange. Level III-retrospective review.

  8. Color dissociation artifacts in double Maddox rod cyclodeviation testing.

    Science.gov (United States)

    Simons, K; Arnoldi, K; Brown, M H

    1994-12-01

    The double Maddox rod test, based on a red Maddox rod in front of one eye and a clear Maddox rod in front of the other, is used to measure cyclodeviation, typically in patients with superior oblique muscle pareses. Discrepant results between the double Maddox rod test and other torsion measures, and reports of "paradoxic" cyclodeviation in the normal eye of some patients with superior oblique paresis, suggest the two-color format of the double Maddox rod test may produce artifactual torsion measures. Forty patients with superior oblique paresis were tested twice using the double Maddox rod test, reversing the red and white Maddox rods between eyes for the second test, and 18 were tested further with same-color red or clear Maddox rods in front of both eyes. With the standard double Maddox rod test, 33 (83%) of 40 patients localized their cyclodeviation to the eye viewing through the red Maddox rod, irrespective of laterality of the paresis or fixation preference. In all 33 patients, laterality of the perceived torsion changed between eyes when testing was repeated with red and white Maddox rods interchanged between eyes. With same-color Maddox rods before both eyes, 17 (94%) of 18 patients localized extorsion to the paretic eye. There was 7.6:1 ratio of luminance transmission and a 1.6:1 ratio of grating spatial frequency bandpass in the plano meridian between the clear and red Maddox rods, which appear to be responsible for the double Maddox rod test artifact. The traditional double Maddox rod test may produce artifactual cyclodeviation measurements. An alternative version of the test, based on same-color Maddox rods in front of both eyes, is proposed. The relatively high spatial frequency bandpass characteristics of the plano meridian of the Maddox rod (as high as 20/25 Snellen equivalent resolution through the clear Maddox rod) also suggests double Maddox rod testing should be conducted in a dark room to avoid biases from visual environment cues.

  9. 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.)

  10. Broadband Vibration Attenuation Using Hybrid Periodic Rods

    Directory of Open Access Journals (Sweden)

    S. Asiri

    2008-12-01

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

  11. Corrosion performance of optimised and advanced fuel rod cladding in PWRs at high burnups

    International Nuclear Information System (INIS)

    Jourdain, P.; Hallstadius, L.; Pati, S.R.; Smith, G.P.; Garde, A.M.

    1997-01-01

    The corrosion behaviour both in-pile and out-of-pile for a number of cladding alloys developed by ABB to meet the current and future needs for fuel rod cladding with improved corrosion resistance is presented. The cladding materials include: 1) Zircaloy-4 (OPTIN) with optimised composition and processing and Zircaloy-2 optimised for Pressurised Water Reactors (PWR), (Zircaloy-2P), and 2) several alternative zirconium-based alloys with compositions outside the composition range for Zircaloys. The data presented originate from fuel rods irradiated in six PWRs to burnups up to about 66 MWd/kgU and from tests conducted in 360 o water autoclave. Also included are in-pile fuel rod growth measurements on some of the alloys. (UK)

  12. 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

  13. Magnetic switch for reactor control rod. [LMFBR

    Science.gov (United States)

    Germer, J.H.

    1982-09-30

    A magnetic reed switch assembly is described for activating an electromagnetic grapple utilized to hold a control rod in position above a reactor core. In normal operation the magnetic field of a permanent magnet is short-circuited by a magnetic shunt, diverting the magnetic field away from the reed switch. The magnetic shunt is made of a material having a Curie-point at the desired release temperature. Above that temperature the material loses its ferromagnetic properties, and the magnetic path is diverted to the reed switch which closes and short-circuits the control circuit for the control rod electro-magnetic grapple which allows the control rod to drop into the reactor core for controlling the reactivity of the core.

  14. Method of diagnosing the control rod operation

    International Nuclear Information System (INIS)

    Nakaniwa, Tomoko; Kudo, Mitsuru.

    1983-01-01

    Purpose: To confirm the soundness of a control rod operation at high accuracy. Method: By utilizing the fact that a control rod position indicating system outputs digitalized decimal signals, position signals are outputted as below: a stop position signal when a magnet is opposed to an odd number limit switch; a drift position signal when it is opposed to an even number limit switch; and a blank position signal when it is not opposed to a limit switch. Since the standard signal pattern is present for the order of signal change due to the displacement of the magnet i.e., stop position→blank position→drift position, absolute position change and the time for passing the limit switch, the standard pattern is stored in an operation diagnosis unit and compared with the data upon actual driving to diagnose the control rod operation collectively. (Sekiya, K.)

  15. Performance analysis of LMFBR control rods

    International Nuclear Information System (INIS)

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

    1975-01-01

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

  16. Elliptical cross section fuel rod study II

    International Nuclear Information System (INIS)

    Taboada, H.; Marajofsky, A.

    1996-01-01

    In this paper it is continued the behavior analysis and comparison between cylindrical fuel rods of circular and elliptical cross sections. Taking into account the accepted models in the literature, the fission gas swelling and release were studied. An analytical comparison between both kinds of rod reveals a sensible gas release reduction in the elliptical case, a 50% swelling reduction due to intragranular bubble coalescence mechanism and an important swelling increase due to migration bubble mechanism. From the safety operation point of view, for the same linear power, an elliptical cross section rod is favored by lower central temperatures, lower gas release rates, greater gas store in ceramic matrix and lower stored energy rates. (author). 6 refs., 8 figs., 1 tab

  17. Storing device for control rod drive

    International Nuclear Information System (INIS)

    Tomatsu, Tsutomu; Miura, Teruo.

    1989-01-01

    A water supply device for supplying clean water, a recycling pump and a filter disposed between the water supply device and a water vessel by way of recycling pipelines are disposed to a water vessel containing storing water for immerging and storing control rod drives for BWR type reactors upon periodical inspection, etc. Clean water is supplied from the water supply device into the control rod drives immerged in the storing water to remove radioactive cruds, etc. deposited at the surface thereof and water is supplied through the recycling pipelines to the filter to remove solid impurities contained therein and the clean water is returned to the water supply device. Since the clean water is always recycled to the inside of the control rod drives, chemical corrosion and electrical corrosion of nitride parts are prevented and radioactive cruds are processed in separated waste processing systems, the atmospheric radiation doses in the operation chamber is reduced. (S.K.)

  18. 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

  19. Oligo(naphthylene–ethynylene) Molecular Rods

    DEFF Research Database (Denmark)

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

    2013-01-01

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

  20. Method for producing titanium aluminide weld rod

    Science.gov (United States)

    Hansen, Jeffrey S.; Turner, Paul C.; Argetsinger, Edward R.

    1995-01-01

    A process for producing titanium aluminide weld rod comprising: attaching one end of a metal tube to a vacuum line; placing a means between said vacuum line and a junction of the metal tube to prevent powder from entering the vacuum line; inducing a vacuum within the tube; placing a mixture of titanium and aluminum powder in the tube and employing means to impact the powder in the tube to a filled tube; heating the tube in the vacuum at a temperature sufficient to initiate a high-temperature synthesis (SHS) reaction between the titanium and aluminum; and lowering the temperature to ambient temperature to obtain a intermetallic titanium aluminide alloy weld rod.

  1. Digital, electromagnetic rod position indicator with compensation

    International Nuclear Information System (INIS)

    Feilchenfeld, M.M.; Geis, C.G.

    1985-01-01

    A digital rod position indicator having discrete coils L 0 , L 1 , L 2 ..... spaced along the travel path of an elongate magnetically permeable member stores in digital form compensation signals for automatically adjusting the location relative to the coils at which a digital output signal representative of the position of the end of the elongate member transitions from one code to the next. The appropriate compensation signal is addressed using the digital output signal and a correction factor which takes into account the direction of movement including reversals. Reference is made to the positioning of the control rods in a pressurized water reactor. (author)

  2. 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

  3. Rod pomnožných proprií, rod v plurálu, rod obecně

    Czech Academy of Sciences Publication Activity Database

    Šimandl, Josef

    2012-01-01

    Roč. 49, 3/4 (2012), s. 124-128 ISSN 1212-5326. [Rod v jazyce a v jazykovědě. Praha, 08.12.2011] Institutional research plan: CEZ:AV0Z90610518 Keywords : gender of nouns * evidence of gender in plural Subject RIV: AI - Linguistics

  4. Hardened plungers and piston rods for high-pressure compressors

    Energy Technology Data Exchange (ETDEWEB)

    1942-07-07

    This report was a summary of information on dimensions, materials, and operating conditions, as well as experience in the use of piston rods and plungers at Gelsenkirchen. The surface hardening of these parts and their resulting life and wear were of prime importance. Nitriding hardening was one of the best processes for the production of wearproof surfaces. Case hardening and autogeneous hardening had been found satisfactory. Heat hardening had been found to be a cheap process in many applications. Surfaces could be obtained by hard chrome plating which would have the same or higher wear resistance as nitriding and excel in the depth of hardness. However, the heat hardening alone produced hard layers which had sufficiently good properties for plungers and piston rods of the booster compressors, gas-circulation pumps, paste presses, compressors and possibly pressure-release machines. This plant possessed a hardening installation which offered the advantage of production of most of the required equipment right at the works. This was particularly important if a grinding machine was available. This arrangement had to be supplemented with a shaft furnace in which parts could be heated to remove stresses before and after machining. 5 tables.

  5. 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)

  6. Residual stresses in cold drawn ferritic rods

    International Nuclear Information System (INIS)

    Atienza, J.M.; Martinez-Perez, M.L.; Ruiz-Hervias, J.; Mompean, F.; Garcia-Hernandez, M.; Elices, M.

    2005-01-01

    The residual stress state generated by cold-drawing in a ferritic steel rod has been determined. Stress profiles in the three principal directions were measured by neutron and X-ray diffraction and calculated by 3D finite element simulation. The agreement between the simulations and the experimental data is excellent

  7. Piston rod seal for a Stirling engine

    Science.gov (United States)

    Shapiro, Wilbur

    1984-01-01

    In a piston rod seal for a Stirling engine, a hydrostatic bearing and differential pressure regulating valve are utilized to provide for a low pressure differential across a rubbing seal between the hydrogen and oil so as to reduce wear on the seal.

  8. Automatic operation device for control rods

    International Nuclear Information System (INIS)

    Sekimizu, Koichi

    1984-01-01

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

  9. 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 between...

  10. On contact numbers in random rod packings

    NARCIS (Netherlands)

    Wouterse, A.; Luding, Stefan; Philipse, A.P.

    2009-01-01

    Random packings of non-spherical granular particles are simulated by combining mechanical contraction and molecular dynamics, to determine contact numbers as a function of density. Particle shapes are varied from spheres to thin rods. The observed contact numbers (and packing densities) agree well

  11. Heat Transfer Enhancement By Three-Dimensional Surface Roughness Technique In Nuclear Fuel Rod Bundles

    Science.gov (United States)

    Najeeb, Umair

    This thesis experimentally investigates the enhancement of single-phase heat transfer, frictional loss and pressure drop characteristics in a Single Heater Element Loop Tester (SHELT). The heater element simulates a single fuel rod for Pressurized Nuclear reactor. In this experimental investigation, the effect of the outer surface roughness of a simulated nuclear rod bundle was studied. The outer surface of a simulated fuel rod was created with a three-dimensional (Diamond-shaped blocks) surface roughness. The angle of corrugation for each diamond was 45 degrees. The length of each side of a diamond block is 1 mm. The depth of each diamond block was 0.3 mm. The pitch of the pattern was 1.614 mm. The simulated fuel rod had an outside diameter of 9.5 mm and wall thickness of 1.5 mm and was placed in a test-section made of 38.1 mm inner diameter, wall thickness 6.35 mm aluminum pipe. The Simulated fuel rod was made of Nickel 200 and Inconel 625 materials. The fuel rod was connected to 10 KW DC power supply. The Inconel 625 material of the rod with an electrical resistance of 32.3 kO was used to generate heat inside the test-section. The heat energy dissipated from the Inconel tube due to the flow of electrical current flows into the working fluid across the rod at constant heat flux conditions. The DI water was employed as working fluid for this experimental investigation. The temperature and pressure readings for both smooth and rough regions of the fuel rod were recorded and compared later to find enhancement in heat transfer coefficient and increment in the pressure drops. Tests were conducted for Reynold's Numbers ranging from 10e4 to 10e5. Enhancement in heat transfer coefficient at all Re was recorded. The maximum heat transfer co-efficient enhancement recorded was 86% at Re = 4.18e5. It was also observed that the pressure drop and friction factor increased by 14.7% due to the increased surface roughness.

  12. 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

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

    International Nuclear Information System (INIS)

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

    1962-08-01

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

  14. Sucker rod string design of the pumping systems

    Directory of Open Access Journals (Sweden)

    Chun Hua Liu

    2015-05-01

    Full Text Available The existing design of sucker rod string mainly focuses on the simplifying assumptions that rod string was exposed to simple tension loading. And its goal was to have equal modified stress at the top of each taper. The improved rod design was to have the same degree of safety at each section, and it used a dynamic force distribution that was proportional along the whole string. However, the available procedures did not provide the desired accuracy of its pertinent analysis, and the operators could not identify the specific phenomena that occur in CBM wells. In this paper, the mathematical models of rod loads and string length were developed based on the cyclic nature of rod string loading; the fatigue endurance method is used to design the single rod string; and the tapered rod string is designed to have an equal equivalent stress at the top of each section. Its application characteristics are demonstrated by the example of CBM wells in Ordos Basin. The interpretations of results show that the previous design gave the single rods a larger diameter and the top rods in the string a greater percent than the proposed method. The calculation should concern about inertial, vibration and friction forces to illustrate the elastic force waves travelling in the rod material with the speed of sound. The single string should be designed using fatigue endurance ratings due to asymmetric pulsating tension of rod loading; and the tapered string should involve a balanced design by setting the fatigue endurance at each section equal. A shorter stroke length gives a greater rod taper percentage and an increased load capacity results to an enhanced rod diameter. The rod diameter increases with the pump size and load capacity for the single string, and the rod taper percentage of the top rod strings increases with plunger diameter for the tapered string. The proposed research improves efficiency of the pumping system, assures good operating conditions, and reduces

  15. Low fluid level in pulse rod shock absorber

    International Nuclear Information System (INIS)

    Aderhold, H.C.

    1974-01-01

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

  16. Ultrasonics aids the identification of failed fuel rods

    International Nuclear Information System (INIS)

    Anon.

    1985-01-01

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

  17. 77 FR 1504 - Stainless Steel Wire Rod From India

    Science.gov (United States)

    2012-01-10

    ... COMMISSION Stainless Steel Wire Rod From India Determination On the basis of the record \\1\\ developed in the... antidumping duty order on stainless steel wire rod From India would be likely to lead to continuation or... contained in USITC Publication 4300 (January 2012), entitled Stainless Steel Wire Rod From India...

  18. ROD INTERNAL PRESSURE QUANTIFICATION AND DISTRIBUTION ANALYSIS USING FRAPCON

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-01-01

    The discharge rod internal pressure (RIP) and cladding hoop stress (CHS) distributions are quantified forWatts BarNuclearUnit 1 (WBN1) fuel rods by modeling core cycle design data, intercycle assembly movements, operation data (including modeling significant trips and downpowers), and as-built fuel enrichments and densities of each fuel rod in FRAPCON-3.5. An alternate model for the amount of helium released from zirconium diboride (ZrB2) integral fuel burnable absorber (IFBA) layers is derived and applied to FRAPCON output data to quantify the RIP and CHS for these fuel rods. SCALE/Polaris is used to quantify fuel rod-specific spectral quantities and the amount of gaseous fission products produced in the fuel for use in FRAPCON inputs. Fuel rods with ZrB2 IFBA layers (i.e., IFBA rods) are determined to have RIP predictions that are elevated when compared to fuel rod without IFBA layers (i.e., standard rods) despite the fact that IFBA rods often have reduced fill pressures and annular fuel blankets. Cumulative distribution functions (CDFs) are prepared from the distribution of RIP predictions for all standard and IFBA rods. The provided CDFs allow for the determination of the portion of WBN1 fuel rods that exceed a specified RIP limit. Lastly, improvements to the computational methodology of FRAPCON are proposed.

  19. Transient rod failure in a pulsing TRIGA Mark I reactor

    International Nuclear Information System (INIS)

    Draper, E.L. Jr.; Atkinson, G.D. Jr.

    1972-01-01

    Full text: On July 7, 1970 the University of Texas at Austin TRIGA Mark I Pulsing Reactor experienced a failure of the transient control rod. Although no danger to personnel or damage to the reactor other than the pulse rod occurred, the failure was promptly reported to the USAEC regional compliance office. The first indication of an abnormal situation was unusual multiplication behavior during the first start-up of the day. As usual for steady state operation, the operator removed the transient rod and began to withdraw the shim and regulating rods. After partial withdrawal, he noticed that the count rate was not increasing as rapidly as was customary. While remaining at the console,the operator had a technician make a visual inspection of the core. The technician observed the transient drive rod was swinging freely in the pool and the poison section was detached. It was concluded, based on the indications of the.reactor instrumentation and visual inspection, that the transient control rod had broken off and remained in position in the core. The regulating and shim rods were inserted and the transient rod was manually cranked to the down position. The manual manipulation of the transient rod, instead of dropping the rod by gravity, was used so that the connecting rod could be reinserted in the control rod guide tube. The reactor core was then partially unloaded so that a critical mass was not present. The transient rod drive and connecting rod were removed from the pool. The poison section was retrieved from its position in the core by welding a tap to a long rod and tapping into the top of the poison section. Visual inspection of the poison section showed that the weld joining the male threads on the poison section to the main body of the control rod had failed. The threads remained screwed in the control rod drive shaft upon separation and the poison section remained fully inserted in the core. A new control rod was fabricated by Gulf General Atomic and shipped

  20. Rebirth of a control rod at the Phenix power plant

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  1. Modeling of microcrack density based damage evolution in ceramic rods

    International Nuclear Information System (INIS)

    Grove, D.J.; Rajendran, A.M.

    2000-01-01

    This paper presents results from simulations of shock wave propagation in ceramic rods with and without confinement. The experiments involved steel and graded-density flyer plates impacting sleeved and unsleeved AD995 ceramic rods. The main objectives of simulating these experiments were: 1) to validate the Rajendran-Grove (RG) ceramic model constants, and 2) to investigate the effects of confinement on damage evolution in ceramic rods, as predicted by the RG model. While the experimental measurements do not indicate the details of damage evolution in the ceramic rod, the numerical modeling has provided some valuable insight into the damage initiation and propagation processes in ceramic rods

  2. Measurement and analysis of CEFR safety and shim rod worth

    International Nuclear Information System (INIS)

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

    2013-01-01

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

  3. Composites reinforcement by rods a SAS study

    CERN Document Server

    Urban, V; Pyckhout-Hintzen, W; Richter, D; Straube, E

    2002-01-01

    The mechanical properties of composites are governed by size, shape and dispersion degree of so-called reinforcing particles. Polymeric fillers based on thermodynamically driven microphase separation of block copolymers offer the opportunity to study a model system of controlled rod-like filler particles. We chose a triblock copolymer (PBPSPB) and carried out SAS measurements with both X-rays and neutrons, in order to characterize separately the hard phase and the cross-linked PB matrix. The properties of the material depend strongly on the way that stress is carried and transferred between the soft matrix and the hard fibers. The failure of the strain-amplification concept and the change of topological contributions to the free energy and scattering factor have to be addressed. In this respect the composite shows a similarity to a two-network system, i.e. interpenetrating rubber and rod-like filler networks. (orig.)

  4. [Rod of Asclepius. Symbol of medicine].

    Science.gov (United States)

    Young, Pablo; Finn, Bárbara C; Bruetman, Julio E; Cesaro Gelos, Jorge; Trimarchi, Hernán

    2013-09-01

    Symbolism is one of the most archaic forms of human thoughts. Symbol derives from the Latin word symbolum, and the latter from the Greek symbolon or symballo, which means "I coincide, I make matches". The Medicine symbol represents a whole series of historical and ethical values. Asclepius Rod with one serpent entwined, has traditionally been the symbol of scientific medicine. In a misconception that has lasted 500 years, the Caduceus of Hermes, entwined by two serpents and with two wings, has been considered the symbol of Medicine. However, the Caduceus is the current symbol of Commerce. Asclepius Rod and the Caduceus of Hermes represent two professions, Medicine and Commerce that, in ethical practice, should not be mixed. Physicians should be aware of their real emblem, its historical origin and meaning.

  5. Experience with a fuel rod enrichment scanner

    International Nuclear Information System (INIS)

    Kubik, R.N.; Pettus, W.G.

    1975-01-01

    This enrichment scanner views all fuel rods produced at B and W's Commercial Nuclear Fuel Plant. The scanner design is derived from the PAPAS System reported by R. A. Forster, H. D. Menlove, and their associates at Los Alamos. The spatial resolution of the system and smoothing of the data are discussed in detail. The cost-effectiveness of multi-detector versus single detector scanners of this general design is also discussed

  6. Synthesis of disk-on-rod antennas

    Science.gov (United States)

    Dubrovka, F. F.; Lenivenko, V. A.

    1993-05-01

    The analysis and synthesis of disk-on-rod antennas (DORAs) with canonical and stepwise disk shapes are considered. A comparison of theoretical and experimental results shows that mathematical models and software developed by solving the appropriate boundary value problems can be used for the design of optimal DORAs. A broadband centimeter-wave DORA is considered as an example of the application of the proposed method for the constructive synthesis of DORAs using multicriterial optimization.

  7. Multiphoton response of retinal rod photoreceptors

    Directory of Open Access Journals (Sweden)

    Vasilios Alexiades

    2007-02-01

    Full Text Available Phototransduction is the process by which light is converted into an electrical response in retinal photoreceptors. Rod photoreceptors contain a stack of (about 1000 disc membranes packed with photopigment rhodopsin molecules, which absorb the photons. We present computational experiments which show the profound effect on the response of the distances (how many discs apart photons happen to be absorbed at. This photon-distribution effect alone can account for much of the observed variability in response.

  8. Rod Driven Frequency Entrainment and Resonance Phenomena

    Directory of Open Access Journals (Sweden)

    Christina Salchow

    2016-08-01

    Full Text Available A controversy exists on photic driving in the human visual cortex evoked by intermittent photic stimulation. Frequency entrainment and resonance phenomena are reported for frequencies higher than 12 Hz in some studies while missing in others. We hypothesized that this might be due to different experimental conditions, since both high and low intensity light stimulation were used. However, most studies do not report radiometric measurements, which makes it impossible to categorize the stimulation according to photopic, mesopic, and scotopic vision. Low intensity light stimulation might lead to scotopic vision, where rod perception dominates. In this study, we investigated photic driving for rod-dominated visual input under scotopic conditions. Twelve healthy volunteers were stimulated with low intensity light flashes at 20 stimulation frequencies, leading to rod activation only. The frequencies were multiples of the individual alpha frequency (α of each volunteer in the range from 0.40–2.30*α. 306-channel whole head magnetoencephalography recordings were analyzed in time, frequency, and spatiotemporal domains with the Topographic Matching Pursuit algorithm. We found resonance phenomena and frequency entrainment for stimulations at or close to the individual alpha frequency (0.90–1.10*α and half of the alpha frequency (0.40–0.55*α. No signs of resonance and frequency entrainment phenomena were revealed around 2.00*α. Instead, on-responses at the beginning and off-responses at the end of each stimulation train were observed for the first time in a photic driving experiment at frequencies of 1.30–2.30*α, indicating that the flicker fusion threshold was reached. All results, the resonance and entrainment as well as the fusion effects, provide evidence for rod-dominated photic driving in the visual cortex.

  9. Hydrodynamic behavior of a bare rod bundle

    International Nuclear Information System (INIS)

    Bartzis, J.G.; Todreas, N.E.

    1977-06-01

    The temperature distribution within the rod bundle of a nuclear reactor is of major importance in nuclear reactor design. However temperature information presupposes knowledge of the hydrodynamic behavior of the coolant which is the most difficult part of the problem due to complexity of the turbulence phenomena. In the present work a 2-equation turbulence model--a strong candidate for analyzing actual three dimensional turbulent flows--has been used to predict fully developed flow of infinite bare rod bundle of various aspect ratios (P/D). The model has been modified to take into account anisotropic effects of eddy viscosity. Secondary flow calculations have been also performed although the model seems to be too rough to predict the secondary flow correctly. Heat transfer calculations have been performed to confirm the importance of anisotropic viscosity in temperature predictions. All numerical calculations for flow and heat have been performed by two computer codes based on the TEACH code. Experimental measurements of the distribution of axial velocity, turbulent axial velocity, turbulent kinetic energy and radial Reynolds stresses were performed in the developing and fully developed regions. A 2-channel Laser Doppler Anemometer working on the Reference mode with forward scattering was used to perform the measurements in a simulated interior subchannel of a triangular rod array with P/D = 1.124. Comparisons between the analytical results and the results of this experiment as well as other experimental data in rod bundle array available in literature are presented. The predictions are in good agreement with the results for the high Reynolds numbers

  10. 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.

  11. Strength and Stiffness Analysis by the Finite-Difference Method of a Concrete Floor Slab Reinforced with Composite Rods During a Fire

    Science.gov (United States)

    Shirko, A. V.; Kamlyuk, A. N.; Drobysh, A. S.; Spiglazov, A. V.

    2017-05-01

    A strength and stiffness comparative analysis has been made of a concrete slab reinforced with composite-reinforcement rods and a slab reinforced with steel rods. The stress-strain state has been assessed for both versions of reinforcement of the slab. The stress-strain state was determined under the action of only static load and with subsequent application of temperature fields, i.e., under standard-fire conditions. It has been shown that the fire resistance of the slab with a composite reinforcement turns out to be 1.6 higher as far as the bearing capacity is concerned, than the fire resistance of the slab with a steel reinforcement, although the initial deflection due to the action of only static load for the slab reinforced with composite rods exceeds six to seven times the deflection of the slab reinforced with steel rods.

  12. Kinetics of folding of the myosin rod.

    Science.gov (United States)

    Bechet, J J; Nozais, M

    1997-02-15

    The kinetics of the unfolding and refolding of the myosin rod have been studied by fluorescence and circular dichroism techniques, at different concentrations of protein and guanidine hydrochloride. The unfolding of the myosin rod was fast and at least biphasic in 2-3 M denaturant, with an initial immediate phase followed by a slower low-amplitude first-order phase. The refolding of the rod in 0.4-2 M guanidine hydrochloride was also at least biphasic; an initial immediate phase preceded a slow second-order phase. At the final denaturant concentration of 0.8 M, the amplitude of the burst phase was weakly dependent on the protein concentration and the rate constant of the refolding slow phase was optimal. These data are incorporated into a folding mechanism with at least three states. The high rates of the first steps of unfolding and refolding may be relevant for the functioning of the native myosin molecule by allowing a transient separation of the two strands of the myosin tail.

  13. 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

  14. Incorporation of squalene into rod outer segments

    International Nuclear Information System (INIS)

    Keller, R.K.; Fliesler, S.J.

    1990-01-01

    We have reported previously that squalene is the major radiolabeled nonsaponifiable lipid product derived from [ 3 H]acetate in short term incubations of frog retinas. In the present study, we demonstrate that newly synthesized squalene is incorporated into rod outer segments under similar in vitro conditions. We show further that squalene is an endogenous constituent of frog rod outer segment membranes; its concentration is approximately 9.5 nmol/mumol of phospholipid or about 9% of the level of cholesterol. Pulse-chase experiments with radiolabeled precursors revealed no metabolism of outer segment squalene to sterols in up to 20 h of chase. Taken together with our previous absolute rate studies, these results suggest that most, if not all, of the squalene synthesized by the frog retina is transported to rod outer segments. Synthesis of protein is not required for squalene transport since puromycin had no effect on squalene incorporation into outer segments. Conversely, inhibition of isoprenoid synthesis with mevinolin had no effect on the incorporation of opsin into the outer segment. These latter results support the conclusion that the de novo synthesis and subsequent intracellular trafficking of opsin and isoprenoid lipids destined for the outer segment occur via independent mechanisms

  15. ELECTROMAGNETIC APPARATUS FOR MOVING A ROD

    Science.gov (United States)

    Young, J.N.

    1957-08-20

    An electromagnetic device for moving an object in a linear path by increments is described. The device is specifically adapted for moving a neutron absorbing control rod into and out of the core of a reactor and consists essentially of an extension member made of magnetic material connected to one end of the control rod and mechanically flexible to grip the walls of a sleeve member when flexed, a magnetic sleeve member coaxial with and slidable between limit stops along the flexible extension, electromagnetic coils substantially centrally located with respect to the flexible extension to flex the extension member into gripping engagement with the sleeve member when ener gized, moving electromagnets at each end of the sleeve to attract the sleeve when energized, and a second gripping electromagnet positioned along the flexible extension at a distance from the previously mentioned electromagnets for gripping the extension member when energized. In use, the second gripping electromagnet is deenergized, the first gripping electromagnet is energized to fix the extension member in the sleeve, and one of the moving electromagnets is energized to attract the sleeve member toward it, thereby moving the control rod.

  16. Analytical prediction of turbulent friction factor for a rod bundle

    International Nuclear Information System (INIS)

    Bae, Jun Ho; Park, Joo Hwan

    2011-01-01

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

  17. Description and characterization of HBWR Series H-1 test rods

    International Nuclear Information System (INIS)

    Wagoner, S.R.; Barner, J.O.; Welty, R.K.

    1979-06-01

    The as-built characterization results are presented for the HBWR Series H-1 test rods to be irradiated as part of the Fuel Performance Improvement Program (FPIP). The irradiation of these rods is to be conducted in the Halden Boiling Water Reactor (HBWR). Series H-1 consists of twelve rods for irradiation and six spares. Rod design types include (1) a reference dished pellet design, (2) an annular pellet design, (3) an annular pellet design combined with graphite-coated cladding, and (4) a packed-particle (vipac) design. The report, which describes the fabrication and detailed characterization results for the rods, is divided into four major sections: (1) experiment description, (2) process development required to fabricate the test rods, (3) methods and procedures used to fabricate and characterize the rods, and (4) a summary of the characterization results

  18. A test research on ventilative well-distributivity under normal temperature for a control rod drive mechanism (Continuous article)

    International Nuclear Information System (INIS)

    Zhu Longxing

    1989-01-01

    A test for cooling of the control rod drive mechnism under normal temperature is described. The relationship between the unbalanced cofficient and the frictional resistance and wind velocity is found by comparing the ventilation in plate top structure of reactor with that in global top structure of reactor

  19. Between a Map and a Data Rod

    Science.gov (United States)

    Teng, William; Rui, Hualan; Strub, Richard; Vollmer, Bruce

    2015-01-01

    A Digital Divide has long stood between how NASA and other satellite-derived data are typically archived (time-step arrays or maps) and how hydrology and other point-time series oriented communities prefer to access those data. In essence, the desired method of data access is orthogonal to the way the data are archived. Our approach to bridging the Divide is part of a larger NASA-supported data rods project to enhance access to and use of NASA and other data by the Consortium of Universities for the Advancement of Hydrologic Science, Inc. (CUAHSI) Hydrologic Information System (HIS) and the larger hydrology community. Our main objective was to determine a way to reorganize data that is optimal for these communities. Two related objectives were to optimally reorganize data in a way that (1) is operational and fits in and leverages the existing Goddard Earth Sciences Data and Information Services Center (GES DISC) operational environment and (2) addresses the scaling up of data sets available as time series from those archived at the GES DISC to potentially include those from other Earth Observing System Data and Information System (EOSDIS) data archives. Through several prototype efforts and lessons learned, we arrived at a non-database solution that satisfied our objectivesconstraints. We describe, in this presentation, how we implemented the operational production of pre-generated data rods and, considering the tradeoffs between length of time series (or number of time steps), resources needed, and performance, how we implemented the operational production of on-the-fly (virtual) data rods. For the virtual data rods, we leveraged a number of existing resources, including the NASA Giovanni Cache and NetCDF Operators (NCO) and used data cubes processed in parallel. Our current benchmark performance for virtual generation of data rods is about a years worth of time series for hourly data (9,000 time steps) in 90 seconds. Our approach is a specific implementation of

  20. ''Fabrice'' process for reconstituting experimental rods in a hot cell from pre-irradiated rods in power reactors

    International Nuclear Information System (INIS)

    Houdaille, B.; Vignesoult, N.; Atabek, R.

    1981-11-01

    The Fabrice process for the ''hot cell refabrication'' of small rods from very long irradiated fuel elements was developed by the CEA, particularly to enable parametric studies to be carried out on the behaviour under irradiation of rods from power stations. This technique, although requiring intricate operations in a hot cell with specially adapted equipment and a very experienced personnel, enables (1) a significant saving to be made compared with the pre-irradiation of new rods and (2) affords a virtually unlimited choice of rod sections irradiated in a power reactor for parametric irradiations. Tests have shown that the initial characteristics of the sample rod are not modified and that the behaviour under irradiation of whole non-refabricated rods and Fabrice rods with the same initial characteristics is identical, for a local specific burn-up under 40,000 MWd/t, vis-a-vis the fuel cladding interaction phenomenon [fr

  1. 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)

  2. 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.)

  3. 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

  4. Sucker-rod pumping handbook production engineering fundamentals and long-stroke rod pumping

    CERN Document Server

    Takacs, Gabor

    2015-01-01

    Sucker-Rod Pumping Handbook presents the latest information on the most common form of production enhancement in today's oil industry, making up roughly two-thirds of the producing oilwell operations in the world. The book begins with an introduction to the main features of sucker rod pumping and an explanation and comparison of lift methods. It goes on to provide the technical and practical knowledge needed to introduce the new and practicing production engineer and operator to the equipment, technology, and applications required to maintain optimum operating conditions.

  5. 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

  6. Minimization of PWR reactor control rods wear

    International Nuclear Information System (INIS)

    Ponzoni Filho, Pedro; Moura Angelkorte, Gunther de

    1995-01-01

    The Rod Cluster Control Assemblies (RCCA's) of Pressurized Water Reactors (PWR's) have experienced a continuously wall cladding wear when Reactor Coolant Pumps (RCP's) are running. Fretting wear is a result of vibrational contact between RCCA rodlets and the guide cards which provide lateral support for the rodlets when RCCA's are withdrawn from the core. A procedure is developed to minimize the rodlets wear, by the shuffling and axial reposition of RCCA's every operating cycle. These shuffling and repositions are based on measurement of the rodlet cladding thickness of all RCCA's. (author). 3 refs, 2 figs, 2 tabs

  7. 4-rod RFQ linac for ion implantation

    Energy Technology Data Exchange (ETDEWEB)

    Fujisawa, Hiroshi; Hamamoto, Nariaki; Inouchi, Yutaka [Nisshin Electric Co. Ltd., Kyoto (Japan)

    1997-03-01

    A 34 MHz 4-rod RFQ linac system has been upgraded in both its rf power efficiency and beam intensity. The linac is able to accelerate in cw operation 0.83 mA of a B{sup +} ion beam from 0.03 to 0.91 MeV with transmission of 61 %. The rf power fed to the RFQ is 29 kW. The unloaded Q-value of the RFQ has been improved approximately 61 % to 5400 by copper-plating stainless steel cooling pipes in the RFQ cavity. (author)

  8. Maintenance of BWR control rod drive mechanisms

    International Nuclear Information System (INIS)

    Greene, R.H.

    1991-01-01

    Control rod drive mechanism (CRDM) replacement and rebuilding is one of the highest dose, most physically demanding, and complicated maintenance activities routinely accomplished by BWR utilities. A recent industry workshop sponsored by the Oak Ridge National Laboratory, which dealt with the effects of CRDM aging, revealed enhancements in maintenance techniques and tooling which have reduced ALARA, improved worker comfort and productivity, and have provided revised guidelines for CRDM changeout selection. Highlights of this workshop and ongoing research on CRDM aging are presented in this paper

  9. Protector predominantly for pump sucker rods

    Energy Technology Data Exchange (ETDEWEB)

    Razhetdinov, U.Z.; Prokopov, O.I.; Sharafutdinov, I.G.; Valishim, Yu.G.

    1982-01-01

    A protector is proposed which includes a cylindrical housing with connecting threaded sections on the ends and rim with spheres attached to the outer surface of the housing. In order to improve reliable operation of the protector by reducing wear of the rocking supports, the rim on the outer surface of the housing is installed at an angle to its axis with the possibility of movement of the sphere in the rim around the sucker rods with interaction of them with the pump-compressor pipes.

  10. Performance predictions and manufacturing concerns of burnable poison rods

    International Nuclear Information System (INIS)

    Copeland, R.A.; Buescher, B.J.

    1977-01-01

    Burnable poison rods for reactors designed by B and W consist of low density pellets, composed of boron carbide dispersed in an alumina matrix (Al 2 O 3 --B 4 C), which are contained in Zircaloy-4 tubes. To predict reliable operation of these rods, the irradiation behavior of the components must be known. Performance models were developed based on experimental irradiation data. During rod fabrication, care must be taken to limit the amount of hydrogen in the rod because of the propensity of Zircaloy to hydride in the presence of high levels of hydrogen. Furthermore, the hygroscopic nature of alumina dictates that care must be taken to avoid moisture (a primary source of hydrogen) in the rods. Manufacturing and quality testing procedures have been developed to provide conformance to the design criteria. Examinations have been performed on irradiated burnable poison rods which verify the adequacy of both performance models and manufacturing procedures

  11. Rapid and accurate control rod calibration measurement and analysis

    International Nuclear Information System (INIS)

    Nelson, George W.; Doane, Harry J.

    1990-01-01

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

  12. Study on evaluating the reactivity worth of the control rods of the PWR 900 MWe

    International Nuclear Information System (INIS)

    Phan Quoc Vuong; Tran Vinh Thanh; Tran Viet Phu

    2015-01-01

    Control rods of a nuclear reactor are divided into two groups: shut down and power control. Reactivity worth of the control rods depends nonlinearly on the rods' compositions and positions where the rods are inserted into the core. Therefore, calculation of control rod worth is of high important. In this study, we calculated the reactivity worth of the power control rod bank of the Mitsubishi PWR 900 MWe. The results are integral and differential worth calibration of the control rods. (author)

  13. Comparison between temperature distributions of an annular fuel rod of circular cross-section and of a hemoglobin shaped cross-section rod for PWR reactors in steady state conditions

    International Nuclear Information System (INIS)

    Oliveira, Maria Vitória A. de; Alvim, Antônio Carlos Marques

    2017-01-01

    The objective of this work is to make a comparison between the temperature distributions of an annular fuel rod of circular cross-section and a hemoglobin shaped cross-section for PWR reactors in steady state conditions. The motivation for this article is due to the fact that the symmetric form of the red globules particles allows the O 2 gases to penetrate the center of the cell homogeneously and quickly. The diffusion equation of gases in any environment is very similar to the heat diffusion equation: Diffusion - Fick's Law; Heat Flow - Fourier; where, the temperature (T) replaces the concentration (c). In previous works the comparison between the shape of solid fuel rods with circular section, and a with hemoglobin-shaped cross-section has proved that this new format optimizes the heat transfer, decreasing the thermal resistance between the center of the UO 2 pellets and the clad. With this, a significant increase in the specific power of the reactor was made possible (more precisely a 23% increase). Currently, the advantages of annular fuel rods are being studied and recent works have shown that 12 x 12 arrays of annular fuel rods perform better, increasing the specific power of the reactor by at least 20% in relation to solid fuel rods, without affecting the safety of the reactor. Our proposal is analyzing the temperature distribution in annular fuel rods with cross sections with red blood cell shape and compare with the theoretical results of the annular fuel rods of circular cross section, initially in steady state. (author)

  14. Comparison between temperature distributions of an annular fuel rod of circular cross-section and of a hemoglobin shaped cross-section rod for PWR reactors in steady state conditions

    Energy Technology Data Exchange (ETDEWEB)

    Oliveira, Maria Vitória A. de; Alvim, Antônio Carlos Marques, E-mail: moliveira@con.ufrj.br, E-mail: alvim@nuclear.ufrj.br [Coordenacao de Pos-Graduacao e Pesquisa de Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear

    2017-07-01

    The objective of this work is to make a comparison between the temperature distributions of an annular fuel rod of circular cross-section and a hemoglobin shaped cross-section for PWR reactors in steady state conditions. The motivation for this article is due to the fact that the symmetric form of the red globules particles allows the O{sub 2} gases to penetrate the center of the cell homogeneously and quickly. The diffusion equation of gases in any environment is very similar to the heat diffusion equation: Diffusion - Fick's Law; Heat Flow - Fourier; where, the temperature (T) replaces the concentration (c). In previous works the comparison between the shape of solid fuel rods with circular section, and a with hemoglobin-shaped cross-section has proved that this new format optimizes the heat transfer, decreasing the thermal resistance between the center of the UO{sub 2} pellets and the clad. With this, a significant increase in the specific power of the reactor was made possible (more precisely a 23% increase). Currently, the advantages of annular fuel rods are being studied and recent works have shown that 12 x 12 arrays of annular fuel rods perform better, increasing the specific power of the reactor by at least 20% in relation to solid fuel rods, without affecting the safety of the reactor. Our proposal is analyzing the temperature distribution in annular fuel rods with cross sections with red blood cell shape and compare with the theoretical results of the annular fuel rods of circular cross section, initially in steady state. (author)

  15. Design of the pulse rod drive mechanism for pulsed reactor

    International Nuclear Information System (INIS)

    You Keyi

    1988-07-01

    The pulse rod drive mechanism is a critical movable device for a pulsed reactor. It is an executor under pulse operations, and it may be used in a shim rod under steady-state operations. The pneumatic-electromechanical driving method is taken in the designing. The structure, operating, calculation of parameters and designing methods of the pulse rod drive mechanism are briefly described in this paper. The testing results of the prototypical mechanism are also presented

  16. Nuclear reactor internals construction and failed fuel rod detection system

    International Nuclear Information System (INIS)

    Frisch, E.; Andrews, H.N.

    1976-01-01

    A system is provided for determining during operation of a nuclear reactor having fluid pressure operated control rod mechanisms the exact location of a fuel assembly with a defective fuel rod. The construction of the reactor internals is simplified in a manner to facilitate the testing for defective fuel rods and the reduce the cost of producing the upper internals of the reactor. 13 claims, 10 drawing figures

  17. 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

  18. 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)

  19. TileCal ROD Hardware and Software Requirements

    CERN Document Server

    Castelo, J; Cuenca, C; Ferrer, A; Fullana, E; Higón, E; Iglesias, C; Munar, A; Poveda, J; Ruiz-Martínez, A; Salvachúa, B; Solans, C; Valls, J A

    2005-01-01

    In this paper we present the specific hardware and firmware requirements and modifications to operate the Liquid Argon Calorimeter (LiArg) ROD motherboard in the Hadronic Tile Calorimeter (TileCal) environment. Although the use of the board is similar for both calorimeters there are still some differences in the operation of the front-end associated to both detectors which make the use of the same board incompatible. We review the evolution of the design of the ROD from the early prototype stages (ROD based on commercial and Demonstrator boards) to the production phases (ROD final board based on the LiArg design), with emphasis on the different operation modes for the TileCal detector. We start with a short review of the TileCal ROD system functionality and then we detail the different ROD hardware requirements for options, the baseline (ROD Demo board) and the final (ROD final high density board). We also summarize the performance parameters of the ROD motherboard based on the final high density option and s...

  20. Fast, accurate control rod calibration using a programmable desk calculator

    International Nuclear Information System (INIS)

    Naugle, N.W.; Randall, John D.

    1972-01-01

    In an attempt to develop a simple least squares program for the rapid calibration of control rods it was necessary to verify that all rods, with the exception of the transient rod, could be accurately defined by a single analytical expression. Since the vertical flux distribution in the core region follows a cosine function, a cosine squared variation was tested. The solution which involves the inversion of a 3 x 3 matrix is performed using a Hewlett Packard Model 9100B programmable desk calculator. The least squares program was applied to a number of control rod calibrations that had previously been analyzed by hand. The agreement was excellent in all cases

  1. 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

  2. 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

  3. Correlation of normal and superconducting transport properties on textured Bi-2212 ceramic thin rods

    International Nuclear Information System (INIS)

    Natividad, E.; Castro, M.; Burriel, R.; Diez, J.C.; Navarro, R.; Angurel, L.A.

    2002-01-01

    The electric and thermal properties well above and below T c of Bi-2212 textured ceramics have been correlated through a careful analysis of the microstructure and the transport measurements. Thin rods with the same Bi-2122 stoichiometry and textured by a laser floating zone technique have been studied with that aim. By changing the growth parameters, it has been possible to produce strong changes in microstructure and critical current density, J c , with small variations in the thermal conductivity. The existence of phase and composition gradients across the thin rods, which explains the variations of T c , makes the relation difficult between the normal state resistivity and J c (77 K). A simple qualitative analysis that takes into account the observed microstructure has been developed to correlate the electric transport properties in the normal and in the superconducting states. (author)

  4. Fatigue behaviour of window and rods

    International Nuclear Information System (INIS)

    Bergeron, J.; Brochard, J.; Cheron, Ch.; Gabriel, F.

    1999-01-01

    The current CEA project pertains to feasibility studies of an internal source of neutrons used for irradiations of various natures. An external source generates the necessary protons thanks to a particle accelerator that delivers a high energy proton beam: 600 MeV, 40 mA. The target is composed of rod assemblies and the spallation material is lead contained in aluminium cladding cooled by low pressured tepid water. The interface between the accelerator and the target named 'window' must both ensure tightness between the accelerator and the target and maintain a differential pressure while being as thin as possible to avoid a too great dissipation of the incident beam. In this respect, the interface is made of Inconel of low thickness in order to be as transparent as possible of the proton beam whose average power density is about of 10 μA.cm -2 , and is cooled by forced convection water of the target. An analysis of nominal and incidental situations of the facility operating mode has been conducted, especially in order to evaluate the consequences of abrupt and frequent shutdown or tripping of the accelerator on the thermomechanical behaviour of the spallation rods and the window, as well as in terms of thermal fatigue. (author)

  5. 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

  6. 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

  7. Rolls-Royce digital Rod Control System

    International Nuclear Information System (INIS)

    Pouillot, M.

    2010-01-01

    Full text of publication follows: Rolls-Royce has developed a new generation of Rod Control System, based on 40 years of experience. The fifth-generation Rod Control System (RCS) from Rolls-Royce offers a reliable, modular design with adaptability to your preferred platform, for modernization projects or new reactors. Flexible implementation provides the option for you to keep existing cabinets, which permits you to optimize installation approach. Main features for the power part: - Control Rod Drive Mechanism (CRDM) type: 3-coil. - Independent control of each sub-bank. - Each sub-bank is controlled by a cycler unit and 3 identical power racks, each including 4 identical power modules and a common power-supply module. - Coil-per-coil digital control: each power module embeds power-conversion, current-control, and current-monitoring functions for one coil. Control and monitoring are carried out by separate electronics in the module. Current is digitized and fully monitored by means of min-max templates. - A double-hold function is included: a power module assigned to a gripper will activate its coil if a fault risking to cause a reactor trip occurs. - Power modules are standardized, hot-pluggable and self-configured: a power module includes a set of parameters for each type of coil SG, MG, LC. The module recognizes the rack it is plugged in, and chooses automatically parameters to be used. Main benefits: - Reduced operational, maintenance, training, and inventory costs: standardization of power modules and integration of control and monitoring on the same PC-card lead to a drastic reduction of spare part types, and simplification of the system. - Easy maintenance: - Replacement of a power module solves nearly all failures due to current control or monitoring for a coil. It is done instantly thanks to hot-plug capability. - On the front plate of power-modules, LEDs provide useful information for diagnostic: current setpoint from cycler, output current bar

  8. Effect of thermal cycling on the strength and superconducting properties of laser floating zone textured Bi-2212 rods

    International Nuclear Information System (INIS)

    Salazar, A.; Pastor, J.Y.; Llorca, J.; Natividad, E.; Gimeno, F.J.; Angurel, L.A.

    2003-01-01

    The influence of the thermal cycling on the flexure strength and superconducting properties of textured Bi 2 Sr 2 CaCu 2 O 8+δ rods processed by the laser-heated floating zone method was studied. Two different temperature ranges (77-300 and 77-393 K) were used to separate the contributions of chemical reactions with liquid water from those due to the thermo-elastic stresses. The rods showed excellent resistance to thermal cycling in both cases and neither the superconducting nor the mechanical properties were significantly affected. In particular, the variation of the critical current at 77 K was below ±4% after 50 thermal cycles. The flexure strength of rods increased or decreased by ∼10% after a few tens of thermal cycles, and crack deflection along the rod axis was observed in the specimens which showed a reduction in strength. This decrease was transitory, and the flexure strength after 50 thermal cycles was similar to that measured on the as-received rods

  9. 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

  10. 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

  11. [Development and Evaluation of a New Selective Culture Medium, KBM Anaero RS-GNR, for Detection of Anaerobic Gram Negative Rods].

    Science.gov (United States)

    Narita, Taeko; Kato, Kyohei; Hanaiwa, Hiroki; Harada, Tetsuhiro; Funashima, Yumiko; Akiwa, Makoto; Sekiguchi, Jun-Ichiro; Nagasawa, Zenzo; Umemura, Tsukuru

    2017-03-22

    The laboratory culture methods for isolating drug-resistant pathogens has been the gold standard in medical microbiology, and play pivotal roles in the overall management of infectious diseases. Recently, several reports have emphasized the development of antibiotics-resistance among anaerobic gram-negative rods, especially Genus Bacteroides and Prevotella . Therefore, a selective culture method to detect these pathogens is needed. We developed here the new selective culture medium, termed "KBM Anaero RS-GNR," for detecting anaerobic Gram-negative rods. Growth capability and selectivity of the agar medium were assessed by using the pure culture suspensions of more than 100 bacterial strains as well as the 13 samples experimentally contaminated with these bacterial strains. This new medium, "KBM Anaero RS-GNR," successfully showed the selective isolation of anaerobic Gram-negative rods. Compared with commercially available medium, "PV Brucella HK Agar, " which is also designed to detect anaerobic Gram-negative rods, there was no significant difference of the overall detection efficiency between two media. However, "KBM Anaero RS-GNR" showed superior to selectivity for anaerobic Gram-negative rods, especially from the samples contaminated with Candida species. Thus, the culture method using KBM Anaero RS-GNR is relevant for isolation of anaerobic Gram-negative rods especially from clinical specimens.

  12. Use of Supplemental Short Pre-Contoured Accessory Rods and Cobalt Chrome Alloy Posterior Rods Reduces Primary Rod Strain and Range of Motion Across the Pedicle Subtraction Osteotomy Level

    DEFF Research Database (Denmark)

    Hallager, Dennis Winge; Gehrchen, Poul Martin; Dahl, Benny

    2016-01-01

    STUDY DESIGN: In vitro cadaveric biomechanical study. OBJECTIVE: To assess effects of 4-rod reconstruction, rod material, and anterior column support on motion and surface rod strain in a pedicle subtraction osteotomy model. SUMMARY OF BACKGROUND DATA: Pedicle subtraction osteotomy (PSO) can...... correct significant sagittal deformity of the lumbar spine; however, revision rates are high. To reduce rod strain and the incidence of rod fracture, clinical use of multi-rod construction, cobalt chrome (CoCr) alloy rods, and interbody spacers adjacent to PSO has been proposed. Investigating both motion...

  13. Electroless silver coating of rod-like glass particles.

    Science.gov (United States)

    Moon, Jee Hyun; Kim, Kyung Hwan; Choi, Hyung Wook; Lee, Sang Wha; Park, Sang Joon

    2008-09-01

    An electroless silver coating of rod-like glass particles was performed and silver glass composite powders were prepared to impart electrical conductivity to these non-conducting glass particles. The low density Ag-coated glass particles may be utilized for manufacturing conducting inorganic materials for electromagnetic interference (EMI) shielding applications and the techniques for controlling the uniform thickness of silver coating can be employed in preparation of biosensor materials. For the surface pretreatment, Sn sensitization was performed and the coating powders were characterized by scanning electron microscopy (SEM), focused ion beam microscopy (FIB), and atomic force microscopy (AFM) along with the surface resistant measurements. In particular, the use of FIB technique for determining directly the Ag-coating thickness was very effective on obtaining the optimum conditions for coating. The surface sensitization and initial silver loading for electroless silver coating could be found and the uniform and smooth silver-coated layer with thickness of 46 nm was prepared at 2 mol/l of Sn and 20% silver loading.

  14. 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.)

  15. 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.

  16. 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.

  17. 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

  18. Simple measuring rod method for the coaxiality of serial holes

    Science.gov (United States)

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

    2017-11-01

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

  19. Numerical investigation of flow past a row of rectangular rods

    Directory of Open Access Journals (Sweden)

    S.Ul. Islam

    2016-09-01

    Full Text Available A numerical study of uniform flow past a row of rectangular rods with aspect ratio defined as R = width/height = 0.5 is performed using the Lattice Boltzmann method. For this study the Reynolds number (Re is fixed at 150, while spacings between the rods (g are taken in the range from 1 to 6. Depending on g, the flow is classified into four patterns: flip-flopping, nearly unsteady-inphase, modulated inphase-antiphase non-synchronized and synchronized. Sudden jumps in physical parameters were observed, attaining either maximum or minimum values, with the change in flow patterns. The mean drag coefficient (Cdmean of middle rod is higher than the second and fourth rod for flip-flopping pattern while in case of nearly unsteady-inphase the middle rod attains minimum drag coefficient. It is also found that the Strouhal number (St of first, second and fifth rod decreases as g increases while that of other two have mixed trend. The results further show that there exist secondary interaction frequencies together with primary vortex shedding frequency due to jet in the gap between rods for 1 ⩽ g ⩽ 3. For the average values of Cdmean and St, an empirical relation is also given as a function of gap spacing. This relation shows that the average values of Cdmean and St approach to those of single rectangular rod with increment in g.

  20. 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.

  1. Evaluation of fuel rods behavior - under irradiation test

    International Nuclear Information System (INIS)

    Lameiras, F.S.; Terra, J.L.; Pinto, L.C.M.; Dias, M.S.; Pinheiro, R.B.

    1981-04-01

    By the accompanying of the irradiation of instrumented test fuel rods simulating the operational conditions in reactors, plus the results of post - irradiation exams, tests, evaluation and calibration of analitic modelling of such fuel rods is done. (E.G.) [pt

  2. Thermal-recovery of modal instability in rod fiber amplifiers

    DEFF Research Database (Denmark)

    Jørgensen, Mette Marie; Laurila, Marko; Noordegraaf, Danny

    2013-01-01

    We investigate the temporal dynamics of Modal instabilities (MI) in ROD fiber amplifiers using a 100 μm core rod fiber in a single-pass amplifier configuration, and we achieve ~200W of extracted output power before the onset of MI. Above the MI threshold, we investigate the temporal dynamics...

  3. The development and validation of control rod calculation methods

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  4. Optimal control rod programs in power reactors

    International Nuclear Information System (INIS)

    Fadilah, S.M.; Lewins, J.

    1975-01-01

    Control rod programming is investigated with respect to optimising the power peaking factor and hence the utilisation of a nuclear reactor. A simplified diffusion model, initially with a finite number of regions, in cylindrical geometry, is used to enable optimal trajectories to be completely synthesised. The average discharge burnup problem is posed both as an external and as an internal optimisation. The connection between optimum power shape and the maximisation of the average discharge burnup is explored in a wider context. It is shown that optimum trajectories combine an initial singular solution of the Haling type with a terminal bang-bang solution. An extension to a higher number of regions and, on passing to the limit, to a diffusion model, provides an alternative proof of Haling's principle without the restriction to monotonic reactivity decrease with burnup. Numerical results in the two-region model are given to show the general scope of optimisation available. (author)

  5. Welding nuclear reactor fuel rod end plugs

    International Nuclear Information System (INIS)

    Yeo, D.

    1984-01-01

    Apparatus for applying a vacuum to a nuclear fuel rod cladding tube's interior through its open end while girth welding an inserted end plug to its other end. An airtight housing has an orifice with a seal which can hermetically engage the tube's open end. A vacuum hose has one end connected to the housing and the other end connected to a vacuum pump. A mechanized device moves the housing to engage or disengage its seal with the tube's open end. Preferably the mechanized device includes an arm having one end attached to the housing and the other end pivotally attached to a moveable table; an arm rotating device to coaxially align the housing's orifice with the welding-positioned tube; and a table moving device to engage the seal of the coaxially aligned orifice with the tube's open end

  6. Welding nuclear reactor fuel rod end plugs

    International Nuclear Information System (INIS)

    Yeo, D.

    1984-01-01

    Apparatus for applying a vacuum to a nuclear fuel rod cladding tube's interior through its open end while girth welding an inserted end plug to its other end. An airtight housing has an orifice with a seal which can hermetically engage the tube's open end. A vacuum hose has one end connected to the housing and the other end connected to a vacuum pump. A mechanized device which moves the housing to engage or disengage its seal with the tube's open end includes at least one arm having one end attached to the housing and the other end pivotally attached to a movable table; an arm rotating device to coaxially align the housing's orifice with the welding-positioned tube; and a table moving device to engage the seal of the coaxially aligned orifice with the tube's open end. (author)

  7. Control rod homogenization in heterogeneous sodium-cooled fast reactors

    International Nuclear Information System (INIS)

    Andersson, Mikael

    2016-01-01

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

  8. Damage development in rod-on-rod impact test on 1100 pure aluminum

    Science.gov (United States)

    Iannitti, G.; Bonora, N.; Bourne, N.; Ruggiero, A.; Testa, G.

    2017-01-01

    Stress triaxiality plays a major role in the nucleation and growth of ductile damage in metals and alloys. Although, the mechanisms responsible for ductile failure are the same at low and high strain rate, in impact dynamics, in addition to time resolved stress triaxiality and plastic strain accumulation, pressure also contributes to establish the condition for ductile failure to occur. In this work, ductile damage development in 1100 commercially pure aluminum was investigated by means of rod-on-rod (ROR) impact tests. Based on numerical simulations, using a continuum damage mechanics (CDM) model that accounts for the role of pressure on damage parameters and stochastic variability of such parameters, the impact velocity for no damage, incipient and fully developed damage were estimated. ROR tests at selected velocities were performed and damage distribution and extent were investigated by sectioning of soft recovered samples. Comparison between numerical simulations and experimental results is presented and discussed.

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

    International Nuclear Information System (INIS)

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

    1984-01-01

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

  10. Stress Analysis of Fuel Rod under Axial Coolant Flow

    International Nuclear Information System (INIS)

    Jin, Hai Lan; Lee, Young Shin; Lee, Hyun Seung; Park, Num Kyu; Jeon, Kyung Rok

    2010-01-01

    A pressurized water reactor(PWR) fuel assembly, is a typical bundle structure, which uses light water as a coolant in most commercial nuclear power plants. Fuel rods that have a very slender and long clad are supported by fuel assembly which consists of several spacer grids. A coolant is a fluid which flows through device to prevent its overheating, transferring the heat produced by the device to other devices that use or dissipate it. But at the same time, the coolant flow will bring out the fluid induced vibration(FIV) of fuel rods and even damaged the fuel rod. This study has been conducted to investigate the flow characteristics and nuclear reactor fuel rod stress under effect of coolant. Fluid structure interaction(FSI) analysis on nuclear reactor fuel rod was performed. Fluid analysis of the coolant which flow along the axial direction and structural analysis under effect of flow velocity were carried out under different output flow velocity conditions

  11. Interfacial colloidal rod dynamics: Coefficients, simulations, and analysis

    Science.gov (United States)

    Yang, Yuguang; Bevan, Michael A.

    2017-08-01

    Colloidal rod diffusion near a wall is modeled and simulated based on a constrained Stokesian dynamic model of chains-of-spheres. By modeling colloidal rods as chains-of-spheres, complete diffusion tensors are computed for colloidal rods in bulk media and near interfaces, including hydrodynamic interactions, translation-rotation coupling, and all diffusion modes in the particle and lab frames. Simulated trajectories based on the chain-of-spheres diffusion tensor are quantified in terms of typical experimental quantities such as mean squared positional and angular displacements as well as autocorrelation functions. Theoretical expressions are reported to predict measured average diffusivities as well as the crossover from short-time anisotropic translational diffusion along the rod's major axis to isotropic diffusion. Diffusion modes are quantified in terms of closed form empirical fits to model results to aid their use in interpretation and prediction of experiments involving colloidal rod diffusion in interfacial and confined systems.

  12. A review of control rod calibration methods for irradiated AGRs

    International Nuclear Information System (INIS)

    Telford, A.R.R.

    1975-10-01

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

  13. Substitute safety rods: Physics design and NTG calibration

    International Nuclear Information System (INIS)

    Baumann, N.P.

    1991-07-01

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

  14. TileCal ROD Motherboard Software Library User's Manual

    CERN Document Server

    Salvachúa, B; Castillo, C; Cuenca, C; Ferrer, A; Fullana, E; Higón, E; Iglesias, C; Munar, A; Poveda, J; Ruiz-Martinez, A; Solans, C; Valls, J

    2005-01-01

    This note describes the software library and an associated standalone application program to handle the TileCal ROD VME motherboard. The library uses the CMT packages vme_rcc and rcc_error, from the ATLAS Online Data Flow to handle the standard crate controller, VP-110 from Concurrent Technologies, and the custom bit3_rcc CMT package to handle an alternative crate controller, the BIT-3 from SBSTM Technologies. The ROD library defines several C++ classes which can be used in either standalone applications to control and debug the RODs or with the TDAQ online software integration of the back-end hardware for the TileCal detector. The library also includes special auxiliary classes to handle additional back-end boards related to the ROD operation like the TBM or the ROD injectors.

  15. Buckling of a thin rod under cylindrical constraint

    Science.gov (United States)

    Miller, Jay; Su, Tianxiang; Wicks, Nathan; Pabon, Jahir; Bertoldi, Katia; Reis, Pedro

    2013-03-01

    We investigate the buckling and post-buckling behavior of a thin elastic rod, under cylindrical constraint, with distributed loading. Our precision model experiments consist of injecting a custom-fabricated rod into a transparent glass pipe. Under imposed velocity (leading to frictional axial loading), a portion of the initially straight rod first buckles into a sinusoidal mode and eventually undergoes a secondary instability into a helical configuration. The buckling and post-buckling behavior is found to be highly dependent on the system's geometry, namely the injected rod length and the aspect ratio of the rod to pipe diameter, as well as material parameters. We quantify the critical loads for this sequence of instabilities, contrast our results with numerical experiments and rationalize the observed behavior through scaling arguments.

  16. Device for replacing the rods of a fuel element of a nuclear reactor

    International Nuclear Information System (INIS)

    Nissel, B.; Kybranz, R.; Will, R.

    1977-01-01

    In order to be able to replace several separate rods (fuel rods or absorber rods), in a fuel element, a special grab is introduced, which consists of several individual gripping devices and is operated by spring loading. (TK) [de

  17. Optimization of fuel rod enrichment distribution to minimize rod power peaking throughout life within BWR fuel assembly

    International Nuclear Information System (INIS)

    Hirano, Yasushi; Hida, Kazuki; Sakurada, Koichi; Yamamoto, Munenari

    1997-01-01

    A practical method was developed for determining the optimum fuel enrichment distribution within a boiling water reactor fuel assembly. The method deals with two different optimization problems, i.e. a combinatorial optimization problem grouping fuel rods into a given number of rod groups with the same enrichment, and a problem determining an optimal enrichment for each fuel rod under the resultant rod-grouping pattern. In solving these problems, the primary goal is to minimize a predefined objective function over a given exposure period. The objective function used here is defined by a linear combination: C 1 X+C 2 X G , where X and X G stand for a control variable to give the constraint respectively for a local power peaking factor and a gadolinium rod power, and C 1 and C 2 are user-definable weighting factor to accommodate the design preference. The algorithm of solving the combinatorial optimization problem starts with finding the optimal enrichment vector without any rod-grouping, and promising candidates of rod-grouping patterns are found by exhaustive enumeration based on the resulting fuel enrichment ordering, and then the latter problem is solved by using the method of approximation programming. The practical application of the present method is shown for a contemporary 8x8 Pu mixed-oxide fuel assembly with 10 gadolinium-poisoned rods. (author)

  18. Study on the effect of fuel rod vibration characteristics on the grid-to-rod fretting wear

    International Nuclear Information System (INIS)

    Kyu Tae Kim

    1997-01-01

    The fretting wear-induced fuel rod failure may be caused by excessive flow-induced vibration and/or inadequate fuel rod support by spacer grid springs. In order to evaluate the fuel rod support conditions, the GRIDFORCE program has been developed. This program takes into account cladding creep rate, initial spring deflection, initial spring force, and spring force relaxation rate as the key fuel design parameters affecting the in-reactor fuel rod supporting conditions. On the other hand, relationship of fuel rod supporting conditions and flow-induced vibration characteristics has been derived, based on fretting wear damage patterns observed in some PWRs. Comparison of the predicted data of the GRIDFORCE program and the fretting wear damage patterns indicates that the GRIDFORCE program can be utilized as an effective tool in evaluating the fretting wear damage

  19. An alternative pathway for signal flow from rod photoreceptors to ganglion cells in mammalian retina.

    OpenAIRE

    DeVries, S H; Baylor, D A

    1995-01-01

    Rod signals in the mammalian retina are thought to reach ganglion cells over the circuit rod-->rod depolarizing bipolar cell-->AII amacrine cell-->cone bipolar cells-->ganglion cells. A possible alternative pathway involves gap junctions linking the rods and cones, the circuit being rod-->cone-->cone bipolar cells-->ganglion cells. It is not clear whether this second pathway indeed relays rod signals to ganglion cells. We studied signal flow in the isolated rabbit retina with a multielectrode...

  20. 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

  1. Measurement Report for the Four-Rod LHC Crab Cavity. Cold Tests held in July 2014

    CERN Document Server

    Navarro Tapia, Maria; Calaga, Rama; Hernandez Chahin, Karim Gibran; Junginger, Tobias; Macpherson, Alick; Torres-Sanchez, Roberto; CERN. Geneva. ATS Department

    2015-01-01

    The performance of the four-rod cavity prototype considered for the HL-LHC upgrade has already been assessed at CERN at cryogenic temperatures three times in the last two years [1, 2, 3]. In this report, the results of the latest measurements, carried out in July 2014, are shown. These measurements were to check the improvement of the cavity performance due to the change of the input and pick-up antennas. An estimation of the residual resistance of the Niobium was also performed.

  2. Operating Stresses in Aircraft-Engine Crankshafts and Connecting Rods. 2 - Instrumentation and Test Results

    Science.gov (United States)

    1945-08-01

    cylinders h and 7 were motored with exhaust and intake valves installed but with valve push rods removed. The power was absorbed by an electric ...four strain gages were interconnected to form an electrical resist- ance bridge of the "&eatstone type (fig. 1(h)) with Advance wire gages in two...20° B.T.C.’ NACA C 11844 7-24-45 -^ !» ’. /•< > t , ,f •c**v7jvt iff- r ,yyt « • ^ff^Wf/fr^ v Ignition Moto ri ng 1500 rpm

  3. Impact loading of a BWR control rod during braking

    International Nuclear Information System (INIS)

    Heeschen, U.

    1977-01-01

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

  4. Evaluation of rod insertion issue for NPP Krsko

    International Nuclear Information System (INIS)

    Gunstek, A.; Kurincic, B.

    1998-01-01

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

  5. 118-C-4 Horizontal Rod Cave characterization plan

    International Nuclear Information System (INIS)

    1997-08-01

    This characterization plan provides instructions for obtaining and analyzing samples for waste designation and disposal. The 118-C-4 Horizontal Rod Cave is located in the 100-C Area about 328 ft (100 m) southeast of the 105-C Reactor (Figure 1). The 118-C-4 Horizontal Rod Cave (Figure 2) is a reinforced concrete bunker approximately 70- ft (21.3-m) long, 7-ft (2.1-m) high, and 12-ft (3.6-m) wide, with triangular-shaped concrete ends 3-ft (0.9-m) high. The rod cave was used to store radiologically contaminated control-rod tips. If control rod tips are present, release of control rod activation products will not change expectations with respect to principal contaminants. The north portion of the cave is empty and the south portion contains two aluminum tubes that may contain rod tips (Figure 3). The caves are contaminated with activation and fission products (e.g., 60Co and 137Cs) common to the 100 Areas (see Appendix for data). Dose rates up to 0.7 mR/hr were measured in the south cave and 0.5 mR/hr in the north cave during an inspection of the facility in December 1996

  6. Anisotropy in CdSe quantum rods

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-01-01

    The size-dependent optical and electronic properties of semiconductor nanocrystals have drawn much attention in the past decade, and have been very well understood for spherical ones. The advent of the synthetic methods to make rod-like CdSe nanocrystals with wurtzite structure has offered us a new opportunity to study their properties as functions of their shape. This dissertation includes three main parts: synthesis of CdSe nanorods with tightly controlled widths and lengths, their optical and dielectric properties, and their large-scale assembly, all of which are either directly or indirectly caused by the uniaxial crystallographic structure of wurtzite CdSe. The hexagonal wurtzite structure is believed to be the primary reason for the growth of CdSe nanorods. It represents itself in the kinetic stabilization of the rod-like particles over the spherical ones in the presence of phosphonic acids. By varying the composition of the surfactant mixture used for synthesis we have achieved tight control of the widths and lengths of the nanorods. The synthesis of monodisperse CdSe nanorods enables us to systematically study their size-dependent properties. For example, room temperature single particle fluorescence spectroscopy has shown that nanorods emit linearly polarized photoluminescence. Theoretical calculations have shown that it is due to the crossing between the two highest occupied electronic levels with increasing aspect ratio. We also measured the permanent electric dipole moment of the nanorods with transient electric birefringence technique. Experimental results on nanorods with different sizes show that the dipole moment is linear to the particle volume, indicating that it originates from the non-centrosymmetric hexagonal lattice. The elongation of the nanocrystals also results in the anisotropic inter-particle interaction. One of the consequences is the formation of liquid crystalline phases when the nanorods are dispersed in solvent to a high enough

  7. 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

  8. Gamma-ray spectroscopy on irradiated fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Terremoto, Luis Antonio Albiac [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil). Centro de Engenharia Nuclear], e-mail: laaterre@ipen.br

    2009-07-01

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

  9. Axial gas flow in irradiated PWR fuel rods

    International Nuclear Information System (INIS)

    Dagbjartsson, S.J.; Murdock, B.A.; Owen, D.E.; MacDonald, P.E.

    1977-09-01

    Transient and steady state axial gas flow experiments were performed on six irradiated, commercial pressurized water reactor fuel rods at ambient temperature and 533 K. Laminar flow equations, as used in the FRAP-T2 and SSYST fuel behavior codes, were used with the gas flow results to calculate effective fuel rod radial gaps. The results of these analyses were compared with measured gap sizes obtained from metallographic examination of one fuel rod. Using measured gap sizes as input, the SSYST code was used to calculate pressure drops and mass fluxes and the results were compared with the experimental gas flow data

  10. Gamma-ray spectroscopy on irradiated fuel rods

    International Nuclear Information System (INIS)

    Terremoto, Luis Antonio Albiac

    2009-01-01

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

  11. System for fuel rod removal from a reactor module

    International Nuclear Information System (INIS)

    Matchett, R.L.; Roof, O.R.; Kikta, T.J.; Wilczynski, R.; Nilsen, R.J.; Bacvinskas, W.S.; Fodor, G.

    1990-01-01

    This patent describes a robotic system for remote underwater withdrawal of the fuel rods from fuel modules of a light water breeder reactor includes a collet/grapple assembly for gripping and removing fuel rods in each module, which is positioned by use of a winch and a radial support means attached to a vertical support tube which is mounted over the fuel module. A programmable logic controller in conjunction with a microcomputer, provides control for the accurate positioning and pulling force of the rod grapple assembly. Closed circuit television cameras are provided which aid in operator interface with the robotic system

  12. Behavior of water reactor fuel rod

    International Nuclear Information System (INIS)

    Yanagisawa, Kazuaki

    1990-08-01

    This paper reviewed the fuels used widely in forms of (1) Zircaloy-sheathed UO 2 fuel in light water-commercial power reactor, (2) Zircaloy-sheathed PuO 2 -UO 2 fuel in plutonium-thermal reactor and advanced reactor (ATR), (3) aluminide and silicide fuel in Material Testing Reactors. From fundamental view points, physical/chemical properties and irradiation behaviors of both fuels and zircaloy claddings are briefly reviewed in chapters 1 and 2. Change of the fuel rod physical parameters with progress of burn-up are summed up in chapter 3. Some fuel troubles and failures encountered in past usage of worldwide LWR fuels are introduced with counterplans taken. In the last session of this chapter, recent results of R and D works have been carried out by fuel vendors are reviewed. Especially, in-core behaviors of PCI-remedy fuels developed to use for high burn-up extension and for load-follow operation are highlighted. Reactor accidents occurred through past forty years are surveyed and reviewed. Fuel behaviors during the reactivity initiated accident (RIA), the power-coolant mismatch (PCM), and the loss-of-coolant accident (LOCA) are taken into this review by using disclosed literatures. Safety criteria being used in Japanese licensing authorities are introduced relating to the fuel design limit. (author)

  13. Hydraulic Rod Drives for the CAREM Reactor

    International Nuclear Information System (INIS)

    Mazzi, R.O

    2000-01-01

    CAREM belongs to those considered innovative reactors and their main design goal is obtain a significant improvement in safety.Requirements for the design of the first shutdown systems (FSS) is one of the mayor challenges from functional and reliability point of view, among most of the system of a nuclear reactor.Thus, the design of First Shutdown System must be in accordance with both, the system and the specific design criteria of the CAREM concept.In order to choose the best option for the control rod drive device, three different alternatives have been analysed in the frame of the Project.This paper discusses the advantages and disadvantages of each option and presents the main reasons to select the hydraulic type as the most promising one.The principles and main characteristics of the selected system are explained and the main goals to be obtained during development activities, in order to obtain a reliable design to successfully comply with operating requirements for reactor service are also presented

  14. Nuclear reactor shutdown control rod assembly

    International Nuclear Information System (INIS)

    Bilibin, K.

    1988-01-01

    This patent describes a nuclear reactor having a reactor core and a reactor coolant flowing therethrough, a temperature responsive, self-actuated nuclear reactor shutdown control rod assembly, comprising: an upper drive line terminating at its lower end with a substantially cylindrical wall member having inner and outer surfaces; a lower drive line having a lower end adapted to be attached to a neutron absorber; a ring movable disposed about the outer surface of the wall member of the upper drive line; thermal actuation means adapted to be in heat exchange relationship with coolant in an associated reactor core and in contact with the ring, and balls located within the openings in the upper drive line. When reactor coolant approaches a predetermined design temperature the actuation means moves the ring sufficiently so that the balls move radially out from the recess and into the space formed by the second portion of the ring thereby removing the vertical support for the lower drive line such that the lower drive line moves downwardly and inserts an associated neutron absorber into an associated reactor core resulting in automatic reduction of reactor power

  15. Control rod driving hydraulic pressure device

    International Nuclear Information System (INIS)

    Ogawa, Masahide.

    1993-01-01

    The present invention concerns a control rod driving hydraulic device of a BWR type reactor, and provides an improvement for a means for supplying mechanical seal flashing water of a pump. That is, a mechanical seal flashing pipeline is branched at the downstream of a pressure-reducing orifice and connected to a minimum flow pipeline. With such a constitution, the minimum flow pipeline is connected to a minimum flow pipeline of an auxiliary pump at the downstream of the pressure-reducing orifice and returned to a suction pipeline of the pump. Pressure at the downstream of the pressure-reducing orifice is set, in the orifice, to a pressure required for mechanical seal flashing. Accordingly, the mechanical seal flashing pipeline is connected and a part of minimum flow rate is utilized, thereby enabling to cool mechanical seals. As a result, flow rate of the mechanical flashing water which has been flown out can be saved. The exhaustion amount from the pump can be reduced, to decrease the shaft power and reduce the capacity of the motor. (I.S.)

  16. Electromotor control rod drive for nuclear reactors

    International Nuclear Information System (INIS)

    Baker, S.M.

    1975-01-01

    The positioning of a control rod arranged in a pressure vessel takes place with a drive. This protrudes out of the pressure vessel through a support and is formed from a rotating field motor with energy source, e.g. alternating current connection. Its stator surrounds a section of a pressure casing which covers the length of the drive. The rotor is arranged in the pressure casing and interacts with a shaft lying in the rotation axis. Furthermore, segments are hinged on it, each of which forms two arms of a rocker. Each segment can be revolved against a storing force in a plane containing the rotation axis, through the stator field acting on one of the rocker arms. In order that the drive motor is automatically blocked should the electricity supply fail, the other rocker arm can be connected with a fixed cased component of the drive having the effect of a friction break or a form-locking mechanical catch. (DG/LH) [de

  17. Fluorescent colloidal silica rods - synthesis and phase behavior

    Science.gov (United States)

    Kuijk, A.

    2012-01-01

    Although the experimental study of spherical colloids has been extensive, similar studies on rod-like particles are rare because suitable model systems are scarce. To fulfill this need, we present the synthesis of monodisperse rod-like silica colloids with tunable dimensions. Rods were produced with diameters of 200 nm and larger and lengths up to 10 µm, which resulted in aspect ratios ranging from 1 to 25. The growth mechanism of these rods involves emulsion droplets of water in pentanol, inside which silica condensation takes place. Since the silica nucleus is attached to the water/pentanol interface, the supply of reactants to the nucleus is anisotropic, causing it to grow on one side only, which results in rod formation. The rods were made suitable for quantitative real-space studies by confocal laser scanning microscopy. Several methods of fluorescent labeling are presented that resulted in constant fluorescence levels, gradients from one rod-end to the other, and even patterns of two colors. Single particle imaging was achieved by creating core-shell rods that had a fluorescent core and a non-fluorescent shell. Alternatively, the rods could be dispersed in a solvent with a low dielectric constant to induce micron-sized double layers. To enable quantitative measurements, a tracking algorithm was developed that identifies the rods' positions and orientations. The newly developed model system was used to study the phase behavior of rods. By combining real-space confocal laser scanning microscopy and small angle X-ray scattering methods, a phase diagram depending on concentration and aspect ratio was constructed, which shows good qualitative agreement with simulation results in literature. This phase diagram includes nematic and smectic phases for the higher aspect ratios. Also, the effect of external fields (electric fields, shear and templates) on the phase behavior was studied. In an electric field, rods aligned themselves with the applied field due to an

  18. Results of Post Irradiation Examinations of VVER Leaky Rods

    Energy Technology Data Exchange (ETDEWEB)

    Markov, D.; Perepelkin, S.; Polenok, V.; Zhitelev, V.; Mayorshina, G. [Head of Fuel Research Department, JSC ' SSC RIAR' , 433510, Dimitrovgrad-10, Ulyanovsk region (Russian Federation)

    2009-06-15

    The most important requirement imposed on fuel elements is to maintain integrity of fuel rod claddings under operation, storage and transportation, since it is directly related to the operational safety. However, failed rod claddings are sometimes observed under reactor operation. Identification and unloading of fuel assemblies with leaky rods from VVER is available only at the time of planned preventive maintenance. An unscheduled reactor shutdown due to the excess of coolant activity limit as well as a preterm unloading of the fuel assembly cause economic damage to nuclear plant. Therefore, models and calculation codes were developed to forecast coolant contamination and failed fuel rod behavior. Criteria based on calculations were set to determine the admissible number of the failed rods in core and the opportunity to continue the reactor operation or pre-term unloading of the fuel assembly with the failed rods. Nevertheless, to prevent the fuel rod failure (for unfailing operation) it is necessary to reveal disadvantages of the design, fabrication method and fuel operation conditions, and to eliminate defects. The most complete and significant information about spent fuel assemblies may be received following the post irradiation material examinations. In order to reveal failure origins and mechanism of changes in VVER fuel and failed rod cladding condition depending on the operation, the examinations of 12 VVER-1000 fuel assemblies and 3 VVER-440 fuel assemblies, operated under normal conditions up to the fuel burnup 13..47 MWd/kgU were carried out. To evaluate the rod cladding condition, reveal defects and determine their parameters, the ultrasonic control of cladding integrity, surface visual inspection, eddy current defectoscopy, measurement of geometrical parameters were applied. In separate cases we used the metallography, measured the hydrogen percentage and carried out the mechanical tests of o-ring samples. The pellet condition was evaluated in

  19. Properties of fiber reinforced plastic rods for prestressing tendons of concrete. 8. ; Dynamic fatigue behavior of reinforced plastics rods. Prestressed concrete yo FRP kinchozai no tokusei. 8. ; FRP rod no doteki hiro tokusei

    Energy Technology Data Exchange (ETDEWEB)

    Uomoto, T.; Nishimura, T. (The University of Tokyo, Tokyo (Japan). Institute of Industrial Science)

    1994-01-01

    This paper describes the result of experiments on dynamic fatigue and creep characteristics of plastic rods reinforced by aramid, glass and carbon fibers (AFRP rods, GFRP rods, and CFRP rods) used as prestressing tendons of concrete. Although the dynamic fatigue limit for generally used FRP rods is not known, the fatigue strength was found to decrease in the order of AFRP rod > CFRP rod > GFRP rod. However, AFRP rods have shown a sharp decrease in the fatigue strength starting from around one million dynamic loading iterations. The AFRP and GFRP rods have had creep fracture existed in fatigue fracture at a high stress of 100 kg/mm[sup 2] or higher, and fatigue fracture existed due to dynamic fatigue below that stress. Residual tensile strength after fatigue due to two-million iterative loadings was about 83% of the average tensile strength for the AFRP rods, and about 69% for the GFRP rods, whereas that for CFRP rods was about the same as the average strength. 8 refs., 6 figs., 1 tab.

  20. 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

  1. Physics analysis of the gang partial rod drive event

    International Nuclear Information System (INIS)

    Boman, C.; Frost, R.L.

    1992-08-01

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

  2. Solitary waves in a magneto-electro-elastic circular rod

    International Nuclear Information System (INIS)

    Xue, C X; Pan, E; Zhang, S Y

    2011-01-01

    A simple nonlinear model is proposed in this paper to study the solitary wave in a circular magneto-electro-elastic rod. Based on the constitutive relation for transversely isotropic piezoelectric and piezomagnetic materials, combined with the differential equations of motion, we derive the longitudinal wave motion equation in a long circular rod. The nonlinearity considered is geometrically associated with the nonlinear normal strain in the longitudinal rod direction and the transverse Poisson's effect is included by introducing the effective Poisson's ratio. The nonlinear solitary wave equation is solved by the Jacobi elliptic function expansion method and numerical examples demonstrate not only the existence of such a wave but also some interesting characteristics of the solitary wave in the rod made of different multiphase coupled materials

  3. Computation of reactor control rod drop time under accident conditions

    International Nuclear Information System (INIS)

    Dou Yikang; Yao Weida; Yang Renan; Jiang Nanyan

    1998-01-01

    The computational method of reactor control rod drop time under accident conditions lies mainly in establishing forced vibration equations for the components under action of outside forces on control rod driven line and motion equation for the control rod moving in vertical direction. The above two kinds of equations are connected by considering the impact effects between control rod and its outside components. Finite difference method is adopted to make discretization of the vibration equations and Wilson-θ method is applied to deal with the time history problem. The non-linearity caused by impact is iteratively treated with modified Newton method. Some experimental results are used to validate the validity and reliability of the computational method. Theoretical and experimental testing problems show that the computer program based on the computational method is applicable and reliable. The program can act as an effective tool of design by analysis and safety analysis for the relevant components

  4. Device for detecting defective nuclear reactor fuel rods

    International Nuclear Information System (INIS)

    Steven, J.

    1976-01-01

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

  5. 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.

  6. Absorber rod bundle actuator in a pressurized water nuclear reactor

    International Nuclear Information System (INIS)

    Martin, J.; Peletan, R.

    1984-01-01

    The invention concerns an absorber rod bundle actuator in a pressurized water reactor with spectral shift control. The device comprises two coaxial control bars. The inner bar is integral with the absorber rod bundle; it has an enlarged zone which acts as a proton under pressure difference across an annular seal which can be radially expanded, the pressure difference allowing to the absorber rod bundles actuating on the piston. When a pressure difference is applied, the seal expands radially by a sufficient amount to make sealing contact with the zone of larger diameter in the outer bar. The invention applies more particularly to reactors with spectral shift control using bundles of fertile rods [fr

  7. The ATLAS LARG ROD G-Links Cooling System

    CERN Document Server

    Hubaut, F; Repain, P; Rossel, F; Vincent, D

    2004-01-01

    In this note is described the water cooling system that will be implemented on the ROD boards of the liquid argon calorimeter detectors in order to guarantee a proper behavior of the optical reception of the data.

  8. Post irradiation examination of control rod assembly of FBTR

    International Nuclear Information System (INIS)

    Anandaraj, V.; Raghu, N.; Venkiteswaran, C.N.; Visweswaran, P.; Vijayakumar, Ran; Jayaraj, V.V.; Padmaprabu, P.; Saravanan, T.; Philip, John; Muralidharan, N.G.; Joseph, Jojo; Kasiviswanathan, K.V.

    2010-01-01

    Six control rods with boron carbide pellets are used in FBTR for shutdown and control of reactor power. One control rod after being subjected to a fluence level of 7.2 x 10 22 n/cm 2 was received for post irradiation examination (PIE) to assess its irradiation behavior and to investigate the incident of dropping of control rod. Examinations carried out include precise dimensional measurements to investigate the possibility of interference between the control rod and outer sheath, Neutron radiography and x-radiograph to assess the integrity of the boron carbide pellets and other internals, density measurements to assess the swelling behaviour of boron carbide pellets and metallographic examinations to study the cracking behaviour and microstructural changes in the pellet and the clad. Depletion of B 10 in the pellet was studied using time of flight mass spectrometry. The paper highlights the examinations and results of the PIE carried out. (author)

  9. Vibrations of post-buckled rods: The singular inextensible limit

    KAUST Repository

    Neukirch, Sébastien

    2012-01-01

    The small-amplitude in-plane vibrations of an elastic rod clamped at both extremities are studied. The rod is modeled as an extensible, shearable, planar Kirchhoff elastic rod under large displacements and rotations, and the vibration frequencies are computed both analytically and numerically as a function of the loading. Of particular interest is the variation of mode frequencies as the load is increased through the buckling threshold. While for some modes there are no qualitative changes in the mode frequencies, other frequencies experience rapid variations after the buckling threshold, the thinner the rod, the more abrupt the variations. Eventually, a mismatch for half of the frequencies at buckling arises between the zero thickness limit of the extensible model and the inextensible model. © 2011 Elsevier Ltd. All rights reserved.

  10. 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

  11. Ultrasonic detection of cracks in uniaxial glass fibre rods

    CSIR Research Space (South Africa)

    Loveday, PW

    2006-01-01

    Full Text Available A non-destructive examination procedure based on a guided wave inspection approach is used for the acoustic examination of glass fibre reinforced composite rods. This paper contains an investigation into the characteristics of guided wave...

  12. Safety coupling for a control rod of a nuclear reactor

    International Nuclear Information System (INIS)

    Mindnich, F.R.; Friedrichs, H.; Schoettle, J.

    1978-01-01

    A coupling is presented between a control rod and the drive shafft arranged below. The construction of this coupling is designed in such a way that the usual sealing maesures against the escape of coolant are reduced. (TK) [de

  13. Failed fuel rod detection method by ultrasonic wave

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  14. On the resonant behavior of longitudinally vibrating accreting rods

    CSIR Research Space (South Africa)

    Shatalov, M

    2012-09-01

    Full Text Available The theory of accreting structures is a new and fast developing branch of analytical mechanics basing on the theory of partial differential and integral equations. In the present paper the authors analyze qualitative properties of accreting rods...

  15. Does rim microstructure formation degrade the fuel rod performance?

    International Nuclear Information System (INIS)

    Baron, D.; Spino, J.

    2002-01-01

    High burnup extension of LWR fuel is progressing to reduce the total process flow and eventually the costs of the nuclear fuel cycle. A particular fuel restructuring at high burnups, commonly observed at the periphery of LWR fuel pellets (rim structure), but also in FBR fuels to some extent and in the Plutonium rich clusters of the MOX Fuels, was considered a priori as a limitation for burnup extension. Since more than ten years this rim effect have been deeply investigated. Its causes and consequences are however not yet totally elucidated. The three steps actually identified of this phenomenon are first a progressive disappearing of the intra-granular Xenon, the outset of numerous 0.5 to 1 m pores and finally a grain subdivision around the pores. Penalty of the porosity increase on the thermal conductivity is obvious. One expect the fission gases to remain trapped in the rim porosity up to a 75 MWd/kgUO 2 local burnup. Above this threshold, 15 to 20 % of the fission gases seem to be quickly released. Microindentation tests conducted at ITU have shown the rim structure to resist fracture extension under punching. It is still open whether this implies certain ductility and viscosity of the material, or if it corresponds to stress relaxation by microcracking. Whatever the case be, it is suggested that the rim material would be able to decrease the interaction stresses and to equalise the cladding strains during a power ramp. Moreover, in the RIA tests, it was concluded so far that the grain de-cohesion caused by gas expansion at the grain boundaries was responsible for the cladding strain and failure. However, not the rim zone was affected by grain de-cohesion but the region adjacent to it. Therefore, in front of the question whether the rim structure degrades the fuel rod behaviour, we continue to argue on its benefit for fuel burnup extension. (author)

  16. Measurement of blockage in deformed LWR multi-rod arrays

    International Nuclear Information System (INIS)

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

    1983-01-01

    This paper critically reviews the current methods used for measuring blockage in multi-rod arrays and discusses their application. A new definition which overcomes the deficiencies of the previous methods is proposed. Also examples of the application of automatic computerised techniques to directly measure rod strain, blockage, sub-channel blockage and perimeter changes from photographs of sections through deformed arrays are presented. (author)

  17. Double-clad nuclear-fuel safety rod

    Science.gov (United States)

    McCarthy, W.H.; Atcheson, D.B.

    1981-12-30

    A device for shutting down a nuclear reactor during an undercooling or overpower event, whether or not the reactor's scram system operates properly. This is accomplished by double-clad fuel safety rods positioned at various locations throughout the reactor core, wherein melting of a secondary internal cladding of the rod allows the fuel column therein to shift from the reactor core to place the reactor in a subcritical condition.

  18. Downflow film boiling in a rod bundle at low pressure

    International Nuclear Information System (INIS)

    Hochreiter, L.E.; Rosal, E.R.; Fayfich, R.R.

    1978-01-01

    A series of low pressure downflow film boiling heat transfer experiments were conducted in a 14-foot (4.27 m) long electrically heater rod bundle containing 336 heater rods. The resulting data was compared with the Dougall-Rohsenow dispersed flow film boiling correlation. The data was found to lie below this correlation as the quality was increased. It is believed that buoyancy effects decreased the heat transfer in downflow film boiling. (author)

  19. CHF prediction in rod bundles using round tube data

    Energy Technology Data Exchange (ETDEWEB)

    Souza, Wallen F.; Veloso, Maria A.F.; Pereira, Cláubia; Costa, Antonella L., E-mail: wallenfds@yahoo.com.br, E-mail: mdora@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear

    2017-07-01

    The present work concerns the use of 1995 CHF table for uniformly heated round tubes, developed jointly by Canadian and Russian researchers, for the prediction of critical heat fluxes in rod bundles geometries. Comparisons between measured and calculated critical heat fluxes indicate that this table could be applied to rod bundles provided that a suitable correction factor is employed. The tolerance limits associated with the departure from nucleate boiling ratio (DNBR) are evaluated by using statistical analysis. (author)

  20. Modelling of fuel rod hydriding failures in water reactors

    International Nuclear Information System (INIS)

    Afanas'eva, E.Yu.; Evdokimov, I.A.; Khoruzhij, O.V.; Likhanskij, V.V.; Sorokin, A.A.

    2003-01-01

    Mechanistic models which were developed to describe primary hydriding phenomena in claddings of initially intact rods with residual moisture are described. The models include the following key processes: fuel rod thermal behavior, UO 2 fuel oxidation in steam-hydrogen atmosphere under irradiation, hydrogen diffusion in zirconium and in the hydride, growth of the hydride phase. Fuel rod thermomechanical behavior is calculated by using RTOP integral fuel code. An oxidation model represents the effects of temperature dynamics and temperature profile along fuel axis and radius on fuel oxidation as well as on hydrogen accumulation inside the fuel rod. Along with ordinary thermal dissociation of water molecules, the oxidation model also addresses radiolysis of the steam-hydrogen mixture due to fission fragments. The present radiolysis model takes into account the effects of the gas mixture composition, temperature and pressure. A new model of cladding hydriding is proposed in which calculation of the massive hydride growth is performed in 2-D geometry. Hydrogen transport in zirconium cladding is modeled with account for thermodiffusion. The RTOP code comprising the models developed allows us to calculate different scenarios of hydriding rod failures under given operation conditions. Test calculations were carried out and compared to available data. It is shown that there are threshold values of initial steam content inside the intact fuel rod which lead to the possibility of through-cladding hydride growth and formation of the primary defect. The threshold values depend on the oxidation state of the cladding inner surface, linear power profile in the fuel rod, fuel rod geometry, cladding temperature conditions and hydrogen diffusivities in zirconium and zirconium hydride

  1. 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)

  2. Normal temperature ventilating homogenity test of control rod drive mechanism

    International Nuclear Information System (INIS)

    Zhu Longxing.

    1987-01-01

    This paper describes the cooling for the control rod drive mechanism. It emphatically introduces some problems which must be considered in the test on ventilating homogenity under normal temperature for the control rod drive mechanism at the top of a reactor such as the selection of cooling shroud assemblies, installed position of ducts, volume of the static pressure container, etc. The test data of blowing-in and induced draft are compared and analysed

  3. Apparatus for injection casting metallic nuclear energy fuel rods

    Science.gov (United States)

    Seidel, Bobby R.; Tracy, Donald B.; Griffiths, Vernon

    1991-01-01

    Molds for making metallic nuclear fuel rods are provided which present reduced risks to the environment by reducing radioactive waste. In one embodiment, the mold is consumable with the fuel rod, and in another embodiment, part of the mold can be re-used. Several molds can be arranged together in a cascaded manner, if desired, or several long cavities can be integrated in a monolithic multiple cavity re-usable mold.

  4. Status and development of RBMK fuel rods and reactor materials

    International Nuclear Information System (INIS)

    Bibilashvili, Yu.K.; Reshetnikov, F.G.; Ioltukhovsky, A.G.

    1998-01-01

    The paper presents current status and development of RBMK fuel rods and reactor materials. With regard to fuel rod cladding the following issues have been discussed: corrosion, tensile properties, welding technology and testing of an alternative cladding alloy with a composition of Zr-Nb-Sn-Fe. Erbium doped fuel has been suggested for safety improvement. Also analysis of fuel reliability is presented in the paper. (author)

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

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  6. Estimating modal instability threshold for photonic crystal rod fiber amplifiers

    DEFF Research Database (Denmark)

    Johansen, Mette Marie; Hansen, Kristian Rymann; Laurila, Marko

    2013-01-01

    We present a semi-analytic numerical model to estimate the transverse modal instability (TMI) threshold for photonic crystal rod amplifiers. The model includes thermally induced waveguide perturbations in the fiber cross section modeled with finite element simulations, and the relative intensity...... noise (RIN) of the seed laser, which seeds mode coupling between the fundamental and higher order mode. The TMI threshold is predicted to ~370 W – 440 W depending on RIN for the distributed modal filtering rod fiber....

  7. Activity determination of the Am-241 radioactive lightning rods

    International Nuclear Information System (INIS)

    Dellamano, Jose C.; Minematsu, Denise; Potiens Jr, Ademar J.

    2008-01-01

    Full text: The radioactive lightning rods had been manufactured in Brazil up to 1989, when the Comissao Nacional de Energia Nuclear (CNEN) lifted the license for manufacture, commerce and installation of these devices. Since this date, the radioactive lightning rods have been replaced for conventional protection systems against electric discharges and have been sent to the institutes subordinated to the CNEN, amongst them the Instituto de Pesquisas Energeticas e Nucleares (IPEN-CNEN/SP). The radioactive lightning rods are constituted in its majority for a central metallic rod where the plates are mounted. Am-241 radioactive sources are fixed in these plates. The treatment of these devices is made in a glove box, where mechanically the sources are separate of the plates and connecting rods, placed in a metallic package and stored for posterior characterization, final packaging, intermediate storage and final disposal. In accordance with manufacturers information had been installed in Brazil, approximately 75,000 units with activities varying between 25 and 92 MBq. Preliminary studies were carried out in some of the 16,000 lightning rods received by the Laboratorio de Rejeitos Radioativos (LRR) of the IPEN-CNEN/SP, and demonstrated that the variation of the values of activity is very bigger. The implantation of a methodology for the radioisotope characterization of the Am-241 removed sources of the radioactive lightning rods is important because the isotope inventory is necessary for the certification of the processes considered for packaging and storage, besides being indispensable data for the final disposal. It is convenient mentioning that one is not about the determination of activity of a radioactive source with geometry and defined characteristics, but the implantation of a measure protocol for groups of sources that will be used in the routine tasks of the LRR. The current work presents the methodology developed for the radioisotope characterization of the Am

  8. Local hydrodynamic characteristics of regular triangular lattice of rods

    International Nuclear Information System (INIS)

    Mantlik, F.; Hejna, J.; Cervenka, J.

    1976-06-01

    Results are presented of an experimental investigation of the friction factor, velocity fields and shear stress distribution around a wetted perimeter in a rod bundle of a triangular lattice with a pitch-to-diameter ratio of 1.17. Measurements were made on 19-rod aerodynamical model at the Reynolds number of 42 300 and 211 000. The results indicated a highly significant effect of secondary flow. (author)

  9. Influence of the post-annealing cooling rate on the superconducting and mechanical properties of LFZ textured Bi-2212 rods

    International Nuclear Information System (INIS)

    Natividad, E; Gomez, J A; Angurel, L A; Salazar, A; Pastor, J Y; Llorca, J

    2002-01-01

    Laser floating zone textured Bi 2 Sr 2 CaCu 2 O 8+δ (Bi-2212) thin rods were manufactured and subjected to a two-step annealing process at 870 deg C and 801 deg C in air. It was found that the subsequent cooling process led to marked changes in electrical properties. Three cooling rates were tested: (i) quenching in liquid nitrogen, (ii) cooling in air inside an alumina tube and (iii) cooling inside the furnace. The results showed that the faster the cooling rate, the higher the normal state resistivity. The T c distribution across the rods was also affected by the cooling rate, but no large differences were observed in the magnitude of the critical current at 77 K since the homogeneity of furnace-cooled samples compensated for the higher outer J c values of fast-cooled ones. The mechanical properties (elastic modulus and flexure strength) were not influenced by the cooling rate, but the samples quenched in liquid nitrogen were often cracked by thermal shock. The elastic modulus and the flexure strength of the rods were deteriorated by the existence of an outer ring of compact, poorly textured material and by the large bubbles found in the central region of the rod. Samples processed by a two-step texturing process which reduced the thickness of the outer ring and eliminated the bubbles had better electrical and mechanical properties

  10. Integrability of a conducting elastic rod in a magnetic field

    Energy Technology Data Exchange (ETDEWEB)

    Sinden, D; Heijden, G H M van der [Centre for Nonlinear Dynamics, University College London, Chadwick Building, Gower Street, London WC1E 6BT (United Kingdom)

    2008-02-01

    We consider the equilibrium equations for a conducting elastic rod placed in a uniform magnetic field, motivated by the problem of electrodynamic space tethers. When expressed in body coordinates the equations are found to sit in a family of non-canonical Hamiltonian systems involving an increasing number of vector fields. These systems, which include the classical Euler and Kirchhoff rods, are shown to be completely integrable in the case of a transversely isotropic rod; they are in fact generated by a Lax pair. For the magnetic rod this gives a physical interpretation to a previously proposed abstract nine-dimensional integrable system. We use the conserved quantities to reduce the equations to a four-dimensional canonical Hamiltonian system, allowing the geometry of the phase space to be investigated through Poincare sections. In the special case where the force in the rod is aligned with the magnetic field the system turns out to be superintegrable, meaning that the phase space breaks down completely into periodic orbits, corresponding to straight twisted rods.

  11. POWER LEVEL EFFECT IN A PWR ROD EJECTION ACCIDENT

    International Nuclear Information System (INIS)

    Diamond, D.J.; Bromley, B.P.; Aronson, A.L.

    2002-01-01

    The purpose of this study is to determine the effect of the initial power level during a rod ejection accident (REA) on the ejected rod worth and the resulting energy deposition in the fuel. The model used is for the hot zero power (HZP) conditions at the end of a typical fuel cycle for the Three Mile Island Unit 1 pressurized water reactor. PARCS , a transient, three-dimensional, two-group neutron nodal diffusion code, coupled with its own thermal-hydraulics model, is used to perform both steady-state and transient simulations. The worth of an ejected control rod is affected by both power level, and the positions of control banks. As the power level is increased, the worth of a single central control rod tends to drop due to thermal-hydraulic feedback and control bank removal, both of which flatten the radial neutron flux and power distributions. Although the peak fuel pellet enthalpy rise during an REA will be greater for a given ejected rod worth at elevated initial power levels, it is more likely the HZP condition will cause a greater net energy deposition because an ejected rod will have the highest worth at HZP. Thus, the HZP condition can be considered the most conservative in a safety evaluation

  12. 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.

  13. 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.

  14. Key developments of a rod control system - 15101

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  15. Reactor core and control rod assembly in FBR type reactor

    International Nuclear Information System (INIS)

    Fujimura, Koji; Kawashima, Katsuyuki; Itooka, Satoshi.

    1993-01-01

    Fuel assemblies and control rod assemblies are attached respectively to reactor core support plates each in a cantilever fashion. Intermediate spacer pads are disposed to the lateral side of a wrapper tube just above the fuel rod region. Intermediate space pads are disposed to the lateral side of a control rod guide tube just above a fuel rod region. The thickness of the intermediate spacer pad for the control rod assembly is made smaller than the thickness of the intermediate spacer pad for the fuel assembly. This can prevent contact between intermediate spacer pads of the control guide tube and the fuel assembly even if the temperature of coolants is elevated to thermally expand the intermediate spacer pad, by which the radial displacement amount of the reactor core region along the direction of the height of the control guide tube is reduced substantially to zero. Accordingly, contribution of the control rod assembly to the radial expansion reactivity can be reduced to zero or negative level, by which the effect of the negative radial expansion reactivity of the reactor is increased to improve the safety upon thermal transient stage, for example, loss of coolant flow rate accident. (I.N.)

  16. Calculation method for control rod dropping time in reactor

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  17. Rod internal pressure quantification and distribution analysis using Frapcon

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-01

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

  18. Instrumentation failure following pedicle subtraction osteotomy: the role of rod material, diameter, and multi-rod constructs.

    Science.gov (United States)

    Luca, Andrea; Ottardi, Claudia; Sasso, Maurizio; Prosdocimo, Liliana; La Barbera, Luigi; Brayda-Bruno, Marco; Galbusera, Fabio; Villa, Tomaso

    2017-03-01

    Pedicle subtraction osteotomy (PSO) has a complication rate noticeably higher than other corrective surgical techniques used for the treatment of spinal sagittal imbalance. In particular, rod breakage and pseudoarthrosis remain burning issues of this technique. Goal of this study was to investigate the biomechanical performance of several hardware constructs. The study was performed using two validated finite element models of the lumbosacral spine (L1-S1) incorporating a PSO on L3 and L4, respectively. Both models were instrumented two levels above and below the osteotomy site. Different combinations of materials (Ti6Al4V and Cr-Co) and device configurations (bilateral single vs. double rod, rod diameters of 5 and 6 mm) were investigated. The loading was represented considering a force of 500 N (imposed along the spinal curvature and connecting the vertebral bodies) and pure moments of 7.5 Nm in flexion-extension, lateral bending and axial rotation. The results were evaluated in terms of range of motion (ROM), load, and stresses acting on the instrumentation. A comparable ROM was found for all the models. The simulations showed a different behavior of the devices: increasing the stiffness an 8-19% increase of the load was calculated on the rod. However, the stress on the instrumentation resulted higher on Cr-Co devices and on smaller rods. The highest stress reduction (up to 50%) was ensured using double rod constructs. The bilateral double parallel rods configuration resulted the best to reduce the stresses on the spinal fixators at the osteotomy site. However, the high loads acting on the rods with respect to the physiologic condition could slow down the bone healing at the osteotomy site.

  19. Calculation of the effectiveness of manual control rods for the reactor of Ignalina NPP Unit 2

    International Nuclear Information System (INIS)

    Bubelis, E.; Pabarcius, R.

    2001-01-01

    On the basis of one of the recent databases of the reactor of Ignalina NPP Unit 2, calculations of the effectiveness of separate manual control rods, groups of manual control rods and axial characteristic of effectiveness of separate manual control rods were performed. The results of the calculations indicated, that all analyzed separate manual control rods have approximately the same effectiveness, which doesn't depend on the location of a control rod in the reactor core layout Manual control rod of the new design has about 10% greater effectiveness than manual control rod of the old design. (author)

  20. Comparative Activity of Several Antimicrobial Agents against Nosocomial Gram-Negative Rods Isolated across Canada

    Directory of Open Access Journals (Sweden)

    Shelley R Scriver

    1995-01-01

    Full Text Available In 1992, a surveillance study was performed in Canada to determine the susceptibility of nosocomial Gram-negative rods to several wide spectrum antimicrobials. Consecutive isolates from 10 institutions, as well as additional strains of selected species of Enterobacteriaceae that are known to possess the Bush group 1 beta-lactamase, were tested for susceptibility to 12 antimicrobials. Third-generation cephalosporin resistance was found to be as high as 29% in Enterobacter cloacae that possesses the Bush group 1 beta-lactamase and less than 4% in those isolates not possessing this enzyme. Cefepime equalled or exceeded the activity of the third-generation cephalosporins against the species of Enterobacteriaceae that demonstrated resistance to the third-generation cephalosporins.

  1. Use of Supplemental Short Pre-Contoured Accessory Rods and Cobalt Chrome Alloy Posterior Rods Reduces Primary Rod Strain and Range of Motion Across the Pedicle Subtraction Osteotomy Level: An In Vitro Biomechanical Study.

    Science.gov (United States)

    Hallager, Dennis Winge; Gehrchen, Martin; Dahl, Benny; Harris, Jonathan A; Gudipally, Manasa; Jenkins, Sean; Wu, Ai-Min; Bucklen, Brandon S

    2016-04-01

    In vitro cadaveric biomechanical study. To assess effects of 4-rod reconstruction, rod material, and anterior column support on motion and surface rod strain in a pedicle subtraction osteotomy model. Pedicle subtraction osteotomy (PSO) can correct significant sagittal deformity of the lumbar spine; however, revision rates are high. To reduce rod strain and the incidence of rod fracture, clinical use of multi-rod construction, cobalt chrome (CoCr) alloy rods, and interbody spacers adjacent to PSO has been proposed. Investigating both motion and rod strain is necessary to determine the biomechanical efficacy of these techniques. Five specimens (T12-S1) underwent PSO at L3 with pedicle screw stabilization at L1-S1. Pedicle subtraction was adjusted to achieve a final lordosis of 70°. Flexion-extension (FE), lateral bending, and axial rotation were applied. Linear strain gauges measured surface rod strain during FE on primary and accessory rods at PSO level. Testing evaluated (1) accessory rods (4-Rod) added at PSO level versus primary rods (2-Rod); (2) Ti versus CoCr rods; and (3) lateral interbody spacers (S) inserted adjacent to PSO. One-way and three-way analysis of variance was performed (P ≤ 0.05). All constructs significantly reduced FE and lateral bending motion relative to intact (P < 0.001). The main effect of accessory rods in reducing FE motion was significant (P = 0.021). Accessory and CoCr rods reduced relative surface strain on the primary rod, irrespective of construct (P < 0.001). CoCr 4-Rod + S provided the greatest reduction in strain (76% decrease; P = 0.003). Accessory and CoCr rods provided greatest reduction in motion and rod strain at PSO level. Interbody devices minimally affected motion-induced strain and might act primarily to maintain disc height. Clinicians must assess whether surface strain and motion reduction minimize the incidence of rod fracture. N/A.

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

    International Nuclear Information System (INIS)

    Yanagisawa, Kazuaki

    1985-12-01

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

  3. Fed-batch production of the hydrophobins RodA and RodB from Aspergillus fumigatus in host Pichia pastoris

    DEFF Research Database (Denmark)

    Pedersen, Mona Højgaard; Borodina, Irina; Frisvad, Jens Christian

    Objectives: Aspergillus fumigatus expresses the hydrophobins RodA and RodB on the surface of its conidia. RodA is known to be important for the pathogenesis of the fungus, but the role of RodB is unknown. The aim was to produce recombinant RodA and RodB for further characterication. Methods....... The expression of the RodA and RodB genes was first studied in culture flasks in buffered complex methanol medium as protein production was dependent on the methanol-induced AOX1 promoter. Later production was scaled up to a 2 L fed-batch fermentor. Hydrophobins were purified using His-select Nickel Affinity gel...

  4. Estimation of rod scale errors in geodetic leveling

    Science.gov (United States)

    Craymer, Michael R.; Vaníček, Petr; Castle, Robert O.

    1995-01-01

    Comparisons among repeated geodetic levelings have often been used for detecting and estimating residual rod scale errors in leveled heights. Individual rod-pair scale errors are estimated by a two-step procedure using a model based on either differences in heights, differences in section height differences, or differences in section tilts. It is shown that the estimated rod-pair scale errors derived from each model are identical only when the data are correctly weighted, and the mathematical correlations are accounted for in the model based on heights. Analyses based on simple regressions of changes in height versus height can easily lead to incorrect conclusions. We also show that the statistically estimated scale errors are not a simple function of height, height difference, or tilt. The models are valid only when terrain slope is constant over adjacent pairs of setups (i.e., smoothly varying terrain). In order to discriminate between rod scale errors and vertical displacements due to crustal motion, the individual rod-pairs should be used in more than one leveling, preferably in areas of contrasting tectonic activity. From an analysis of 37 separately calibrated rod-pairs used in 55 levelings in southern California, we found eight statistically significant coefficients that could be reasonably attributed to rod scale errors, only one of which was larger than the expected random error in the applied calibration-based scale correction. However, significant differences with other independent checks indicate that caution should be exercised before accepting these results as evidence of scale error. Further refinements of the technique are clearly needed if the results are to be routinely applied in practice.

  5. Control Rod Driveline Reactivity Feedback Model for Liquid Metal Reactors

    International Nuclear Information System (INIS)

    Kwon, Young-Min; Jeong, Hae-Yong; Chang, Won-Pyo; Cho, Chung-Ho; Lee, Yong-Bum

    2008-01-01

    The thermal expansion of the control rod drivelines (CRDL) is one important passive mitigator under all unprotected accident conditions in the metal and oxide cores. When the CRDL are washed by hot sodium in the coolant outlet plenum, the CRDL thermally expands and causes the control rods to be inserted further down into the active core region, providing a negative reactivity feedback. Since the control rods are attached to the top of the vessel head and the core attaches to the bottom of the reactor vessel (RV), the expansion of the vessel wall as it heats will either lower the core or raise the control rods supports. This contrary thermal expansion of the reactor vessel wall pulls the control rods out of the core somewhat, providing a positive reactivity feedback. However this is not a safety factor early in a transient because its time constant is relatively large. The total elongated length is calculated by subtracting the vessel expansion from the CRDL expansion to determine the net control rod expansion into the core. The system-wide safety analysis code SSC-K includes the CRDL/RV reactivity feedback model in which control rod and vessel expansions are calculated using single-nod temperatures for the vessel and CRDL masses. The KALIMER design has the upper internal structures (UIS) in which the CRDLs are positioned outside the structure where they are exposed to the mixed sodium temperature exiting the core. A new method to determine the CRDL expansion is suggested. Two dimensional hot pool thermal hydraulic model (HP2D) originally developed for the analysis of the stratification phenomena in the hot pool is utilized for a detailed heat transfer between the CRDL mass and the hot pool coolant. However, the reactor vessel wall temperature is still calculated by a simple lumped model

  6. Strategy for Fuel Rod Receipt, Characterization, Sample Allocation for the Demonstration Sister Rods

    Energy Technology Data Exchange (ETDEWEB)

    Marschman, Steven C. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Warmann, Stephan A. [Portage, Inc., Idaho Falls, ID (United States); Rusch, Chris [NAC International, Inc., Norcross, GA (United States)

    2014-03-01

    , inert gas backfilling, and transfer to an Independent Spent Fuel Storage Installation (ISFSI) for multi-year storage. To document the initial condition of the used fuel prior to emplacement in a storage system, “sister ” fuel rods will be harvested and sent to a national laboratory for characterization and archival purposes. This report supports the demonstration by describing how sister rods will be shipped and received at a national laboratory, and recommending basic nondestructive and destructive analyses to assure the fuel rods are adequately characterized for UFDC work. For this report, a hub-and-spoke model is proposed, with one location serving as the hub for fuel rod receipt and characterization. In this model, fuel and/or clad would be sent to other locations when capabilities at the hub were inadequate or nonexistent. This model has been proposed to reduce DOE-NE’s obligation for waste cleanup and decontamination of equipment.

  7. 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

  8. Analysis of Explanted Magnetically Controlled Growing Rods From Seven UK Spinal Centers.

    Science.gov (United States)

    Joyce, Thomas J; Smith, Simon L; Rushton, Paul R P; Bowey, Andrew J; Gibson, Michael J

    2018-01-01

    Analysis of explanted MAGnetic Expansion Control (MAGEC) growing rods. To analyze explanted MAGEC rods used in management of early onset scoliosis and identify the mode of failure in such cases. Magnetically controlled growing rods are increasingly used as the option of choice for early onset scoliosis. However, being more complex than conventional growing rods they are perhaps more likely to succumb to multifarious failure modes. In addition, metallosis has been reported around failed MAGEC rods. Explanted MAGEC rods from seven UK spinal centers were obtained for independent analysis. Thirty-four MAGEC rods, from 18 children, explanted for reasons including failure of rod lengthening and maximum rod distraction reached, were cut open to allow internal components to be evaluated and assessed. Externally, all MAGEC rods showed localized marks, which were termed "growth marks" as they indicated growth of the rod in vivo, on the extending bar component. After cutting open, titanium wear debris was found inside all 34 (100%) MAGEC rods. Ninety-one percent (31/34) of MAGEC rods showed measurable wear of the extending bar, towards the magnet end. Substantial damage to the radial bearing was seen inside 74% (25/34) of MAGEC rods while O-ring seal failure was seen in 53% (18/34) of cases. In 44% (15/34) of MAGEC rods the drive pin was fractured but this was felt to be an effect of rod failure, not a cause. The combination of high volumes of titanium wear debris alongside O-ring seal damage likely accounts for the metallosis reported clinically around some MAGEC rods. Based on this explant data, a failure mechanism in MAGEC rods due to the natural off axis loading in the spine was proposed. This is the largest data set reporting a complete analysis of explanted MAGEC rods to date. 4.

  9. 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

  10. 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)

  11. Device and method of cooling control rod drives

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  12. Failed fuel rod detection system and computerized manipulator during outages

    International Nuclear Information System (INIS)

    Boehm, H.H.; Foerch, H.

    1984-01-01

    During regular outages spent fuel assemblies need to be replaced and relocated within the core. Defective fuel rods in particular fuel assemblies have to be removed from further service and before delivery of such faulty fuel assemblies to a reprocessing plant. The system which Brown Boveri Reaktor GmbH and Krautkraemer have developed in the Federal Republic of Germany is capable of directly locating the defective rods in a proper fuel assembly. Inspection times are comparable to those of standard sipping methods, with the advantages of immediately available results and direct identification of the defective fuel rods. During the repair of fuel assemblies this system allows withdrawal of individual defective rods. With the sipping method all the fuel rods of a defective fuel assembly need to be removed and inspected by eddy current testing. During steam generator inspection and repair personnel are exposed to ample radiation. A remotely controlled, computerized manipulator was used to significantly reduce the radiation dose by automating steps in the procedures; at the same time inspection and repair times were reduced. The main features of the manipulator are a rigid component construction of the leg and two arms, and a resolver control for horizontal and vertical motion that enables rapid and accurate access to a desired tube (author)

  13. The graphic model of the spent fuel rod extracting system

    International Nuclear Information System (INIS)

    Yoon, Jee Sup; Kim, Sung Hyun

    1997-01-01

    The spent fuel rod extracting system is being developed in KAERI to deal with problems associated with utilization of storage pools at nuclear power plants. This system consists of an equipment system for extracting rods from spent fuel assemblies, a machine controller, and a supervisory controller. The performance of extraction system has been investigated through a series of experiments. Even though the system is designed to automatically perform sequential procedures, several problems have been found such as the gripper stucking to fuel rod caused by misaligned positioning and the socket jamming of impact wrench into the nut, etc. Up to this end the graphical model of the rod extracting system has been made so that possible sequences of operations including error detection and recovery actions are verified by using a graphic simulation before real operations. For the implementation, IGRIP is being used as a multifunctional tool for developing the rod extraction system. IGRIP is not only an excellent visualization tool, but it also highlights modeling virtual machine. (author). 6 refs., 1 tab., 6 figs

  14. Substitute safety rods: Physics of operation and irradiation

    International Nuclear Information System (INIS)

    Baumann, N.P.

    1991-01-01

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

  15. 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....

  16. Fault pseudotachylyte: a coseismic lightning rod

    Science.gov (United States)

    Ferre, E. C.; Conder, J. A.; MathanaSekaran, N.; Geissman, J. W.

    2013-12-01

    of melt during the formation of a pseudotachylite vein. The increase in melt temperature is the most important factor affecting electrical conductivity in the fault plane. When the melt temperature rises from 1300 to 2000K, its electrical conductivity increases about 80 times. This implies that once a continuous pseudotachylite sheet-like vein is formed during an earthquake, the vein has a much higher electrical conductivity than its host-rock. The dramatic increase in electrical conductivity along the pseudotachylite plane might be synchronous with the generation of the coseismic electrical current. Thus, regardless of its origin, any electrical current produced during an earthquake will travel along the pseudotachylite plane which acts as a lightning rod. The magnetization of a solid due to an electrical current results from Biot-Savart law which states that an electrical current generates a magnetic field. The solidification of the pseudotachylite vein does not happen at once but proceeds from the margin inwards as an electrical current may still pass through the conducting pseudotachylite. Therefore, the host-rock of the pseudotachylite vein or its solidified margin can be magnetized by a coseismic current.

  17. 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.

  18. Refabrication of fuel rods - qualification of the end plug welds

    International Nuclear Information System (INIS)

    Sannen, Leo; Gys, August; Parthoens, Yves

    2005-01-01

    Refabrication of irradiated fuel rods is applied at SCK/CEN, both to make short fuel rodlets for tests in research reactors and to reconstitute full-size rods for their reinsertion in the original fuel assembly as an elegant back end solution for industrial fuel rods after their use in fuel research programs. In both cases the end cap welds have to be qualified thoroughly, to prove their proper performance either under irradiation and/or during long-term storage. The paper describes the qualification process that is applied at the hot laboratory LHMA at SCK/CEN to qualify the welding methodology and the actual welds made according this methodology. The results obtained on a typical refabrication case are included. (Author)

  19. Unfolding/refolding studies of the myosin rod.

    Science.gov (United States)

    Nozais, M; Bechet, J J

    1993-12-15

    The effect of guanidine hydrochloride on the gel-filtration chromatography, viscosity, far ultraviolet circular dichroism and fluorescence emission intensity of the myosin rod was studied under equilibrium conditions. The normalized transition curves for each of these methods were comparable with a midpoint at a guanidine hydrochloride concentration of 1.75-2 M. The curves were not, however, superposable, suggesting that the loss of helix content and the dissociation of the two chains of the myosin rod were not tightly linked. Furthermore, they were unexpectedly independent of the protein concentration over 0.05-20 microM. These phenomena are interpreted taking into account the large size of the molecule. A step-wise process is proposed as a model for the unfolding of the myosin rod.

  20. Nuclear reactor fuel rod behavior modelling and current trends

    International Nuclear Information System (INIS)

    Colak, Ue.

    2001-01-01

    Safety assessment of nuclear reactors is carried out by simulating the events to taking place in nuclear reactors by realistic computer codes. Such codes are developed in a way that each event is represented by differential equations derived based on physical laws. Nuclear fuel is an important barrier against radioactive fission gas release. The release of radioactivity to environment is the main concern and this can be avoided by preserving the integrity of fuel rod. Therefore, safety analyses should cover an assessment of fuel rod behavior with certain extent. In this study, common approaches for fuel behavior modeling are discussed. Methods utilized by widely accepted computer codes are reviewed. Shortcomings of these methods are explained. Current research topics to improve code reliability and problems encountered in fuel rod behavior modeling are presented

  1. Topologically Directed Assemblies of Semiconducting Sphere-Rod Conjugates.

    Science.gov (United States)

    Lin, Zhiwei; Yang, Xing; Xu, Hui; Sakurai, Tsuneaki; Matsuda, Wakana; Seki, Shu; Zhou, Yangbin; Sun, Jian; Wu, Kuan-Yi; Yan, Xiao-Yun; Zhang, Ruimeng; Huang, Mingjun; Mao, Jialin; Wesdemiotis, Chrys; Aida, Takuzo; Zhang, Wei; Cheng, Stephen Z D

    2017-12-27

    Spontaneous organizations of designed elements with explicit shape and symmetry are essential for developing useful structures and materials. We report the topologically directed assemblies of four categories (a total of 24) of sphere-rod conjugates, composed of a sphere-like fullerene (C 60 ) derivative and a rod-like oligofluorene(s) (OF), both of which are promising organic semiconductor materials. Although the packing of either spheres or rods has been well-studied, conjugates having both shapes substantially enrich resultant assembled structures. Mandated by their shapes and topologies, directed assemblies of these conjugates result not only in diverse unconventional semiconducting supramolecular lattices with controlled domain sizes but also in tunable charge transport properties of the resulting structures. These results demonstrate the importance of persistent molecular topology on hierarchically assembled structures and their final properties.

  2. CFD modeling of secondary flows in fuel rod bundles

    International Nuclear Information System (INIS)

    Baglietto, Emilio; Ninokata, Hisashi

    2004-01-01

    An optimized non-linear eddy viscosity model is introduced, for calculations of detailed coolant velocity distribution in a tight lattice fuel bundle. The low Reynolds formulation has been optimized based on DNS data for channel flow. The non-linear stress-strain relationship has been modified in the coefficients to model the flow anisotropy, which causes the formation of turbulence driven secondary flows inside the bundle subchannels. Predictions of the model are first compared to experimental measurements of secondary flows in a triangularly arrayed rod bundle with p/d=1.3. Subsequently wall shear stress and velocity predictions are compared with different experimental data for a rod bundle with p/d=1.17. The model shows to be able to correctly reproduce the scale of the secondary motion, and to accurately reproduce both wall shear stress and velocity distributions inside the rod bundle subchannels. (author)

  3. Disassembling and cleaning apparatus for control rod drives

    International Nuclear Information System (INIS)

    Yoshida, Tomiji; Sasaki, Masayoshi; Numata, Nobumasa; Oowada, Masataka; Arazoe, Masatoshi.

    1981-01-01

    Purpose: To shorten the working period, as well as decrease the number of operators and radiation exposure doses for disassembling and cleaning of the control rod drives. Constitution: The apparatus comprises a cleaning tank, clamp units, inner and outer surface cleaning nozzles, a high pressure pump, a drying tank unit, electronic components and the likes. This enables to render the various works automatic, such as wash out of outer and inner filters and strainers at high radiation dose levels, extraction and cleaning of a piston tube and an index tube, cleaning for the control rod drive main body, and cleaning and drying of small components, whereby the disassembling, cleaning and assembling works for the control rod drives, among all, the disassembling and cleaning works resulting relatively high radiation exposure doses to the operator can be made in a remote control and semi-automatic manner. (Furukawa, Y.)

  4. Characterisation of Plasma Filled Rod Pinch electron beam diode operation

    Science.gov (United States)

    MacDonald, James; Bland, Simon; Chittenden, Jeremy

    2016-10-01

    The plasma filled rod pinch diode (aka PFRP) offers a small radiographic spot size and a high brightness source. It operates in a very similar to plasma opening switches and dense plasma focus devices - with a plasma prefill, supplied via a number of simple coaxial plasma guns, being snowploughed along a thin rod cathode, before detaching at the end. The aim of this study is to model the PFRP and understand the factors that affect its performance, potentially improving future output. Given the dependence on the PFRP on the prefill, we are making detailed measurements of the density (1015-1018 cm-3), velocity, ionisation and temperature of the plasma emitted from a plasma gun/set of plasma guns. This will then be used to provide initial conditions to the Gorgon 3D MHD code, and the dynamics of the entire rod pinch process studied.

  5. A Method for Determining Reactivity-Time Function of Safety Rods

    International Nuclear Information System (INIS)

    Milovanovic, S.; Pesic, M.

    1994-01-01

    For accidental analysis of HERBE fast-thermal core, an accurate reactivity-time function for reactor safety rods is necessary. The HERBE core was designed with four safety rods: two of them are the actual safety rods, and the other two are additional safety rods which include holds during motion. The reactivity-time function is determined in two steps: (1) safety rods reactivity-position function is measured using inverse method; (2) rod drop position-time function is measured using a new method. In previously proposed method, it was determined by measurement of rod drop times and assuming constant acceleration during any particular interval of a rod motion. The complex dependence of the reactivity-time function for the HERBE safety rods during reactor shutdown is determined by combining both previously obtained reactivity worth data and measurements of safety rods trajectory. Integral reactivity-time function of the safety rods, including rods interference reactivity effects, is shown. In this new method an improvement for accurate safety rod position measurement, compared to previously proposed method, is obtained. At the same time, the assumption of the constant acceleration of the safety rods in the motion intervals is validated

  6. 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.

  7. An evaluation of control rod motion simulator of research reactor

    International Nuclear Information System (INIS)

    Sanda

    2010-01-01

    Motion simulator for rod control research reactor has been carried out using a servo motor. Reactor rod motion control at any point should be in the right position, one of the motors that can move in a precise and correct is the servo motor. To ensure that the servo motor to move in accordance with the desired program, then the servo motor function test should be carried out to ensure having good performance. Tests carried out on meshes stress disorder, the load is stable within a certain period and travel time safety control rod up and down, travel time regulating control rods up and down and travel time compensation control rods up and down. In testing the breakdown voltage V out nets at 24 V, 6.5 A with 12 Q load deviation obtained V0= V1 = 0.1% and 0.65% and for the stability of the load in a certain time deviation V = 0.7125% , next to the breakdown voltage V out nets at 12 V, 4.2 A with a 6 Q load deviation obtained V0= V1 = 0.275% and 1.158% for the stability of the load in a certain time deviation V = 1.463% and the net-voltage noise nets on V out 24 V, 4.5 A with 12 Q load deviation obtained V0 = V1 = 0.196% and 0.496% and for the stability of the load in a certain time deviation V = 0.3625%. While the travel time of a safety control rod up and down, up and down the regulator and compensation rise and fall showed a steady linear graph. The results show that the performance of the servo motor is very stable with the working area below the tolerance limit, it is 5% - 10%.(author)

  8. Core design of super LWR with double tube water rods

    International Nuclear Information System (INIS)

    Wu, Jianhui; Oka, Yoshiaki

    2014-01-01

    Highlights: • Supercritical light water cooled and moderated reactor with double tube water rods is developed. • Double-row fuel rod assembly and out-in fuel loading pattern are applied. • Separation plates in peripheral assemblies increase average outlet temperature. • Neutronic and thermal design criteria are satisfied during the cycle. - Abstract: Double tube water rods are employed in core design of super LWR to simplify the upper core structure and refueling procedure. The light water moderator flows up in the inner tube from the bottom of the core, then, changes the flow direction at the top of the core into the outer tube and flows out at the bottom of the core. It eliminates the moderator guide/distribution tubes into the single tube water rods from the top dome of the reactor pressure vessel of the previous super LWR design. Two rows of fuel rods are filled between the water rods in the fuel assembly. Out-in refueling pattern is adopted to flatten radial power distribution. The peripheral fuel assemblies of the core are divided into four flow zones by separation plates for increasing the average core outlet temperature. Three enrichment zones are used for axial power flattening. The equilibrium core is analyzed based on neutronic/thermal-hydraulic coupled model. The results show that, by applying the separation plates in peripheral fuel assemblies and low gadolinia enrichment, the maximum cladding surface temperature (MCST) is limited to 653 °C with the average outlet temperature of 500 °C. The inherent safety is satisfied by the negative void reactivity effects and sufficient shutdown margin

  9. 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

  10. Experiments to understand the corrosion process of fuel rod claddings

    International Nuclear Information System (INIS)

    Groeschel, F.; Hermann, A.

    1997-01-01

    Fuel rods in light water reactors have to respond to the trends in increased burn-up and extended dwelling time in reactor. Waterside corrosion of the cladding affecting wall thickness, mechanical stability due to hydriding and the heat transfer due to the low thermal conductivity of the oxide scale may become the limiting factors. The corrosion process is complex and involves a large variety of mechanisms. Understanding of the process is important for safe operation and a prerequisite for development of improved materials. A variety of analytical techniques and mechanical tests, including examination of irradiated pathfinder rods, are used to tackle the different aspects. (author) 6 figs., 1 tab., 17 refs

  11. Grid supports design for dual-cooled fuel rods

    Directory of Open Access Journals (Sweden)

    J Kim

    2016-09-01

    In this paper, the minimum spring force to prevent dual-cooled fuel rods from dropping during normal reactor operation is calculated. The spring characteristics of a cantilever type and a hemi-sphere type are predicted. A finite element analysis is carried out by using the commercial code ABAQUS. The analysis results are verified by experiments. Finally, it is checked whether the property of the suggested springs satisfies the minimum required spring force. Based on the obtained results, a kind of spacer grid candidate for dual cooled fuel rods, i.e. a spacer grid with hybrid supports is suggested.

  12. Determination of Ultimate Torque for Multiply Connected Cross Section Rod

    Directory of Open Access Journals (Sweden)

    V. L. Danilov

    2015-01-01

    Full Text Available The aim of this work is to determine load-carrying capability of the multiply cross-section rod. This calculation is based on the model of the ideal plasticity of the material, so that the desired ultimate torque is a torque at which the entire cross section goes into a plastic state.The article discusses the cylindrical multiply cross-section rod. To satisfy the equilibrium equation and the condition of plasticity simultaneously, two stress function Ф and φ are introduced. By mathematical transformations it has been proved that Ф is constant along the path, and a formula to find its values on the contours has been obtained. The paper also presents the rationale of the line of stress discontinuity and obtained relationships, which allow us to derive the equations break lines for simple interaction of neighboring circuits, such as two lines, straight lines and circles, circles and a different sign of the curvature.After substitution into the boundary condition at the end of the stress function Ф and mathematical transformations a formula is obtained to determine the ultimate torque for the multiply cross-section rod.Using the doubly connected cross-section and three-connected cross-section rods as an example the application of the formula of ultimate torque is studied.For doubly connected cross-section rod, the paper offers a formula of the torque versus the radius of the rod, the aperture radius and the distance between their centers. It also clearly demonstrates the torque dependence both on the ratio of the radii and on the displacement of hole. It is shown that the value of the torque is more influenced by the displacement of hole, rather than by the ratio of the radii.For the three-connected cross-section rod the paper shows the integration feature that consists in selection of a coordinate system. As an example, the ultimate torque is found by two methods: analytical one and 3D modeling. The method of 3D modeling is based on the Nadai

  13. Periodicity effects of axial waves in elastic compound rods

    DEFF Research Database (Denmark)

    Nielsen, R. B.; Sorokin, S. V.

    2015-01-01

    Floquet analysis is applied to the Bernoulli-Euler model for axial waves in a periodic rod. Explicit asymptotic formulae for the stop band borders are given and the topology of the stop band pattern is explained. Eigenfrequencies of the symmetric unit cell are determined by the Phase-closure Prin......Floquet analysis is applied to the Bernoulli-Euler model for axial waves in a periodic rod. Explicit asymptotic formulae for the stop band borders are given and the topology of the stop band pattern is explained. Eigenfrequencies of the symmetric unit cell are determined by the Phase...

  14. Detection of Microcracks in Trunnion Rods Using Ultrasonic Guided Waves

    Science.gov (United States)

    2015-07-01

    detection in rod waveguides, such as rock and soil anchors, are well described in a number of papers (Beard and Lowe 2003; He 2006 et al.; Yang 2011...anchor corrosion in soils (Beard et al.2003; Yang 2011; He et al. 2006). In general, these specific applications need high sensitivity at the rod...dimensional (3D) finite element modeling ( FEM ) as a means to avoid ERDC/ITL TR-15-1 17 pull-off requirements. It is important to note that this LF

  15. Review of design technology of control rod position indicators

    International Nuclear Information System (INIS)

    Yu, Je Yong; Huh, Hyung; Kim, Ji Ho; Kim, Jong In; Chang, Moon Hee

    1999-10-01

    An integral reactor SMART is under development at KAERI. The design characteristics of SMART are radically different from those employer in currently operating loop type water reactors in Korea. The objective of this report is to review the design technology of position indicator, and to study the various sensors which can be used in rod position indicator. Design criteria that rod position indicator should satisfy are also examined. Following position indicators are reviewed in this report. 1. Digital positioning indicator (DRPI), 2. Reed switch type position indicator (RSPT), 3. Choke sensor type position indicator, 4. Ultrasonic sensor type position indicator, 5. Comparison of each position indicator. (author)

  16. Operation of a 473 MHz four-rod cavity RFQ

    International Nuclear Information System (INIS)

    Kazimi, R.; Huson, F.R.; Mackay, W.W.; Meitzler, C.R.

    1992-01-01

    We have constructed a new type of four-rod Radio Frequency Quadrupole to operate at 473 MHz. Four-rod structures have not previously been built for such a high frequency. The RFQ is designed to accelerate 10 mA of H - ions from 30 keV to 0.5 MeV. The rf measurements and beam test of the RFQ have been performed successfully. Here we present operational results of the RFQ system including measurements of the beam current, the required rf power, energy, energy spread, and emittance. (Author) 8 refs., 6 figs., 2 tabs

  17. 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

  18. Exploration of a Permanent Magnet Synchronous Generator with Compensated Reactance Windings in Parallel Rod Configuration

    Science.gov (United States)

    Lyan, Oleg; Jankunas, Valdas; Guseinoviene, Eleonora; Pašilis, Aleksas; Senulis, Audrius; Knolis, Audrius; Kurt, Erol

    2018-02-01

    In this study, a permanent magnet synchronous generator (PMSG) topology with compensated reactance windings in parallel rod configuration is proposed to reduce the armature reactance X L and to achieve higher efficiency of PMSG. The PMSG was designed using iron-cored bifilar coil topology to overcome problems of market-dominant rotary type generators. Often the problem is a comparatively high armature reactance X L, which is usually bigger than armature resistance R a. Therefore, the topology is proposed to partially compensate or negligibly reduce the PMSG reactance. The study was performed by using finite element method (FEM) analysis and experimental investigation. FEM analysis was used to investigate magnetic field flux distribution and density in PMSG. The PMSG experimental analyses of no-load losses and electromotive force versus frequency (i.e., speed) was performed. Also terminal voltage, power output and efficiency relation with load current at different frequencies have been evaluated. The reactance of PMSG has low value and a linear relation with operating frequency. The low reactance gives a small variation of efficiency (from 90% to 95%) in a wide range of load (from 3 A to 10 A) and operation frequency (from 44 Hz to 114 Hz). The comparison of PMSG characteristics with parallel and series winding connection showed insignificant power variation. The research results showed that compensated reactance winding in parallel rod configuration in PMSG design provides lower reactance and therefore, higher efficiency under wider load and frequency variation.

  19. Conceptual Design Study on Electromagnets of Control Rod Drive Mechanism of a SFR

    International Nuclear Information System (INIS)

    Lee, Jaehan; Koo, Gyeonghoi

    2013-01-01

    The prototype SFR has six primary control rod assemblies(CRAs) and three secondary shutdown assemblies. The primary control system is used for power control, burnup compensation and reactor shutdown in response to demands from the plant control or protection systems. This paper describes the design concept of primary control rod drive mechanism shortly, and performs the parametric design studies for the electromagnet device of the drive mechanism to maximize CRA gripping force. The electromagnetic core usually confines and guides the magnetic field. The major parameters influenced on the electromagnetic force are the geometry and arrangement of the electromagnet and armature for a given coil specification. A typical equation calculating the electromagnetic force for a solenoid type is represented in equation. The first one is the increasing of the flux cross section area (Α c , Α g ) in magnetic field connecting of air gap, armature and electromagnets. Secondly, the reducing of the path lengths (l c , l g ) of the armature and electromagnet makes the magnetic flux (Β) resistance to be low. An electromagnet field analyses are performed for the initial design values of the electromagnet device. The gripping force is about 3 times of CRA weight when one coil is power on. The parametric studies on air gap, core sizes configuring of the electromagnet cores are performed to maximize the electromagnetic force

  20. A constitutive model for the flow through an assembly of circular section rods

    International Nuclear Information System (INIS)

    Moeller, S.V.

    1979-08-01

    The determination of the flow through an uniform array of rod bundle is made by means of the Continuum Theories of Mixtures, which gives balance equations for the system. The hypotheses of isothermal and fully developed turbulent flow are made. Constitutive equations for the resistive force are determined from Jakob's and Rowe's correlations, and its behaviour analysed for a standard case. Comparison of these equations with Bottgenbach's experiments shows good agreement of the direction of the pressure, although direct comparison between present theory and his theory is not possible. For the confirmation of the model an experiment is performed, this consisting of measuring pressure drop (Euler's Number) in the axial and transverse direction of a random array rod bundle at various angles as functions of velocity (Reynold's Number), which has good agreement, except on axial direction. At last, a sample problem is formulated with the purpose of showing the applicability of the model, this being the determination of pressure field due to the influence of a baffle. (Author) [pt

  1. High burnup fuel onset conditions in dry storage. Prediction of EOL rod internal pressure

    Energy Technology Data Exchange (ETDEWEB)

    Feria, F.; Herranz, L.E.

    2015-07-01

    During dry storage, cladding resistance to failure can be affected by several degrading mechanisms like creep or hydrides radial reorientation. The driving force of these effects is the stress at which the cladding is submitted. The maximum stress in the cladding is determined by the end-of-reactor-life (EOL) rod internal pressure, PEOL, at the maximum temperature attained during dry storage. Thus, PEOL sets the initial conditions of storage for potential time-dependent changes in the cladding. Based on FRAPCON-3.5 calculations, the aim of this work is to analyse the PEOL of a PWR fuel rod irradiated to burnups greater than 60 GWd/tU, where limited information is available. In order to be conservative, demanding irradiation histories have been used with a peak linear power of 44 kW/m. FRAPCON-3.5 results show an increasing exponential trend of PEOL with burnup, from which a simple correlation has been derived. The comparison with experimental data found in the literature confirms the enveloping nature of the predicted curve. Based on that, a conservative prediction of cladding stress in dry storage has been obtained. The comparison with a critical stress threshold related to hydrides embrittlement seems to point out that this issue should not be a concern at burnups below 65 GWd/tU. (Author)

  2. The improvement of control rod in experimental fast reactor JOYO. The development of a sodium bonded type control rod

    Energy Technology Data Exchange (ETDEWEB)

    Soga, T.; Miyakawa, S.; Mitsugi, T. [Japan Nuclear Cycle Development Inst., Oarai Engineering Center, Irradiation Center, Irradiation and Administration Section, Oarai, Ibaraki (Japan)

    1999-06-01

    Currently, the lifetime of control rods in JOYO is limited by Absorber-Cladding Mechanical Interaction (ACMI) due to swelling of B{sub 4}C(boron carbide) pellets accelerated by relocation of pellet fragments. A sodium bonded type control rod was developed which improves the thermal conductivity by means of charging sodium into the gap between B{sub 4}C and cladding and by utilizing a shroud which wraps the pellet fragments in a thin tube. This new design will be able to enlarge the gap between B{sub 4}C and cladding, without heating B{sub 4}C or fragment relocation, thus extending the life of the control rod. The sodium bonded type will be fabricated as the ninth reload control rods in JOYO. (1) The specification of a sodium bonded type control rod was determined with the wide gap between B{sub 4}C and cladding. In the design simulation, main component temperature were below the maximum limit. And the local heating by helium bubble generated from B{sub 4}C in the sodium gap, was not a serious problem in the analysis which was considered. (2) A structural design for the sodium entrance into the pin was determined. A formula was developed which the limit for sodium charging given physical dimension of the structure and sodium property. Result from sodium out-pile experiments validated the theoretical formula. (3) The analysis of ACMI indicated a lifetime extension of the sodium bonded type by 4.6% in comparison with lifetime of the helium bonded type of 1.6%. This is due to the boron10 burn-up rate being three times higher in the sodium bonded type than in the helium bonded type. To achieve a target burn-up 10% in the future, it will be necessary to modify design based on irradiation data which will be obtained by practical use of the sodium bonded control rods in JOYO. (4) The effects due to Absorber-Cladding Chemical Interaction (ACCI) were reduced by controlling the cladding temperature and chromium coating to the cladding's inner surface. It was confirmed

  3. Status of work on the final repository concept concerning direct disposal of spent fuel rods in fuel rod casks (BSK)

    International Nuclear Information System (INIS)

    Filbert, W.; Wehrmann, J.; Bollingerfehr, W.; Graf, R.; Fopp, S.

    2008-01-01

    The reference concept in Germany on direct final storage of spent fuel rods is the burial of POLLUX containers in the final repository salt dome. The POLLUX container is self-shielded. The final storage concept also includes un-shielded borehole storage of high-level waste and packages of compacted waste. GNS has developed a spent fuel container (BSK-3) for unshielded borehole storage with a mass of 5.2 tons that can carry the fuel rods of three PWR reactors of 9 BWR reactors. The advantages of BSK storage include space saving, faster storage processes, less requirements concerning technical barriers, cost savings for self-shielded casks.

  4. Cholesteric colloidal liquid crystals from phytosterol rod-like particles

    NARCIS (Netherlands)

    Rossi, L.; Sacanna, S.; Velikov, K.P.

    2011-01-01

    We report the first observation of chiral colloidal liquid crystals of rod-like particles from a low molecular weight organic compound— phytosterols. Based on the particles shape and crystal structure, we attribute this phenomenon to chiral distribution of surface charge on the surface of

  5. Rigid rod spaced fullerene as building block for nanoclusters

    Indian Academy of Sciences (India)

    By using phenylacetylene based rigid-rod linkers (PhA), we have successfully synthesized two fullerene derivatives, C60-PhA and C60-PhA-C60. The absorption spectral features of C60, as well as that of the phenylacetylene moiety are retained in the monomeric forms of these fullerene derivatives, ruling out the possibility ...

  6. Problem of radioactive lightning rods in the Republic of Croatia

    International Nuclear Information System (INIS)

    Novakovic, M.

    1994-01-01

    It became evident that as in most countries in Europe and other world, the radioactive lightning preventers will be prohibited in Croatia. It has to be done gradually and in phases. About 50% of whole number of radioactive lightning rods is mounted on hotels, and other are on industrial objects. Request for immediate replacement of them can almost fully load the available storage with radioactive waste, and the ex users should spent a significant sums of money to built an alternative lightning protection. One of the options is to use dismounted sources and use them for some other convenient purpose even for renewing the other radioactive lightning rod. In our opinion the best is to prohibit installation of the new lightning rods and existing ones dismount after elapsing the time for replacement of the radioactive attachment. After some years all radioactive lightning rods would be dismounted with smaller financial burden to ex users and community resulting also with less net amounts of radioactive waste

  7. Further delineation of spondylometaphyseal dysplasia with cone-rod dystrophy

    NARCIS (Netherlands)

    Sousa, Sérgio B.; Russell-Eggitt, Isabelle; Hall, Christine; Hall, Bryan D.; Hennekam, Raoul C. M.

    2008-01-01

    There are several entities that combine a skeletal dysplasia with a retinal dystrophy. Recently, another possibly autosomal recessive entity was added to this group characterized by a specific spondylometaphyseal dysplasia and a cone-rod dystrophy, without other significant impairments. The entity

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

    International Nuclear Information System (INIS)

    Sease, J.D.

    1998-01-01

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

  9. CEA data and methods for control rod calculations

    International Nuclear Information System (INIS)

    Salvatores, M.

    1988-01-01

    Methods and data used at CEA for LMFBR control rod calculations are presented. The performances of both the design methods and of the refined methods for the SUPER PHENIX start-up experiments are given. New developments for future core design and operation are also reported, together with the role of integral experiments to validate them. (author). 15 refs, 4 figs, 11 tabs

  10. Efficient Electroluminescence from a New Conjugated Rigid-Rod Polyquinoline

    National Research Council Canada - National Science Library

    Zhang, Xuejun, Ph.D

    1998-01-01

    A new conjugated rigid-rod polyquinoline, poly (2,2'-p-phenylene)-6,6'-bis(4-(p-tert-butylphenyl)quinoline)), was synthesized and its thin film was used as the emissive layer in light-emitting diodes...

  11. Hydraulic pressure control unit for control rod drive

    International Nuclear Information System (INIS)

    Watabe, Yukio.

    1990-01-01

    The pressure invention concerns a hydraulic pressure control unit for control rod drives in BWR type reactors. The space above a floating piston possessed by an accumulator and the housing of control rod drives are connected by means of a pipeline. The pipeline has a scram valve which is opened upon occurrence of reactor scram. A pump is disposed between the accumulator and the scram valve for communicating a discharge port to apply a high pressure water to the accumulator. According to the present invention, a control unit is disposed between the scram valve and the housing of the control rod drives in the hydraulic pressure control unit for maintaining the cross sectional area of the flow channel of the pipeline to a usual size when the pressure in a pressure vessel is under a rated operation pressure, while limiting the cross sectional area of the flow channel when the pressure is lower than that in the rated operation. Thus, whole insertion of the control rod substantially at a constant speed is enabled irrespective of the level of the pressure in the pressure vessel. (I.S.)

  12. Hemispheric Correlates of the Rod-And-Frame Test.

    Science.gov (United States)

    Berlin, Donna F.; Languis, Marlin L.

    1981-01-01

    Right-handed sixth graders were administered the WISC Block Design and verbal and nonverbal versions of the Rod-and-Frame Test (RFT), measuring field dependence/independence. Results seemed to reflect a right hemisphere processing for the nonverbal RFT and a possible sex bias against girls in its traditional verbal administration. (Author/SJL)

  13. 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

  14. Determination of the control rod worth for research reactors

    International Nuclear Information System (INIS)

    Aldama, D.L.; Gual, M.R.

    2000-01-01

    Nowadays there is a big interest in developing neutronic analysis methods for research reactor and particularly for the determination of the control rods worth under different operation conditions and core configurations. The reactivity associated with the control rods is of interest in the shutdown margin and in calculations of possible abnormal conditions related to reactivity accidents. For theses studies several computer codes have been developed. The present work is aimed at the validation of the calculation methods of the Nuclear Technology Center of Cuba. For this purpose, in order to evaluate the safety of this type of installations, the reactivity worth of the control rods of the cylindrical configuration of the Brazilian critical assembly IPEN/MB-01 is determined. These calculations, however, are a relatively complex task that requires the use of three-dimensional models. Because of this, the validation of the calculation methods used for this purpose is of great importance. In fact, it is one of the requirements called upon by the quality assurance programs for the development, maintenance and utilization of the calculation codes used in safety analysis. For the calculation of control rod worth the lattice code WIMS-D/4 [8] and the diffusion code SNAP-3D [9] were used. This work presents the obtained results and gives a comparison with the experimental values

  15. Safety rod/thimble melt failure characterization experiments

    International Nuclear Information System (INIS)

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

    1992-05-01

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

  16. Boiling water reactor radiation shielded Control Rod Drive Housing Supports

    Energy Technology Data Exchange (ETDEWEB)

    Baversten, B.; Linden, M.J. [ABB Combustion Engineering Nuclear Operations, Windsor, CT (United States)

    1995-03-01

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

  17. Non-linear waves in heterogeneous elastic rods via homogenization

    KAUST Repository

    Quezada de Luna, Manuel

    2012-03-01

    We consider the propagation of a planar loop on a heterogeneous elastic rod with a periodic microstructure consisting of two alternating homogeneous regions with different material properties. The analysis is carried out using a second-order homogenization theory based on a multiple scale asymptotic expansion. © 2011 Elsevier Ltd. All rights reserved.

  18. Phase behaviour of rod-like colloid + flexible polymer mixtures

    NARCIS (Netherlands)

    Lekkerkerker, H.N.W.; Stroobants, A.

    The effect of non-adsorbing, flexible polymer on the isotropic-nematic transition in dispersions of rod-like colloids is investigated. A widening of the biphasic gap is observed, in combination with a marked polymer partitioning between the coexisting phases. Under certain conditions, areas of

  19. Longitudinal vibrations of a Rayleigh-Bishop rod

    CSIR Research Space (South Africa)

    Fedotov, IA

    2010-01-01

    Full Text Available In this work, for analyzing the longitudinal vibrations of a conic rod, the authors used the Rayleigh–Bishop model, which generalizes the Rayleigh model and takes into account both lateral displacements and the shear stress in the transverse cross...

  20. Measurement of safety rod trajectory using digital optical encoder

    International Nuclear Information System (INIS)

    Milovanovic, S.; Pesic, M.; Milovanovic, T.

    1996-01-01

    In this paper a rod drop position - time function z(t) is measured using new method - with a digital displacement transducer and digital optical incremental encoder connected to a serial communication port of a personal computer. Some improvements, comparing to the previously proposed method are explained. Also, some possible applications in analysis of reactor dynamics and kinetics are proposed. (author)

  1. 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.

  2. Possibilities and limits of the reactivity determination of control rods

    International Nuclear Information System (INIS)

    Buenemann, D.

    1975-01-01

    Basic physical facts of the reactivity determination of control rods are presented. A survey of currrently applied methods is given, and the drawbacks of the various methods are pointed out. Special problems are presented by the interpretation of highly subcritical assemblies which are not really important in practical reactor operation but desirable for a consistant comparison between theory and experiments. (orig./AK) [de

  3. 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

  4. Electromagnetic design calculation of the control rod drive mechanism

    International Nuclear Information System (INIS)

    Zhu Qirong; Zhu Jingchang

    1991-01-01

    Electromagnetic design calculation of the step-by-step magnetic jacking control rod drive mechanism includes magnetic field force calculation and design calculation of magnetomotive force for three electromagnetic iron and their coilds. The basic principle and method of electromagnetic design calculation had been expounded to take the lift magnet and lift coil for example

  5. Effect of Cuisenaire Rods' approach on students' interest in decimal ...

    African Journals Online (AJOL)

    This study explored the effectiveness of Cuisenaire Rods' approach in arousing students' interest in decimal fractions. Two research questions were posed and three hypotheses were formulated to guide the study. A sample of 200 JS3 students from a randomly selected schools in Makurdi Metropolis of Benue State, served ...

  6. Cross-correlated imaging of distributed mode filtering rod fiber

    DEFF Research Database (Denmark)

    Laurila, Marko; Barankov, Roman; Jørgensen, Mette Marie

    2013-01-01

    We analyze the modal properties of an 85μm core distributed mode filtering rod fiber using cross-correlated (C2) imaging. We evaluate suppression of higher-order modes (HOMs) under severely misaligned mode excitation and identify a single-mode regime where HOMs are suppressed by more than 20dB....

  7. Frequency resolved transverse mode instability in rod fiber amplifiers

    DEFF Research Database (Denmark)

    Johansen, Mette Marie; Laurila, Marko; Maack, Martin D.

    2013-01-01

    Frequency dynamics of transverse mode instabilities (TMIs) are investigated by testing three 285/100 rod fibers in a single-pass amplifier setup reaching up to ~200W of extracted output power without beam instabilities. The pump power is increased well above the TMI threshold to uncover output dy...

  8. Flexible Stabilisation of the Degenerative Lumbar Spine Using PEEK Rods

    Directory of Open Access Journals (Sweden)

    Jacques Benezech

    2016-01-01

    Full Text Available Posterior lumbar interbody fusion using cages, titanium rods, and pedicle screws is considered today as the gold standard of surgical treatment of lumbar degenerative disease and has produced satisfying long-term fusion rates. However this rigid material could change the physiological distribution of load at the instrumental and adjacent segments, a main cause of implant failure and adjacent segment disease, responsible for a high rate of further surgery in the following years. More recently, semirigid instrumentation systems using rods made of polyetheretherketone (PEEK have been introduced. This clinical study of 21 patients focuses on the clinical and radiological outcomes of patients with lumbar degenerative disease treated with Initial VEOS PEEK®-Optima system (Innov’Spine, France composed of rods made from PEEK-OPTIMA® polymer (Invibio Biomaterial Solutions, UK without arthrodesis. With an average follow-up of 2 years and half, the chances of reoperation were significantly reduced (4.8%, quality of life was improved (ODI = 16%, and the adjacent disc was preserved in more than 70% of cases. Based on these results, combined with the biomechanical and clinical data already published, PEEK rods systems can be considered as a safe and effective alternative solution to rigid ones.

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

    Energy Technology Data Exchange (ETDEWEB)

    Sease, J.D.

    1998-03-01

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

  10. Local heat transfer coefficient for turbulent flow in rod bundles

    International Nuclear Information System (INIS)

    Fernandez y Fernandez, E.; Carajilescov, P.

    1983-03-01

    The correlation of the local heat transfer coefficients in heated triangular array of rod bundles, in terms of the flow hydrodynamic parameters is presented. The analysis is made first for fluid with Prandtl numbers varying from moderated to high (Pr>0.2), and then extended to fluids with low Prandtl numbers (0.004 [pt

  11. Mechanical behaviour of PWR fuel rods during intermediate storage

    International Nuclear Information System (INIS)

    Bouffioux, P.; Dalmas, R.; Bernaudat, C.

    2000-01-01

    EDF, which owns the irradiated fuel coming from its NPPs, has initiated studies regarding the mechanical behaviour of a fuel rod and the integrity of its cladding, in the case where the spent fuel is stored for a significant duration. During the phases following in-reactor irradiation (ageing in a water-pool, transport and intermediate storage), many phenomena, which are strongly coupled, may influence the cladding integrity: - residual power and temperature decay; - helium production and release in the free volume of the rod (especially for MOX fuel); - fuel column swelling; - cladding creep-out under the inner gas pressure of the fuel rod; - metallurgical changes due to high temperatures during transportation. In parallel, the quantification of the radiological risk is based on the definition of a cladding integrity criterion. Up to now, this criterion requires that the clad hoop strain due to creep-out does not exceed 1%. A more accurate criterion is being investigated. The study and modelling of all the phenomena mentioned above are included in a R and D programme. This programme also aims at redefining the cladding integrity criterion, which is assumed to be too conservative. The R and D programme will be presented. In order to predict the overall behaviour of the rod during the intermediate storage phases, the AVACYC code has been developed. It includes the models developed in the R and D programme. The input data of the AVACYC code are provided by the results of in-reactor rod behaviour simulations, using the thermal-mechanical CYRANO3 code. Its main results are the evolution vs. time of hoop stresses in the cladding, rod internal pressure and cladding hoop strains. Chained CYRANO-AVACYC calculations have been used to simulate the behaviour of MOX fuel rods irradiated up to 40 GWd/t and stored under air during 100 years, or under water during 50 years. For such fuels, where the residual power remains high, we show that a large part of the cladding strain

  12. Coolant mixing in LMFBR rod bundles and outlet plenum mixing transients. Progress report, September 1, 1976--November 30, 1976

    Energy Technology Data Exchange (ETDEWEB)

    Todreas, N.E.; Golay, M.W.; Wolf, L.

    1976-01-01

    Information is presented concerning bundle geometry with wrapped and bare rods, subchannel geometry with bare rods, LMFBR outlet plenum flow mixing, and theoretical determination of local temperature fields in LMFBR fuel rod bundles.

  13. Coolant mixing in LMFBR rod bundles and outlet plenum mixing transients. Progress report, March 1, 1977--May 31, 1977

    International Nuclear Information System (INIS)

    Todreas, N.E.; Golay, M.W.; Wolf, L.

    1977-01-01

    Progress is summarized in the following tasks: (1) bundle flow studies (wrapped and bare rods); (2) subchannel flow studies (bare rods); (3) LMFBR outlet plenum flow mixing; and (4) theoretical determination of local temperature fields in LMFBR fuel rod bundles

  14. Effect of severity of rod contour on posterior rod failure in the setting of lumbar pedicle subtraction osteotomy (PSO): a biomechanical study.

    Science.gov (United States)

    Tang, Jessica A; Leasure, Jeremi M; Smith, Justin S; Buckley, Jenni M; Kondrashov, Dimitriy; Ames, Christopher P

    2013-02-01

    Rod failure has been reported clinically in pedicle subtraction osteotomy (PSO) to correct flat back deformity. To characterize the fatigue life of posterior screw-rod constructs in the setting of PSO as a function of the severity of rod contour angle. A modified ASTM F1717 to 04 was used. Rods were contoured to the appropriate angle for the equivalent 20-, 40-, or 60-degree PSO angles. Testing was performed on a mechanical test frame at 400/40 N and 250/25 N, and specimens were cycled at 4 Hz to failure or run-out at 2,000,000 cycles. The effect of the screw-rod system on fatigue strength of curved rods was compared using Cox proportional hazards regression. At 400 N/40 N, Cox proportional hazards regression indicated that contouring rods from a 20-degree PSO angle to either 40 or 60 degrees significantly decreased fatigue life (hazard ratio = 7863.6, P = .0144). However, contouring rods from a 40-degree PSO angle to 60 degrees had no significant effect on the fatigue life (P > .05). At 250 N/25 N, Cox proportional hazards regression indicated that contouring rods from a 20-degree PSO angle to either 40 or 60 degrees significantly decreased fatigue life (hazard ratio = 7863.6, P = .0144). Furthermore, contouring rods from a 40-degree PSO angle to 60 degrees had a significant effect on the fatigue life (hazard ratio = 7863.6, P = .0144). Results suggest that in the setting of PSO, the fatigue life of posterior spinal fixation rods depends largely on the severity of the rod angle used to maintain the vertebral angle created by the PSO and is significantly lowered by rod contouring.

  15. Burnout data for flow of boiling water in vertical round ducts, annuli and rod clusters

    Energy Technology Data Exchange (ETDEWEB)

    Becker, Kurt M.; Hernborg, Gunnar; Bode, Manfred; Eriksson, O.

    1965-07-01

    The present report contains the tables of the burnout data obtained for flow in vertical channels at the Heat Engineering Laboratory of AB Atomenergi in Sweden. The data covers measurements in round ducts, annuli, 3-rod and 7-rod clusters.

  16. Design of water rod cores of a direct cycle supercritical-pressure light water reactor

    International Nuclear Information System (INIS)

    Okano, Yasushi; Koshizuka, Sei-Ichi; Oka, Yoshiaki

    1994-01-01

    A conceptual design of a direct-cycle supercritical-pressure light water reactor with water rods is presented. Three types of water rods are analyzed: single, semi-double and full-double tubes. A water rod replaces seven fuel rods in a triangular lattice. The coolant density change in the water rods and the fuel channel is calculated using a code developed in the present study. The full double tube is most superior in terms of the distribution of the moderator. The number of fuel rods to water rods is 198:19, which makes optimum moderation. The average enrichment becomes 4.13%. The axial power flattening is finally achieved by partial length fuel rods and enrichment split of 0.25%. The discharge burnup is 45 GWd/t. (author)

  17. Burnout data for flow of boiling water in vertical round ducts, annuli and rod clusters

    International Nuclear Information System (INIS)

    Becker, Kurt M.; Hernborg, Gunnar; Bode, Manfred; Eriksson, O.

    1965-01-01

    The present report contains the tables of the burnout data obtained for flow in vertical channels at the Heat Engineering Laboratory of AB Atomenergi in Sweden. The data covers measurements in round ducts, annuli, 3-rod and 7-rod clusters

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

    Science.gov (United States)

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

    2016-03-01

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

  19. COST IMPACT OF ROD CONSOLIDATION ON THE VIABILITY ASSESSMENT DESIGN

    Energy Technology Data Exchange (ETDEWEB)

    D. Lancaster

    1999-03-29

    The cost impact to the Civilian Radioactive Waste Management System of using rod consolidation is evaluated. Previous work has demonstrated that the fuel rods of two assemblies can be packed into a canister that can fit into the same size space as that used to store a single assembly. The remaining fuel assembly hardware can be compacted into the same size canisters with a ratio of 1 hardware canister per each 6 to 12 assemblies. Transportation casks of the same size as currently available can load twice the number of assemblies by placing the compacted assemblies in the slots currently designed for a single assembly. Waste packages similarly could contain twice the number of assemblies; however, thermal constraints would require considering either a low burnup or cooling. The analysis evaluates the impact of rod consolidation on CRWMS costs for consolidation at prior to transportation and for consolidation at the Monitored Geological Repository surface facility. For this study, no design changes were made to either the transport casks or waste packages. Waste package designs used for the Viability Assessment design were employed but derated to make the thermal limits. A logistics analysis of the waste was performed to determine the number of each waste package with each loading. A review of past rod consolidation experience found cost estimates which range from $10/kgU to $32/kgU. $30/kgU was assumed for rod consolidation costs prior to transportation. Transportation cost savings are about $17/kgU and waste package cost savings are about $21/kgU. The net saving to the system is approximately $500 million if the consolidation is performed prior to transportation. If consolidation were performed at the repository surface facilities, it would cost approximately $15/kgU. No transportation savings would be realized. The net savings for consolidation at the repository site would be about $400 million dollars.

  20. p53 selectively regulates developmental apoptosis of rod photoreceptors.

    Directory of Open Access Journals (Sweden)

    Linda Vuong

    Full Text Available Retinal cells become post-mitotic early during post-natal development. It is likely that p53, a well-known cell cycle regulator, is involved in regulating the genesis, differentiation and death of retinal cells. Furthermore, retinal cells are under constant oxidative stress that can result in DNA damage, due to the extremely high level of metabolic activity associated with phototransduction. If not repaired, this damage may result in p53-dependent cell death and ensuing vision loss. In this study, the role of p53 during retinal development and in the post-mitotic retina is investigated. A previously described super p53 transgenic mouse that expresses an extra copy of the mouse p53 gene driven by its endogenous promoter is utilized. Another transgenic mouse (HIP that expresses the p53 gene in rod and cone photoreceptors driven by the human interphotoreceptor retinoid binding protein promoter was generated. The electroretinogram (ERG of the super p53 mouse exhibited reduced rod-driven scotopic a and b wave and cone-driven photopic b wave responses. This deficit resulted from a reduced number of rod photoreceptors and inner nuclear layer cells. However, the reduced photopic signal arose only from lost inner retinal neurons, as cone numbers did not change. Furthermore, cell loss was non-progressive and resulted from increased apoptosis during retinal developmental as determined by TUNEL staining. In contrast, the continuous and specific expression of p53 in rod and cone photoreceptors in the mature retinas of HIP mice led to the selective loss of both rods and cones. These findings strongly support a role for p53 in regulating developmental apoptosis in the retina and suggest a potential role, either direct or indirect, for p53 in the degenerative photoreceptor loss associated with human blinding disorders.

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

    Science.gov (United States)

    Smith, J.E.

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

  2. 49 CFR 230.96 - Main, side, and valve motion rods.

    Science.gov (United States)

    2010-10-01

    ... bore of main rod bearings shall not exceed pin diameters more than 3/32 inch at front or back end. The total lost motion at both ends shall not exceed 5/32 inch. (g) Side rod bearings. The bore of side rod.... The total amount of side motion of each rod on its crank pin shall not exceed 1/4 inch. (e) Oil and...

  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

    broth. Protein bands of expected sizes were detected by SDS-PAGE and western blotting, and the identity was further confirmed by tandem mass spectrometry. Both proteins were purified using his-affinity chromatography, and the high level of purity was verified by silver-stained SDS-PAGE. Recombinant Rod...

  4. Operational experience gained with the Failed Fuel Rod Detection System in nuclear power plants

    International Nuclear Information System (INIS)

    Boehm, H.H.; Foerch, H.

    1985-01-01

    Fuel assemblies containing defective fuel rods are releasing fission products, and consequently have to be removed from further service in the core. Partially spent fuel assemblies can only be reinserted into the core after removal of the defective rods. Spent fuel assemblies have to be freed from these failed rods before being shipped to a reprocessing plant

  5. 78 FR 71565 - Steel Threaded Rod from India: Postponement of Preliminary Determination of Antidumping Duty...

    Science.gov (United States)

    2013-11-29

    ... International Trade Administration Steel Threaded Rod from India: Postponement of Preliminary Determination of...'') published a notice of initiation of the antidumping duty investigation of steel threaded rod from India.\\1... later than December 20, 2013. \\1\\ See Steel Threaded Rod From India and Thailand: Initiation of...

  6. Leader inception field from a vertical rod conductor efficiency of electrical triggering techniques

    Energy Technology Data Exchange (ETDEWEB)

    Berger, G. [CNRS-SUPELEC, Gif sur Yvette (France)

    1996-12-31

    Vertical rod conductors have been tested in high voltage laboratory under simulated lightning conditions. These experiments led to the determination of the inception field necessary to launch an upward positive leader, which is a function of the rod height, confirming Rizk`s models. The efficiency of electrical triggering techniques added to a single rod is also investigated.

  7. Results of automatic system implementation for the friction control rods execution in Cofrentes nuclear power plant

    International Nuclear Information System (INIS)

    Palomo, M.; Urrea, M.; Arnaldos, A.

    2011-01-01

    The purpose of this presentation is to show the obtained results in Cofrentes Nuclear Power Plant (Spain) of Control Rods PCC/24 Friction Test Procedure. In order to perform this, a Control Rod Friction Test System has been developed. Principally, this system consists on software and data acquisition hardware that obtains and analyzes the control rod pressure variation on which the test is being made. The PCC/24 Procedure objective is to detect an excessive friction in the control rod movement that could cause a CRD (Control Rod Drive) movement slower than usual. This test is necessary every time that an anomalous alteration is produced in the reactor core that could affect to a fuel rod, and it is executed before the time measure of control rods rapid scram test of the affected rods. This test has to be carried out to all the reactor control rods and takes valuable time during plant refuelling. So, by means of an automatic system to perform the test, we obtain an important time saving during refuelling. On the other hand, the on-line monitoring of the control rod insertion and changes in differential pressure, permits a control rod operation fast and safe validation. Moreover, an automatic individual report of every rod is generated by the system and a final global result report of the entire test developed in refuelling is generated. The mentioned reports can be attached directly to the procedure documents obtaining an office data processing important saving time.(author)

  8. Results of automatic system implementation for the friction control rods execution in Cofrentes nuclear power plant

    International Nuclear Information System (INIS)

    Curiel, M.; Palomo, M. J.; Urrea, M.; Arnaldos, A.

    2010-10-01

    The purpose of this presentation is to show the obtained results in Cofrentes nuclear power plant (Spain) of control rods Pcc/24 friction test procedure. In order to perform this, a control rod friction test system has been developed. Principally, this system consists on software and data acquisition hardware that obtains and analyzes the control rod pressure variation on which the test is being made. The Pcc/24 procedure objective is to detect an excessive friction in the control rod movement that could cause a control rod drive movement slower than usual. This test is necessary every time that an anomalous alteration is produced in the reactor core that could affect to a fuel rod, and it is executed before the time measure of control rods rapid scram test of the affected rods. This test has to be carried out to all the reactor control rods and takes valuable time during plant refuelling. So, by means of an automatic system to perform the test, we obtain an important time saving during refuelling. On the other hand, the on-line monitoring of the control rod insertion and changes in differential pressure, permits a control rod operation fast and safe validation. Moreover, an automatic individual report of every rod is generated by the system and a final global result report of the entire test developed in refuelling is generated. The mentioned reports can be attached directly to the procedure documents obtaining an office data processing important saving time. (Author)

  9. 76 FR 78882 - Carbon and Certain Alloy Steel Wire Rod From Mexico: Affirmative Preliminary Determination of...

    Science.gov (United States)

    2011-12-20

    ... Steel Wire Rod From Mexico: Affirmative Preliminary Determination of Circumvention of the Antidumping... circumvention inquiry into whether Deacero S.A. de C.V. (Deacero) and Ternium Mexico S.A. de C.V. (Ternium... wire rod are covered by this circumvention inquiry. \\3\\ See Carbon and Certain Alloy Steel Wire Rod...

  10. Crystallization kinetics and morphology in phase separating and sedimenting mixtures of colloidal spheres and rods

    NARCIS (Netherlands)

    Lekkerkerker, H.N.W.; Oversteegen, S.M.; Wijnhoven, J.E.G.J.; Vonk, C.

    2004-01-01

    The crystallization of sedimentating silica spheres in the presence of silica-coated boehmite rods in low-salt dimethylformamide is studied by means of confocal scanning laser microscopy. As expected, addition of rods gives rise to a net attraction due to the depletion effect. Upon increasing rod

  11. Results of automatic system implementation for the friction control rods execution in Cofrentes nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Curiel, M. [Logistica y Acondicionamientos Industriales SAU, Sorolla Center, local 10, Av. de las Cortes Valencianas, 46015 Valencia (Spain); Palomo, M. J. [ISIRYM, Universidad Politecnica de Valencia, Camino de Vera s/n, Valencia (Spain); Urrea, M. [Iberdrola Generacion S. A., Central Nuclear Cofrentes, Carretera Almansa Requena s/n, 04662 Cofrentes, Valencia (Spain); Arnaldos, A., E-mail: m.curiel@lainsa.co [TITANIA Servicios Tecnologicos SL, Sorolla Center, local 10, Av. de las Cortes Valencianas No. 58, 46015 Valencia (Spain)

    2010-10-15

    The purpose of this presentation is to show the obtained results in Cofrentes nuclear power plant (Spain) of control rods Pcc/24 friction test procedure. In order to perform this, a control rod friction test system has been developed. Principally, this system consists on software and data acquisition hardware that obtains and analyzes the control rod pressure variation on which the test is being made. The Pcc/24 procedure objective is to detect an excessive friction in the control rod movement that could cause a control rod drive movement slower than usual. This test is necessary every time that an anomalous alteration is produced in the reactor core that could affect to a fuel rod, and it is executed before the time measure of control rods rapid scram test of the affected rods. This test has to be carried out to all the reactor control rods and takes valuable time during plant refuelling. So, by means of an automatic system to perform the test, we obtain an important time saving during refuelling. On the other hand, the on-line monitoring of the control rod insertion and changes in differential pressure, permits a control rod operation fast and safe validation. Moreover, an automatic individual report of every rod is generated by the system and a final global result report of the entire test developed in refuelling is generated. The mentioned reports can be attached directly to the procedure documents obtaining an office data processing important saving time. (Author)

  12. 78 FR 66382 - Certain Steel Threaded Rod From India and Thailand

    Science.gov (United States)

    2013-11-05

    ...)] Certain Steel Threaded Rod From India and Thailand Determinations On the basis of the record \\1\\ developed... injured by reason of imports from India and Thailand of certain steel threaded rod, provided for primarily... threaded rod from Thailand. Accordingly, effective June 27, 2013, the Commission instituted countervailing...

  13. Results of automatic system implementation for the friction control rods execution in Cofrentes nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Palomo, M., E-mail: mpalomo@iqn.upv.es [Universidad Politecnica de Valencia (UPV) (Spain); Urrea, M., E-mail: matias.urrea@iberdrola.es [Iberdrola Generacion S.A. Valencia (Spain). C.N. Cofrentes; Curiel, M., E-mail: m.curiel@lainsa.com [Logistica y Acondicionamientos Industriales (LAINSA), Valencia (Spain); Arnaldos, A., E-mail: a.arnaldos@titaniast.com [TITANIA Servicios Teconologicos, Valencia (Spain)

    2011-07-01

    The purpose of this presentation is to show the obtained results in Cofrentes Nuclear Power Plant (Spain) of Control Rods PCC/24 Friction Test Procedure. In order to perform this, a Control Rod Friction Test System has been developed. Principally, this system consists on software and data acquisition hardware that obtains and analyzes the control rod pressure variation on which the test is being made. The PCC/24 Procedure objective is to detect an excessive friction in the control rod movement that could cause a CRD (Control Rod Drive) movement slower than usual. This test is necessary every time that an anomalous alteration is produced in the reactor core that could affect to a fuel rod, and it is executed before the time measure of control rods rapid scram test of the affected rods. This test has to be carried out to all the reactor control rods and takes valuable time during plant refuelling. So, by means of an automatic system to perform the test, we obtain an important time saving during refuelling. On the other hand, the on-line monitoring of the control rod insertion and changes in differential pressure, permits a control rod operation fast and safe validation. Moreover, an automatic individual report of every rod is generated by the system and a final global result report of the entire test developed in refuelling is generated. The mentioned reports can be attached directly to the procedure documents obtaining an office data processing important saving time.(author)

  14. SIVAR - Computer code for simulation of fuel rod behavior in PWR during fast transients

    International Nuclear Information System (INIS)

    Dias, A.F.V.

    1980-10-01

    Fuel rod behavior during a stationary and a transitory operation, is studied. A computer code aiming at simulating PWR type rods, was developed; however, it can be adapted for simulating other type of rods. A finite difference method was used. (E.G.) [pt

  15. Investigation of the group growing process for monocrystalline germanium rods by Stepanov's method in a rectilinear thermal zone

    CERN Document Server

    Egorov, L P; Zatulovskii, L M; Chaikin, P M; Gulyaev, Y V; Zhvirblyanskii, V Yu; Levinzon, D I; Smirnov, Yu M; Sachkov, G V

    1973-01-01

    The apparatus used had a floating former for the stable growth of 7.9 germanium rods of 8.9 mm. dia. from a crucible charge of approximately 3 kg. The control of the crystal growth front is discussed in experimental and theoretical terms. It is claimed that inspection of the crystal front during growth, made possible in this apparatus by the provision of ports in the heater screens, greatly facilitates control and increases the quality (dislocation density, specific resistance etc.) of the product. (1 refs).

  16. Fabrication and characterization of rod-like nano-hydroxyapatite on MAO coating supported on Mg-Zn-Ca alloy

    Science.gov (United States)

    Gao, J. H.; Guan, S. K.; Chen, J.; Wang, L. G.; Zhu, S. J.; Hu, J. H.; Ren, Z. W.

    2011-01-01

    The poor corrosion resistance of magnesium alloys is a dominant problem that limits their clinical application. In order to solve this challenge, micro-arc oxidation (MAO) was used to fabricate a porous coating on magnesium alloys and then electrochemical deposition (ED) was done to fabricate rod-like nano-hydroxyapatite (RNHA) on MAO coating. The cross-section morphology of the composite coatings and its corresponding energy dispersion spectroscopy (EDS) surficial scanning map of calcium revealed that HA rods were successfully deposited into the pores. The three dimensional morphology and scanning electron microscopy (SEM) image of the composite coatings showed that the distribution of the HA rods was dense and uniform. Atomic force microscope (AFM) observation of the composite coatings showed that the diameters of HA rods varied from 95 nm to 116 nm and the root mean square roughness (RMS) of the composite coatings was about 42 nm, which were favorable for cellular survival. The bonding strength between the HA film and MAO coating increased to 12.3 MPa, almost two times higher than that of the direct electrochemical deposition coating (6.3 MPa). Compared with that of the substrate, the corrosion potential of Mg-Zn-Ca alloy with composite coatings increased by 161 mV and its corrosion current density decreased from 3.36 × 10 -4 A/cm 2 to 2.40 × 10 -7 A/cm 2 which was due to the enhancement of bonding strength and the deposition of RNHA in the MAO pores. Immersion tests were carried out at 36.5 ± 0.5 °C in simulated body fluid (SBF). It was found that RNHA can induce the rapid precipitation of calcium orthophosphates in comparison with conventional HA coatings. Thus magnesium alloy coated with the composite coatings is a promising candidate as biodegradable bone implants.

  17. Fabrication and characterization of rod-like nano-hydroxyapatite on MAO coating supported on Mg-Zn-Ca alloy

    International Nuclear Information System (INIS)

    Gao, J.H.; Guan, S.K.; Chen, J.; Wang, L.G.; Zhu, S.J.; Hu, J.H.; Ren, Z.W.

    2011-01-01

    The poor corrosion resistance of magnesium alloys is a dominant problem that limits their clinical application. In order to solve this challenge, micro-arc oxidation (MAO) was used to fabricate a porous coating on magnesium alloys and then electrochemical deposition (ED) was done to fabricate rod-like nano-hydroxyapatite (RNHA) on MAO coating. The cross-section morphology of the composite coatings and its corresponding energy dispersion spectroscopy (EDS) surficial scanning map of calcium revealed that HA rods were successfully deposited into the pores. The three dimensional morphology and scanning electron microscopy (SEM) image of the composite coatings showed that the distribution of the HA rods was dense and uniform. Atomic force microscope (AFM) observation of the composite coatings showed that the diameters of HA rods varied from 95 nm to 116 nm and the root mean square roughness (RMS) of the composite coatings was about 42 nm, which were favorable for cellular survival. The bonding strength between the HA film and MAO coating increased to 12.3 MPa, almost two times higher than that of the direct electrochemical deposition coating (6.3 MPa). Compared with that of the substrate, the corrosion potential of Mg-Zn-Ca alloy with composite coatings increased by 161 mV and its corrosion current density decreased from 3.36 x 10 -4 A/cm 2 to 2.40 x 10 -7 A/cm 2 which was due to the enhancement of bonding strength and the deposition of RNHA in the MAO pores. Immersion tests were carried out at 36.5 ± 0.5 deg. C in simulated body fluid (SBF). It was found that RNHA can induce the rapid precipitation of calcium orthophosphates in comparison with conventional HA coatings. Thus magnesium alloy coated with the composite coatings is a promising candidate as biodegradable bone implants.

  18. 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

  19. Fabrication Of Bar graph Assembly For Control Rod Indicator Of Reactor Kartini

    International Nuclear Information System (INIS)

    Supriyanto

    1996-01-01

    The Bar graph instrument for monitor control rod position have been made. This instrument is located on the panel diagram in the control room of reactor Kartini. In this instrument, every control rod position is indicated by 10 LED on, that can move up or down as moving of control rod, total 20 LED is needed one assembly that is 10 LED for full down position and 10 LED for full up position. This bar graph can only monitor the control rod position every 10% of maximum control rod movement

  20. Modal properties of the flexural vibrating package of rods linked by spacer grids

    Directory of Open Access Journals (Sweden)

    Zeman V.

    2011-06-01

    Full Text Available The paper deals with the modelling and modal analysis of the large package of identical parallel rods linked by transverse springs (spacer grids placed on several level spacings. The rod discretization by finite element method is based on Rayleigh beam theory. For the cyclic and central symmetric package of rods (such as fuel rods in nuclear fuel assembly the system decomposition on the identical revolved rod segments was applied. A modal synthesis method with condensation is used for modelling of the whole system. The presented method is the first step for modelling the nuclear fuel assembly vibration caused by excitation determined by the support plate motion of the reactor core.

  1. Countercurrent flow-limiting characteristics of a Savannah River Plant control rod septifoil

    International Nuclear Information System (INIS)

    Anderson, J.L.

    1992-07-01

    Experiments were performed at the Idaho National Engineering Laboratory to investigate the counter-current flow limiting characteristics of a Savannah River Plant control rod septifoil assembly. These experiments were unheated, using air and water as the working fluids. Results are presented in terms of the Wallis flooding correlation for several different control rod configurations. Flooding was observed to occur in the vicinity of the inlet slots/holes of the septifoil, rather than within the rod bundle at the location of the minimum flow area. Nearly identical flooding characteristics of the septifoil were observed for configurations with zero, three, and four rods inserted, but significantly different results occurred with 5 rods inserted

  2. Absorber rod for nuclear reactors in a pebble bed of spherical operating elements

    International Nuclear Information System (INIS)

    Reinstein, D.; Gnutzmann, H.

    1978-01-01

    The claim refers to the constructional configuration of an absorber rod, whose and penetrating into the pebble bed has an opening to reduce the fracture rate, so that the operating elements can escape into a channel within the absorber rod. To suit this to the direction of movement of the elements a part of the end of the rod is flexibly connected to the hollow absorber rod via a joint. In this way the mechanical load of the element particles is reduced and simultaneously one achieves that much lower force is required to insert the absorber rod into the pebble bed. (UA) [de

  3. Experimental Investigation of Surface Wave Plasma Excited by a Cylindrical Dielectric Rod

    International Nuclear Information System (INIS)

    Wu Zhonghang; Li Zebin; He Kongduo; Yang Xilu; Chen Zhenliu; Ou Qiongrong; Liang Rongqing; Ju Jiaqi; Yan Hang

    2014-01-01

    An improved surface wave plasma source equipped with a cylindrical quartz rod has been developed, which has great potential in processing inner wall of cylindrical workpieces. A cylindrical quartz rod not only excites the plasma around the rod, but also guides surface wave plasma along the rod. The distributions of plasma density and plasma temperature under different incident microwave powers and pressures are diagnosed by a Langmuir probe. The electron density near the rod is around the order of 10 11 cm −3 . When the incident power is 450 W, the length of surface wave plasma column can reach up to 420 mm at 20 Pa. (low temperature plasma)

  4. Planar dynamics of large-deformation rods under moving loads

    Science.gov (United States)

    Zhao, X. W.; van der Heijden, G. H. M.

    2018-01-01

    We formulate the problem of a slender structure (a rod) undergoing large deformation under the action of a moving mass or load motivated by inspection robots crawling along bridge cables or high-voltage power lines. The rod is described by means of geometrically exact Cosserat theory which allows for arbitrary planar flexural, extensional and shear deformations. The equations of motion are discretised using the generalised-α method. The formulation is shown to handle the discontinuities of the problem well. Application of the method to a cable and an arch problem reveals interesting nonlinear phenomena. For the cable problem we find that large deformations have a resonance detuning effect on cable dynamics. The problem also offers a compelling illustration of the Timoshenko paradox. For the arch problem we find a stabilising (delay) effect on the in-plane collapse of the arch, with failure suppressed entirely at sufficiently high speed.

  5. Theoretical rheology of suspensions of ferromagnetic rod-like particles

    International Nuclear Information System (INIS)

    Altenberger, A.R.; Dahler, J.S.

    1989-01-01

    The authors extend the linear response-like derivation of the generalized Navier-Stokes equation to non-Newtonian flows with rate of strain-dependent transport coefficients. They derive a time correlation function expression for the viscosity tensor and point out possible ambiguities in the operational definitions of viscosity coefficients. Their analysis is specific to a suspension of polar, rod-like ferromagnetic particles. A commentary is included about the approximations that lead from the time correlation function and the molecular definition of the viscosity tensor to the standard, Brownian dynamics model used in the theoretical rheology of suspensions. Some theoretical difficulties and logical inconsistencies are pointed out. Preliminary results for the transport coefficients of dilute suspension of magnetic rod-like particles are presented

  6. Control-rod removal-blocking monitor (RBM)

    International Nuclear Information System (INIS)

    Matsumiya, Shoichi.

    1983-01-01

    Purpose: To provide an apparatus capable of checking with ease whether predetermined local power region neutron signals are selected correctly depending on selected control rods. Constitution: A computer is adapted to read the entire LPRM (local power range monitor) signals and a matrix calculation function is further provided for obtaining predetermined selected LPRM signals from the entire LPRM signals by the selected control rods. Then, a comparison function is provided for comparing the above mentioned selected LPRM signals read from RBM and signals obtained in the inside of the computer. The comparison functions judge whether both of them coincide to each other or not and, if it is judged not coincide, produces an abnormal-operation signal. Since the operator can thus recognize the failure in the RBM and carry out repairement directly, the reliability can be improved. (Seki, T.)

  7. Zircaloy sheathed thermocouples for PWR fuel rod temperature measurements

    International Nuclear Information System (INIS)

    Anderson, J.V.; Wesley, R.D.; Wilkins, S.C.

    1979-01-01

    Small diameter zircaloy sheathed thermocouples have been developed by EG and G Idaho, Inc., at the Idaho National Engineering Laboratory. Surface mounted thermocouples were developed to measure the temperature of zircaloy clad fuel rods used in the Thermal Fuels Behavior Program (TFBP), and embedded thermocouples were developed for use by the Loss-of-Fluid Test (LOFT) Program for support tests using zircaloy clad electrically heated nuclear fuel rod simulators. The first objective of this developmental effort was to produce zircaloy sheathed thermocouples to replace titanium sheathed thermocouples and thereby eliminate the long-term corrosion of the titanium-to-zircaloy attachment weld. The second objective was to reduce the sheath diameter to obtain faster thermal response and minimize cladding temperature disturbance due to thermocouple attachment

  8. Probabilistic thermo-chemical analysis of a pultruded composite rod

    DEFF Research Database (Denmark)

    Baran, Ismet; Tutum, Cem Celal; Hattel, Jesper Henri

    2012-01-01

    In the present study the deterministic thermo-chemical pultrusion simulation of a composite rod taken from the literature [7] is used as a validation case. The predicted centerline temperature and cure degree profiles of the rod match well with those in the literature [7]. Following the validation...... case, the probabilistic design of the pultrusion process, which has not been considered until now, is performed. The effect of statistical variations in the material (i.e. fiber and resin) and resin kinetic properties, as well as process parameters such as pulling speed and inlet temperature...... on the product quality (degree of cure) are examined by means of Monte Carlo Simulation (MCS) with Latin Hypercube Sampling (LHS) technique. The variations in the activation energy as well as the density of the resin are found to have a strong influence on the centerline degree of cure at the exit whereas...

  9. Research on the Fatigue Life Prediction Method of Thrust Rod

    Directory of Open Access Journals (Sweden)

    Guoyu Feng

    2016-01-01

    Full Text Available Purpose of this paper is to investigate the fatigue life prediction method of the thrust rod based on the continuum damage mechanics. The equivalent stress used as damage parameters established rubber fatigue life prediction model. Through the finite element simulation and material test, the model parameters and the fatigue damage dangerous positions were obtained. By equivalent stress life model, uniaxial fatigue life of the V-type thrust rod is analyzed to predict the ratio of life and the life of the test was 1.73, within an acceptable range, and the fatigue damage occurring position and finite element analysis are basically the same. Fatigue life analysis shows that the method is of correct, theoretical, and practical value.

  10. Heater rod temperature change at boiling transition under flow oscillation

    International Nuclear Information System (INIS)

    Kasai, Shigeru; Toba, Akio; Takigawa, Yukio; Ebata, Shigeo; Morooka, Shin-ichi; Shirakawa, Ken-etsu; Utsuno, Hideaki.

    1986-01-01

    The experiments were performed to investigate the boiling transition phenomenon under flow oscillation (OSBT) during thermal hydraulic instability. It was found, from the experimental results, that the thermal hydraulic instability did not immediately lead to the boiling transition (BT) and, even when the BT occurred due to a power increase, the change in the heater rod temperature was periodically up and down with a saw-toothed shape and no excursion occurred. To investigate the temperature change characteristics, an analysis was also performed using the transient thermal hydraulics code. The analytical results showed that the shape of the heater rod temperature change was well simulated by presuming a repeat of alternate BT and rewetting. Based on these results, further analysis has been performed with the lumped parameter model to investigate the temperature profile characteristics as well as the effects of the post-BT heat transfer coefficient and the flow oscillation period on the maximum temperature. (author)

  11. Anti-ejection system for control rod drives

    International Nuclear Information System (INIS)

    Matthews, J.C.

    1977-01-01

    A linearly movable latch mechanism is provided to move into engagement with a deformable collet whenever an undesired ejection of a leadscrew is initiated from a nuclear reactor mounted control rod drive. Such an undesired ejection would occur in the event of a rupture in a housing of the control rod drive. The collet is deformed by the linear movement of the latch mechanism to wedge itself against the leadscrew and prevent the ejection of the leadscrew from the housing. The latch mechanism is made to be controllably engageable with the leadscrew and when thus engaged to allow the leadscrew to move in a control direction while moving with the leadscrew to engage and deform the collet when the leadscrew moves in an ejection direction. 13 claims, 2 figures

  12. Development of control system of coating of rod hydraulic cylinders

    Science.gov (United States)

    Aizhambaeva, S. Zh; Maximova, A. V.

    2018-01-01

    In this article, requirements to materials of hydraulic cylinders and methods of eliminating the main factors affecting the quality of the applied coatings rod hydraulic cylinders. The chromium plating process - one of ways of increase of anti-friction properties of coatings rods, stability to the wear and corrosion. The article gives description of differences of the stand-speed chromium plating process from other types of chromium plating that determines a conclusion about cutting time of chromium plating process. Conducting the analysis of technological equipment suggested addressing the modernization of high-speed chromium plating processes by automation and mechanization. Control system developed by design of schematic block diagram of a modernized and stand-speed chromium plating process.

  13. The behaviour of control rod absorber under irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Bourgoin, J. [Electricite de France, Avoine (France). Groupe des Laboratoires; Couvreur, F.; Gosset, D. [CEA-Saclay, Service d' Etude des Materiaux Irradies, 91191, Gif-sur-Yvette (France); Defoort, F.; Monchanin, M. [Framatome Nuclear Fuel, 10 rue J. Recamier 69456, Lyon (France); Thibault, X. [Electricite de France, SEPTEN, 12-14 avenue Dutrievoz, 69628, Villeurbanne (France)

    1999-12-01

    Increase of rod diameters and cracking of PWR control rod claddings may occur in operation. In order to understand the contribution of the absorber properties to this damage, EDF and FRAMATOME launched a programme of examinations concerning the silver-indium-cadmium alloy constituting the absorber bars. Density measurements and microstructural investigations such as micrography, microanalysis were carried out in the EDF Hot Laboratory, X-ray diffraction was performed by CEA. The results show that transmutations induce chemical modifications inside the FCC alloy and, further, formation of an HCP phase similar to the {zeta} phase of the silver alloys. The chemical and crystallographic changes account for the major part of the absorber swelling. (orig.)

  14. Large-scale transport across narrow gaps in rod bundles

    Energy Technology Data Exchange (ETDEWEB)

    Guellouz, M.S.; Tavoularis, S. [Univ. of Ottawa (Canada)

    1995-09-01

    Flow visualization and how-wire anemometry were used to investigate the velocity field in a rectangular channel containing a single cylindrical rod, which could be traversed on the centreplane to form gaps of different widths with the plane wall. The presence of large-scale, quasi-periodic structures in the vicinity of the gap has been demonstrated through flow visualization, spectral analysis and space-time correlation measurements. These structures are seen to exist even for relatively large gaps, at least up to W/D=1.350 (W is the sum of the rod diameter, D, and the gap width). The above measurements appear to compatible with the field of a street of three-dimensional, counter-rotating vortices, whose detailed structure, however, remains to be determined. The convection speed and the streamwise spacing of these vortices have been determined as functions of the gap size.

  15. High-resolution flow structure measurements in a rod bundle

    International Nuclear Information System (INIS)

    Ylönen, A. T.

    2013-01-01

    Flow behaviour inside a rod bundle has been an active research topic since the early days of the nuclear power industry. Of particular interest in previous studies have been topics such as flow mixing, two-phase flow structure and mapping of two-phase flow transitions. The optimisation of fuel element design can only be achieved by truly understanding the nature of flow. The ultimate goal in this research is to enhance the heat transfer and increase the critical heat flux, which would improve the fuel economy. A better understanding of the flow would also improve nuclear safety as departure from nucleate boiling (DNB) can be predicted more accurately. The motivation for the current project (SUBFLOW) was to increase knowledge of the complex flow phenomena inside a rod bundle. A dedicated sub-channel flow test facility was designed and constructed at the Paul Scherrer Institut (PSI), Villigen, Switzerland. An adiabatic test loop has an up-scaled (1:2.6) vertical fuel rod bundle model with a 4 × 4 geometry. For the very first time, the wire-mesh sensor measurement technique was implemented in a rod bundle as two 64×64 conductivity wire-mesh sensors were installed in the upper part of the test section. The measurement technique enables one to study single- and two-phase flow behaviour with high spatial and temporal resolution. The research topics addressed in this thesis cover a wide range of flow conditions with and without a spacer grid in a rod bundle. The experimental campaign was started by studying natural mixing of a passive scalar to characterise the development of turbulent diffusion in an injection sub-channel and, later on, cross-mixing between adjacent sub-channels. The results were also used in comparison with the in-house CFD code PSI-Boil that is being developed at PSI. The code could estimate the mixing inside the sub-channel and the transition to cross-mixing with a good accuracy. As a natural transition, the SUBFLOW experiments were continued by

  16. Turbulent flow through a wall subchannel of a rod bundle

    International Nuclear Information System (INIS)

    Rehme, K.

    1978-04-01

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

  17. Validation of fuel rod performance analysis code COPERNIC

    International Nuclear Information System (INIS)

    Han Yebin; Wang Jun; Ren Qisen; Liu Tong; Zhou Yuemin

    2012-01-01

    IAEA has sponsored the FUMEX Ⅲ (FUel Modeling at Extended Burnup) coordinated research project to improve computer code used for fuel behaviour simulation. As one of over thirty international participants, CGNPC has been engaged in testing and developing the fuel modelling code COPERNIC against data and cases provided by the IAEA and OECD/NEA. Investigations focused on high burnup and transient analysis, and include dimensional change model- ling. Data from several 6 calculation cases have been compared with COPERNIC predictions by far. Due to different purposes of tests, these cases had different designs including rod refabrication and annular pellet and were under different operation conditions including normal operation and ramp test. The comparison and preliminary analysis between predicted and measured results in such as fuel temperature, cladding outer diameter, cladding corrosion layer thickness, and fission gas release have been conducted, which demonstrated that the COPERNIC code was applicable to different rod designs under different operation conditions with an accurate prediction. (authors)

  18. The readout driver (ROD) for the ATLAS liquid argon calorimeters

    CERN Document Server

    Efthymiopoulos, I

    2001-01-01

    The Readout Driver (ROD) for the Liquid Argon calorimeter of the ATLAS detector is described. Each ROD module receives triggered data from 256 calorimeter cells via two fiber-optics 1.28 Gbit/s links with a 100 kHz event rate (25 kbit/event). Its principal function is to determine the precise energy and timing of the signal from discrete samples of the waveform, taken each period of the LHC clock (25 ns). In addition, it checks, histograms, and formats the digital data stream. A demonstrator system, consisting of a motherboard and several daughter-board processing units (PUs) was constructed and is currently used for tests in the lab. The design of this prototype board is presented here. The board offers maximum modularity and allows the development and testing of different PU designs based on today's leading integer and floating point DSPs. (3 refs).

  19. Transmission efficiency measurement at the FNAL 4-rod RFQ

    Energy Technology Data Exchange (ETDEWEB)

    Carneiro, J. P. [Fermilab; Garcia, F. G. [Fermilab; Ostiguy, J. F. [Fermilab; Saini, A. [Fermilab; Zwaska, R. [Fermilab; Mustapha, B. [Argonne; Ostroumov, P. [Argonne

    2014-12-01

    This paper presents measurements of the beam transmission performed on the 4-rod RFQ currently under operation at Fermilab. The beam current has been measured at the RFQ exit as a function of the magnetic field strength in the two LEBT solenoids. This measurement is compared with scans performed on the FermiGrid with the beam dynamics code TRACK. A particular attention is given to the impact, on the RFQ beam transmission, of the space-charge neutralization in the LEBT.

  20. Microstructural observations on the terminal penetration of long rod projectile

    OpenAIRE

    Krushna Kumbhar; P. Ponguru Senthil; A.K. Gogia

    2017-01-01

    Present study focuses on the terminal penetration of tungsten heavy alloy (WHA) long rod penetrator impacted against armour steel at an impact velocity of 1600 m/s. The residual penetrator and armour steel target recovered after the ballistic test have been characterized using optical microscope, scanning electron microscope (SEM) and electron probe micro analyzer (EPMA). Metallurgical changes in target steel and WHA remnant have been analysed. Large shear stresses and shear localization have...

  1. First Test Results of the 4-ROD Crab Cavity

    CERN Document Server

    Ambattu, P; Burt, G; Calaga, R; Capatina, O; Calatroni, S; Ciapala, E; Doherty, D; Ferreira, L; Jensen, E; Hall, B; Lingwood, C; Maesen, P; Mongelluzzo, A; Renaglia, T; Therasse, M

    2013-01-01

    The first compact prototype crab cavity with the 4rod geometry has undergone surface treatment and cold testing. Due to the complex geometry and unique fabrication procedure, RF validation of the field at beyond the nominal operating voltage at a sufficiently high Q0 is an important pre-requisite. Preliminary results of the first cold tests are presented along with cavity performance at different stages of the cavity processing is described.

  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. Control rod drive mechanism with shock absorber for nuclear reactor

    International Nuclear Information System (INIS)

    Chevereau, G.

    1989-01-01

    The mechanism usable in a PWR has a shaft carrying the bar vertically displaceable in the reactor internals and a dash pot with a hydraulic cylinder and a piston. The cylinder has a large diameter perforated upper section to the cylinder, a small diameter lower section, a piston traversed by the control rod sized to fit into the upper section and forced downwards when the control descends. The shock absorbing chamber is defined between the piston and the upper section [fr

  4. Auxiliary device for the assembly of cruciform BWR control rods

    International Nuclear Information System (INIS)

    Lippert, H.J.

    1980-01-01

    The mounting auxiliary equipment has got a frame of the size of a core cell that can be put on the core grid. Diagonally arranged link bodies serve to center the cross-shaped control rod. The mounting auxiliary equipment consists of a single piece and therefore facilitates manipulation for which a diagonal handle is provided at the upper end. It may be swung aside. (RW) [de

  5. Auxiliary device for the assembly of cruciform BWR control rods

    International Nuclear Information System (INIS)

    Lippert, H.J.

    1982-01-01

    The mounting auxiliary equipment has got a frame of the size of a core cell that can be put on the core grid. Diagonally arranged link bodies serve to center the cross-shaped control rod. The mounting auxiliary equipment consists of a single piece and therefore facilitates manipulation for which a diagonal handle is provided at the upper end. It may be swung aside. (orig./HP)

  6. Dipole stabilizer rods for 400 keV deuteron RFQ

    International Nuclear Information System (INIS)

    Sista, V.L.S. Rao; Srivastava, S.C.L.; Pande, Rajni; Roy, Shweta; Singh, P.

    2009-01-01

    In our 400 keV deuteron RFQ for neutron production, the destructive dipolar modes are very close to the required quadrupolar mode. In order to increase the spacing between the quadrupole and dipole modes the dipolar stabilizer rods (DSR's) are used. The design of the DSR's is done using the computer code CST Microwave studio. The variation of the quadrupole and dipolar mode frequencies with the radius and length of the DSR's are studied. (author)

  7. Characterization of the retinal proteome during rod photoreceptor genesis

    Directory of Open Access Journals (Sweden)

    Hecker Laura A

    2010-01-01

    Full Text Available Abstract Background The process of rod photoreceptor genesis, cell fate determination and differentiation is complex and multi-factorial. Previous studies have defined a model of photoreceptor differentiation that relies on intrinsic changes within the presumptive photoreceptor cells as well as changes in surrounding tissue that are extrinsic to the cell. We have used a proteomics approach to identify proteins that are dynamically expressed in the mouse retina during rod genesis and differentiation. Findings A series of six developmental ages from E13 to P5 were used to define changes in retinal protein expression during rod photoreceptor genesis and early differentiation. Retinal proteins were separated by isoelectric focus point and molecular weight. Gels were analyzed for changes in protein spot intensity across developmental time. Protein spots that peaked in expression at E17, P0 and P5 were picked from gels for identification. There were 239 spots that were picked for identification based on their dynamic expression during the developmental period of maximal rod photoreceptor genesis and differentiation. Of the 239 spots, 60 of them were reliably identified and represented a single protein. Ten proteins were represented by multiple spots, suggesting they were post-translationally modified. Of the 42 unique dynamically expressed proteins identified, 16 had been previously reported to be associated with the developing retina. Conclusions Our results represent the first proteomics study of the developing mouse retina that includes prenatal development. We identified 26 dynamically expressed proteins in the developing mouse retina whose expression had not been previously associated with retinal development.

  8. Hydrodynamic behavior of a bare rod bundle. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Bartzis, J.G.; Todreas, N.E.

    1977-06-01

    The temperature distribution within the rod bundle of a nuclear reactor is of major importance in nuclear reactor design. However temperature information presupposes knowledge of the hydrodynamic behavior of the coolant which is the most difficult part of the problem due to complexity of the turbulence phenomena. In the present work a 2-equation turbulence model--a strong candidate for analyzing actual three dimensional turbulent flows--has been used to predict fully developed flow of infinite bare rod bundle of various aspect ratios (P/D). The model has been modified to take into account anisotropic effects of eddy viscosity. Secondary flow calculations have been also performed although the model seems to be too rough to predict the secondary flow correctly. Heat transfer calculations have been performed to confirm the importance of anisotropic viscosity in temperature predictions. All numerical calculations for flow and heat have been performed by two computer codes based on the TEACH code. Experimental measurements of the distribution of axial velocity, turbulent axial velocity, turbulent kinetic energy and radial Reynolds stresses were performed in the developing and fully developed regions. A 2-channel Laser Doppler Anemometer working on the Reference mode with forward scattering was used to perform the measurements in a simulated interior subchannel of a triangular rod array with P/D = 1.124. Comparisons between the analytical results and the results of this experiment as well as other experimental data in rod bundle array available in literature are presented. The predictions are in good agreement with the results for the high Reynolds numbers.

  9. Calculation of fission gases internal pressure in nuclear fuel rods

    International Nuclear Information System (INIS)

    Vasconcelos Santana, M. de.

    1981-12-01

    Models concerning the principal phenomena, particularly thermal expansion, fuel swelling, densification, reestructuring, relocation, mechanical strain, fission gas production and release, direct or indirectly important to calculate the internal pressure in nuclear fuel rods were analysed and selected. Through these analyses a computer code was developed to calculate fuel pin internal pressure evolution. Three different models were utilized to calculate the internal pressure in order to select the best and the most conservative estimate. (Author) [pt

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

    Science.gov (United States)

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

    2000-01-01

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

  11. Improvement to the control rod drive of a nuclear reactor

    International Nuclear Information System (INIS)

    Desfontaines, Guy.

    1981-01-01

    Improvement to the devices that move the control rods of a nuclear reactor. The slow movements of the rods are generally carried out by screw and nut gear, the nut being blocked as to rotation and the screw as to translation movement. Additionally, a mechanism enables the control rods to be inserted rapidly by release of the screw and nut gear, the nut remaining constantly in gear with the screw. The presence of extra poles and coils under the stator of the actuating motor of the screw add length and weight to the mechanism and hence increase the strains and deformations which affect the latter in the event of an earthquake. The device of the invention makes it possible to overcome this drawback and leads to a more simple mechanism. It is characterized in that the rotor of the motor actuating the screw is also provided with clamps, in its high position, controlled by electromagnetic action as from the coils of the actuating motor stator so that they are in the closed position on the screw when the stator is powered and in the open position when it is no longer so, in order to allow the screw and nut assembly drop, and in that it includes a device to lock the clamps, enabling these to be kept in the open position when the control screw is not in the high holding position [fr

  12. Minimum slugging velocity in fluidized beds containing vertical rods

    Energy Technology Data Exchange (ETDEWEB)

    Coronella, C.J.; Lee, S.Y.; Seader, J.D. (University of Utah, Salt Lake City, UT (United States). Dept. of Chemical Engineering)

    1994-09-01

    A new method for determining the onset of slugging in fluidized beds is presented. Pressure-drop fluctuations, measured from below the distributor to the gas exit line, are transformed to the frequency domain by the power spectral desity function (PSDF). The dominant frequency of the PSDF corresponds to the eruption frequency of bubbles or slugs. A fluidized bed is in the slugging regime when this dominant frequency, f[sub d], remains constant with changing gas velocity. This method is an improvement over previous methods because of the simple nature of the apparatus required, and because it is possible to locate the pressure probes so that they do not interfere with the fluidization or undergo rapid wear from the constant particle movement. This method was used to determine the gas velocity corresponding to the transition from the bubbling to the slugging regime for a 10cm diameter bed of sand fluidized with air and containing three 1.9cm diameter vertical rods at 5.2cm centre-to-centre triangular spacing and extending the length of the bed, and to compare the results with those from the same bed without any internal rods. The presence of the vertical rods inhibited the onset of the slugging regime, and significantly extended the bubbling regime to higher gas velocities. 32 refs., 12 figs.

  13. 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

  14. Technique of manufacturing specimen of irradiated fuel rods

    International Nuclear Information System (INIS)

    Min, Duck Seok; Seo, Hang Seok; Min, Duck Kee; Koo, Dae Seo; Lee, Eun Pyo; Yang, Song Yeol

    1999-04-01

    Technique of manufacturing specimen of irradiated fuel rods to perform efficient PIE is developed by analyzing the relation between requiring time of manufacturing specimen and manufacturing method in irradiated fuel rods. It takes within an hour to grind 1 mm of specimen thickness under 150 rpm in speed of grinding, 600 g gravity in force using no.120, no.240, no.320 of grinding paper. In case of no.400 of grinding paper, it takes more an hour to grind the same thickness as above. It takes up to a quarter to grind 80-130 μm in specimen thickness using no.400 of grinding paper. When grinding time goes beyond 15 minutes, the grinding thickness of specimen does not exist. The polishing of specimen with 150 Rpms in speed of grinding machine, 600 g gravity in force, 10 minutes in polishing time using diamond paste 15 μm on polishing cloths amounts to 50 μm in specimen thickness. In case of diamond paste 9 μm on polishing cloth, the polishing of specimen amounts to 20 μm. The polishing thickness of specimen with 15 minutes in polishing time using 6 μm, 3 μm, 1 μm, 1/4 μm does not exist. Technique of manufacturing specimen of irradiated fuel rods will have application to the destructive examination of PIE. (author). 6 refs., 1 tab., 10 figs

  15. Large eddy simulation of a fuel rod subchannel

    International Nuclear Information System (INIS)

    Mayer, Gusztav

    2007-01-01

    In a VVER-440 reactor the measured outlet temperature is related to fuel limit parameters and the power upgrading plans of VVER-440 reactors motivated us to obtain more information on the mixing process of the fuel assemblies. In a VVER-440 rod bundle the fuel rods are arranged in triangular array. Measurement shows (Krauss and Meyer, 1998) that the classical engineering approach, which tries to trace the characterization of such systems back to equivalent (hydraulic diameter) pipe flows, does not give reasonable results. Due to the different turbulence characteristics, the mixing is more intensive in rod bundles than it would be expected based on equivalent pipe flow correlations. As a possible explanation of the high mixing, secondary flow was deduced from measurements by several experimentalists (Trupp and Azad, 1975). Another candidate to explain the high mixing is the so-called flow pulsation phenomenon (Krauss and Meyer, 1998). In this paper we present subchannel simulations (Mayer et al. 2007) using large eddy simulation (LES) methodology and the lattice Boltzmann method (LBM) without the spacers at Reynolds number 21000. The simulation results are compared with the measurements of Trupp and Azad (1975). The mean axial velocity profile shows good agreement with the measurement data. Secondary flow has been observed directly in the simulation results. Reasonable agreement has been achieved for most Reynolds stresses. Nevertheless, the calculated normal stresses show small, but systematic deviation from the measurement data. (author)

  16. Interim transfer canister for consolidating nuclear fuel rods

    International Nuclear Information System (INIS)

    Formanek, F.J.

    1987-01-01

    This patent describes a canister for receiving and consolidating a group of uniformly spaced apart nuclear fuel rods, comprising: a rectangular, vertically oriented straight back panel; a pair of oppositely disposed side panels connected perpendicularly to the back panel, having a vertical straight upper portion and an inwardly tapered lower portion; a front panel opposite the back panel and connected to the side panels, having a straight vertical upper portion and inwardly tapered lower portion; whereby the back, side and front panels define a rectangular upper opening at the upper end of the canister and a generally rectangular lower opening at the other end, the lower opening having a cross-sectional area less than one-half that of the upper opening; parallel plate members spanning the canister from the front panel to the back panel, each plate spaced from the other the same uniform distance, the plates extending downwardly into the tapered portion of the canister while remaining spaced above the tapered sidewalls; first base means at the lower end of the canister, removably mounted and having an oblique orientation generally downward from the front panel to the back panel, for guiding the fuel rods to be inserted preferentially toward the lower portion of the back panel; and second base means removably mounted within the canister below first base means and oriented transversely to the longitudinal extent of the canister, for supporting the fuel rods when the first base means is removed from the canister

  17. Simulation error propagation for a dynamic rod worth measurement technique

    International Nuclear Information System (INIS)

    Kastanya, D.F.; Turinsky, P.J.

    1996-01-01

    KRSKO nuclear station, subsequently adapted by Westinghouse, introduced the dynamic rod worth measurement (DRWM) technique for measuring pressurized water reactor rod worths. This technique has the potential for reduced test time and primary loop waste water versus alternatives. The measurement is performed starting from a slightly supercritical state with all rods out (ARO), driving a bank in at the maximum stepping rate, and recording the ex-core detectors responses and bank position as a function of time. The static bank worth is obtained by (1) using the ex-core detectors' responses to obtain the core average flux (2) using the core average flux in the inverse point-kinetics equations to obtain the dynamic bank worth (3) converting the dynamic bank worth to the static bank worth. In this data interpretation process, various calculated quantities obtained from a core simulator are utilized. This paper presents an analysis of the sensitivity to the impact of core simulator errors on the deduced static bank worth

  18. Accuracy and Repeatability of Trajectory Rod Measurement Using Laser Scanners.

    Science.gov (United States)

    Liscio, Eugene; Guryn, Helen; Stoewner, Daniella

    2017-12-22

    Three-dimensional (3D) technologies contribute greatly to bullet trajectory analysis and shooting reconstruction. There are few papers which address the errors associated with utilizing laser scanning for bullet trajectory documentation. This study examined the accuracy and precision of laser scanning for documenting trajectory rods in drywall for angles between 25° and 90°. The inherent error range of 0.02°-2.10° was noted while the overall error for laser scanning ranged between 0.04° and 1.98°. The inter- and intraobserver errors for trajectory rod placement and virtual trajectory marking showed that the range of variation for rod placement was between 0.1°-1° in drywall and 0.05°-0.5° in plywood. Virtual trajectory marking accuracy tests showed that 75% of data values were below 0.91° and 0.61° on azimuth and vertical angles, respectively. In conclusion, many contributing factors affect bullet trajectory analysis, and the use of 3D technologies can aid in reduction of errors associated with documentation. © 2017 American Academy of Forensic Sciences.

  19. 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

  20. Control rod effects with plutonium recycle in a PWR

    International Nuclear Information System (INIS)

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

    1979-03-01

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

  1. Anti-ejection device, which can be released, for control rods of nuclear reactor

    International Nuclear Information System (INIS)

    Belz, G.

    1983-01-01

    The present invention proposes an anti-ejection device which allows to withdraw the control rod out of a PWR reactor core if the locking systems of the rod translation are streck. This device prohibits the control rod ejection as long as an effort lower than a predetermined value is not applied on the control rod. This limit value is determined with regard of the efforts which may be applied on the control rod in case of an external accidental source. Nevertheless, if the anti-ejection mechanism remains stuck, it is however possible to withdraw the control rod out of the core applying on its control rod drives an effort higher than the limit value [fr

  2. Analysis of Subchannel and Rod Bundle PSBT Experiments with CATHARE 3

    Directory of Open Access Journals (Sweden)

    M. Valette

    2012-01-01

    Full Text Available This paper presents the assessment of CATHARE 3 against PWR subchannel and rod bundle tests of the PSBT benchmark. Noticeable measurements were the following: void fraction in single subchannel and rod bundle, multiple liquid temperatures at subchannel exit in rod bundle, and DNB power and location in rod bundle. All these results were obtained both in steady and transient conditions. Void fraction values are satisfactory predicted by CATHARE 3 in single subchannels with the pipe module. More dispersed predictions of void values are obtained in rod bundles with the CATHARE 3 3D module at subchannel scale. Single-phase liquid mixing tests and DNB tests in rod bundle are also analyzed. After calibrating the mixing in liquid single phase with specific tests, DNB tests using void mixing give mitigated results, perhaps linked to inappropriate use of CHF lookup tables in such rod bundles with many spacers.

  3. Effect of electrode shape on grounding resistances - Part 2

    DEFF Research Database (Denmark)

    Tomaskovicova, Sonia; Ingeman-Nielsen, Thomas; Christiansen, Anders V.

    2016-01-01

    Although electric resistivity tomography (ERT) is now regarded as a standard tool in permafrost monitoring, high grounding resistances continue to limit the acquisition of time series over complete freeze-thaw cycles. In an attempt to alleviate the grounding resistance problem, we have tested three...... electrode designs featuring increasing sizes and surface area, in the laboratory and at three different field sites in Greenland. Grounding resistance measurements showed that changing the electrode shape (using plates instead of rods) reduced the grounding resistances at all sites by 28%-69% during...... unfrozen and frozen ground conditions. Using meshes instead of plates (the same rectangular shape and a larger effective surface area) further improved the grounding resistances by 29%-37% in winter. Replacement of rod electrodes of one entire permanent permafrost monitoring array by meshes resulted...

  4. Minor actinide transmutation on PWR burnable poison rods

    International Nuclear Information System (INIS)

    Hu, Wenchao; Liu, Bin; Ouyang, Xiaoping; Tu, Jing; Liu, Fang; Huang, Liming; Fu, Juan; Meng, Haiyan

    2015-01-01

    Highlights: • Key issues associated with MA transmutation are the appropriate loading pattern. • Commercial PWRs are the only choice to transmute MAs in large scale currently. • Considerable amount of MA can be loaded to PWR without disturbing k eff markedly. • Loading MA to PWR burnable poison rods for transmutation is an optimal loading pattern. - Abstract: Minor actinides are the primary contributors to long term radiotoxicity in spent fuel. The majority of commercial reactors in operation in the world are PWRs, so to study the minor actinide transmutation characteristics in the PWRs and ultimately realize the successful minor actinide transmutation in PWRs are crucial problem in the area of the nuclear waste disposal. The key issues associated with the minor actinide transmutation are the appropriate loading patterns when introducing minor actinides to the PWR core. We study two different minor actinide transmutation materials loading patterns on the PWR burnable poison rods, one is to coat a thin layer of minor actinide in the water gap between the zircaloy cladding and the stainless steel which is filled with water, another one is that minor actinides substitute for burnable poison directly within burnable poison rods. Simulation calculation indicates that the two loading patterns can load approximately equivalent to 5–6 PWR annual minor actinide yields without disturbing the PWR k eff markedly. The PWR k eff can return criticality again by slightly reducing the boric acid concentration in the coolant of PWR or removing some burnable poison rods without coating the minor actinide transmutation materials from PWR core. In other words, loading minor actinide transmutation material to PWR does not consume extra neutron, minor actinide just consumes the neutrons which absorbed by the removed control poisons. Both minor actinide loading patterns are technically feasible; most importantly do not need to modify the configuration of the PWR core and

  5. Multiple rod-cone and cone-rod photoreceptor transmutations in snakes: evidence from visual opsin gene expression.

    Science.gov (United States)

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

    2016-01-27

    In 1934, Gordon Walls forwarded his radical theory of retinal photoreceptor 'transmutation'. This proposed that rods and cones used for scotopic and photopic vision, respectively, were not fixed but could evolve into each other via a series of morphologically distinguishable intermediates. Walls' prime evidence came from series of diurnal and nocturnal geckos and snakes that appeared to have pure-cone or pure-rod retinas (in forms that Walls believed evolved from ancestors with the reverse complement) or which possessed intermediate photoreceptor cells. Walls was limited in testing his theory because the precise identity of visual pigments present in photoreceptors was then unknown. Subsequent molecular research has hitherto neglected this topic but presents new opportunities. We identify three visual opsin genes, rh1, sws1 and lws, in retinal mRNA of an ecologically and taxonomically diverse sample of snakes central to Walls' theory. We conclude that photoreceptors with superficially rod- or cone-like morphology are not limited to containing scotopic or photopic opsins, respectively. Walls' theory is essentially correct, and more research is needed to identify the patterns, processes and functional implications of transmutation. Future research will help to clarify the fundamental properties and physiology of photoreceptors adapted to function in different light levels. © 2016 The Author(s).

  6. Manual materials handling in mining: the effect of rod heights and foot positions when lifting "in-the-hole" drill rods.

    Science.gov (United States)

    Plamondon, André; Delisle, Alain; Trimble, Karin; Desjardins, Pierre; Rickwood, Trevor

    2006-11-01

    There is a paucity of studies focusing on the lifting of rods or long awkward heavy objects. In-the-hole (ITH) drilling is a heavy repetitive mining task, which has been identified as having a relatively high incidence and severity rate of musculoskeletal injuries. The purpose of this study was to examine how the load experienced by ITH drill operators changed when lifting a vertical drilling rod (1.61m, 35kg) using two rod heights and four different foot positions. In addition, a symmetrical lift with a lifting index (LI) of 1.4 also served as a comparison to determine possible risk of low back injury. Eleven experienced ITH drill operators participated in the study. Each subject was required to lift a vertical drilling rod until the upper body was in an erect posture using four different foot positions (0 degrees =subject facing the rod, 45 degrees =subject oblique to the rod, 90 degrees =subject right side to the rod and freestyle). In addition, two rod height conditions were studied where the base of the vertical rod was supported either (1) at ground level (height of rod CG=0.83m) or (2) on a 20cm rack (height of rod CG=1.03m). Finally, each subject lifted a 21.5kg box in the sagittal plane, which corresponded to an LI of 1.4 in the NIOSH lifting equation. Reflective markers were placed on the subjects, and three video cameras and one force plate were used to record the forces and the motion of the subjects' segments. Two surface electrodes were applied on the right and the left erector spinae (ES) at the level of L3. Back loading was defined by the level of the peak moments, the mechanical work and erector spinae muscle activity (EMG). It was found that the vertical height of the rod had the most significant impact on back loading, while the effect of the initial foot positioning relative to the rod was limited by the technique adopted by the drillers. Moreover, it was found that some of the subjects used techniques less strenuous for the back than others

  7. Report on the shearing, dissolution and analysis of GRIP-II rod 79-453 (validation rod); Light Water Breeder Reactor proof-of-breeding analytical support project

    International Nuclear Information System (INIS)

    Levitz, N.M.; Parks, J.E.; Winsch, I.O.; Meyer, R.J.; Graczyk, D.G.; Tomlinson, G.; Deeken, P.G.

    1981-10-01

    This report covers the processing and analysis of the fuel-bearing section (M-5138) of an irradiated experimental Light Water Breeder Reactor fuel rod, GRIP-II rod No. 79-453; this section has been designated the Validation Rod. Process steps included precision shearing of the rod into eight comminuted segments, dissolution of the segments, and chemical and radiometric analyses of the resulting solutions. The shearing and dissolution were carried out fully remotely in an existing pilot-scale facility installed in a shielded cell. Data are provided on physical parameters of the rod section and segments, uranium assays and isotopic abundances, and selected fission products. An error analysis of the individual measurements and analyses is included

  8. Dimensional analysis and extended hydrodynamic theory applied to long-rod penetration of ceramics

    Directory of Open Access Journals (Sweden)

    J.D. Clayton

    2016-08-01

    Full Text Available Principles of dimensional analysis are applied in a new interpretation of penetration of ceramic targets subjected to hypervelocity impact. The analysis results in a power series representation – in terms of inverse velocity – of normalized depth of penetration that reduces to the hydrodynamic solution at high impact velocities. Specifically considered are test data from four literature sources involving penetration of confined thick ceramic targets by tungsten long rod projectiles. The ceramics are AD-995 alumina, aluminum nitride, silicon carbide, and boron carbide. Test data can be accurately represented by the linear form of the power series, whereby the same value of a single fitting parameter applies remarkably well for all four ceramics. Comparison of the present model with others in the literature (e.g., Tate's theory demonstrates a target resistance stress that depends on impact velocity, linearly in the limiting case. Comparison of the present analysis with recent research involving penetration of thin ceramic tiles at lower typical impact velocities confirms the importance of target properties related to fracture and shear strength at the Hugoniot Elastic Limit (HEL only in the latter. In contrast, in the former (i.e., hypervelocity and thick target experiments, the current analysis demonstrates dominant dependence of penetration depth only by target mass density. Such comparisons suggest transitions from microstructure-controlled to density-controlled penetration resistance with increasing impact velocity and ceramic target thickness.

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

    Science.gov (United States)

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

    2014-01-01

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

  10. Monte Carlo simulation of neutron transport in a homogeneous reactor with a partially inserted control rod

    International Nuclear Information System (INIS)

    Karlsson, J.K.H.; Linden, P.

    1997-01-01

    The neutron transport in a bare, cylindrical and homogeneous reactor, with and without the presence of a central partially inserted control rod, has been simulated by using a Monte Carlo transport code. The behaviour of both the flux and current in this system have been investigated. We have found that the flux and especially the current are strongly affected by the presence of the control rod in its close vicinity. The results indicate the feasibility to identify the position and especially the tip of the rod from the flux and current. Further, the direction to the rod can be found from the current vector. The information content regarding the position of the rod, in both the neutron flux and the current, decays strongly as a function of distance and it is dependent on the size of the rod. In our model, the practical range over which the flux or current can be a useful indicator of the position of the tip of the rod is about 10-15 cm for a rod with a diameter of 2 cm. The practical range for identification of the position of the rod is greater for a rod of larger diameter

  11. Rod consolidation of RG and E's [Rochester Gas and Electric Corporation] spent PWR [pressurized water reactor] fuel

    International Nuclear Information System (INIS)

    Bailey, W.J.

    1987-05-01

    The rod consolidation demonstration involved pulling the fuel rods from five fuel assemblies from Unit 1 of RG and E's R.E. Ginna Nuclear Power Plant. Slow and careful rod pulling efforts were used for the first and second fuel assemblies. Rod pulling then proceeded smoothly and rapidly after some minor modifications were made to the UST and D consolidation equipment. The compaction ratios attained ranged from 1.85 to 2.00 (rods with collapsed cladding were replaced by dummy rods in one fuel assembly to demonstrate the 2:1 compaction ratio capability). This demonstration involved 895 PWR fuel rods, among which there were some known defective rods (over 50 had collapsed cladding); no rods were broken or dropped during the demonstration. However, one of the rods with collapsed cladding unexplainably broke during handling operations (i.e., reconfiguration in the failed fuel canister), subsequent to the rod consolidation demonstration. The broken rod created no facility problems; the pieces were encapsulated for subsequent storage. Another broken rod was found during postdemonstration cutting operations on the nonfuel-bearing structural components from the five assemblies; evidence indicates it was broken prior to any rod consolidation operations. During the demonstration, burnish-type lines or scratches were visible on the rods that were pulled; however, experience indicates that such lines are generally produced when rods are pulled (or pushed) through the spacer grids. Rods with collapsed cladding would not enter the funnel (the transition device between the fuel assembly and the canister that aids in obtaining high compaction ratios). Reforming of the flattened areas of the cladding on those rods was attempted to make the rod cross sections more nearly circular; some of the reformed rods passed through the funnel and into the canister

  12. Rewetting of a low superheated rod with saturated water

    Energy Technology Data Exchange (ETDEWEB)

    Portillo, O.; Reyes, R.; Wayner, P.C. Jr.

    1999-07-01

    The study of the rewetting of a superheated surface has application in several technological fields. It is related to the control mechanism for loss of coolant accident (LOCA) in nuclear reactors. An adsorption model as the precursory mechanism for rewetting of a superheated surface is extended from its application to non-polar liquids to a polar fluid, and modeling calculations are compared with experimental data found in the literature. The adsorption model is based on interfacial forces acting at the tip of the rewetting front, the three-phase region. In this region, solid, liquid and vapor interfaces generate a contact angle that depends on the degree of superheat and describes the velocity of rewetting. The contact angle is a function of interfacial forces calculated through the disjoining pressure of the adsorbed film precursory of the rewetting. The influences of van der Waals and electrostatic intermolecular forces in the film thickness are analyzed. The authors find that the order of magnitude of the film thickness in the controlling region is of a few angstroms: thus, only van der Waals intermolecular forces define the interactions. For the prediction of the velocity of rewetting the temperature profile along the rod's surface is required and a one-dimensional and a two-dimensional heat conduction balances are solved. The thermophysical properties in the adsorption model are predicted by ASPEN PLUS data bank and from ASME steam tables. Variations of the predicted values have a strong influence on the results. The surface boundary condition on the rod contains an evaporative heat transfer coefficient that is calculated from the fitted experimental rewetting velocities and the two-dimensional temperature field in the rod. Using this calculation scheme the values of the evaporative heat transfer coefficient are obtained in the normal range of values. Therefore the adsorption model gives results that are consistent with experimental observations.

  13. Rod monochromacy and the coevolution of cetacean retinal opsins.

    Directory of Open Access Journals (Sweden)

    Robert W Meredith

    2013-04-01

    Full Text Available Cetaceans have a long history of commitment to a fully aquatic lifestyle that extends back to the Eocene. Extant species have evolved a spectacular array of adaptations in conjunction with their deployment into a diverse array of aquatic habitats. Sensory systems are among those that have experienced radical transformations in the evolutionary history of this clade. In the case of vision, previous studies have demonstrated important changes in the genes encoding rod opsin (RH1, short-wavelength sensitive opsin 1 (SWS1, and long-wavelength sensitive opsin (LWS in selected cetaceans, but have not examined the full complement of opsin genes across the complete range of cetacean families. Here, we report protein-coding sequences for RH1 and both color opsin genes (SWS1, LWS from representatives of all extant cetacean families. We examine competing hypotheses pertaining to the timing of blue shifts in RH1 relative to SWS1 inactivation in the early history of Cetacea, and we test the hypothesis that some cetaceans are rod monochomats. Molecular evolutionary analyses contradict the "coastal" hypothesis, wherein SWS1 was pseudogenized in the common ancestor of Cetacea, and instead suggest that RH1 was blue-shifted in the common ancestor of Cetacea before SWS1 was independently knocked out in baleen whales (Mysticeti and in toothed whales (Odontoceti. Further, molecular evidence implies that LWS was inactivated convergently on at least five occasions in Cetacea: (1 Balaenidae (bowhead and right whales, (2 Balaenopteroidea (rorquals plus gray whale, (3 Mesoplodon bidens (Sowerby's beaked whale, (4 Physeter macrocephalus (giant sperm whale, and (5 Kogia breviceps (pygmy sperm whale. All of these cetaceans are known to dive to depths of at least 100 m where the underwater light field is dim and dominated by blue light. The knockout of both SWS1 and LWS in multiple cetacean lineages renders these taxa rod monochromats, a condition previously unknown among

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

    Science.gov (United States)

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

    2013-11-01

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

  15. Nuclear fuel rod with retainer for pellet stack

    International Nuclear Information System (INIS)

    Cloue, J.M.

    1986-01-01

    The rod, usable in pressurized water reactors, comprises a stack of fuel pellets and means holding the stack against an end plug of the fuel can during handling operations. These means include a radially expansive element (retainer) of which the shape is so that when it is free at ambient temperature it is gripping the inside of the casing, and a temperature sensitive spacer which contracts the retainer to release it from the casing at a temperature between the ambient and the operating temperature of a reactor [fr

  16. Breaking of rod-shaped model material during compression

    Directory of Open Access Journals (Sweden)

    Lukas Kulaviak

    2017-01-01

    Full Text Available The breakage of a model anisometric dry granular material caused by uniaxial compression was studied. The bed of uniform rod-like pasta particles (8 mm long, aspect ratio 1:8 was compressed (Gamlen Tablet Press and their size distribution was measured after each run (Dynamic Image Analysing. The compression dynamics was recorded and the effect of several parameters was tested (rate of compression, volume of granular bed, pressure magnitude and mode of application. Besides the experiments, numerical modelling of the compressed breakable material was performed as well, employing the DEM approach (Discrete Element Method. The comparison between the data and the model looks promising.

  17. Control rod drive WWER 1000 – tuning of input parameters

    Directory of Open Access Journals (Sweden)

    Markov P.

    2007-10-01

    Full Text Available The article picks up on the contributions presented at the conferences Computational Mechanics 2005 and 2006, in which a calculational model of an upgraded control rod linear stepping drive for the reactors WWER 1000 (LKP-M/3 was described and results of analysis of dynamical response of its individual parts when moving up- and downwards were included. The contribution deals with the tuning of input parameters of the 3rd generation drive with the objective of reaching its running as smooth as possible so as to get a minimum wear of its parts as a result and hence to achieve maximum life-time.

  18. Heat transfer in rod bundles with severe clad deformations

    International Nuclear Information System (INIS)

    Ihle, P.

    1984-04-01

    The content of the paper is focused on heat transfer conditions during the reflood phase of a LOCA in slightly to severely deformed PWR fuel rod bundle geometries. The status of analytical and, especially, of experimental work is described as far as it is possible within this frame. Emphasis is placed on the presentation of the results of ''Flooding Experiments with Blocked Arrays'' (FEBA), a program performed at the Kernforschungszentrum Karlsruhe in the frame of the Project Nuclear Safety (PNS). (orig./WL) [de

  19. Achieving anisotropy in metamaterials made of dielectric cylindrical rods

    DEFF Research Database (Denmark)

    Peng, Liang; Ran, Lixin; Mortensen, Asger

    2010-01-01

    We show that anisotropic negative effective dispersion relation can be achieved in pure dielectric rod-type metamaterials by turning from the symmetry of a square lattice to that of a rectangular one. Theoretical predictions and conclusions are verified by both numerical calculations and computer...... based simulations. The proposed anisotropic metamaterial, is used to construct a refocusing slab lens and a subdiffraction hyperlens. The all-dielectric origin makes it more straightforward to address loss and scaling, thus facilitating future applications in both the terahertz and optical range....

  20. Heat removal in gas-cooled fuel rod clusters

    International Nuclear Information System (INIS)

    Rehme, K.

    1975-01-01

    For a thermo- and fluid-dynamic analysis of fuel rod cluster subchannels for gas-cooled breeder reactors, the following values must be verified: a) friction coefficient as flow parameter; b) Stanton number as heat transfer parameter; c) influence of spacers on friction coefficient and Stanton number; d) heat and mass exchange between subchannels with different temperatures. These parameters are established by combining results of single experiments and of integral experiments. Mention is made of further studies to be performed in order to determine the heat removal from gas-cooled fast breeder fuel elements. (HR) [de