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Sample records for boron-carbide control rod

  1. Detection of a leaking boron-carbide control rod in a TRIGA Mark I reactor

    Energy Technology Data Exchange (ETDEWEB)

    Blotcky, A J; Arsenault, L J [General Medical Research, Veterans Administration Hospital, Omaha (United States)

    1974-07-01

    During a routine quarterly inspection of the boron-carbide control rods of the Omaha Veterans Administration Hospital 18 kW Triga Mark I reactor, a pin hole leak was detected approximately 3 mm from the chamfered edge. The leak was found by observing bubbles when the rod was withdrawn from the reactor tank for visual observation, and could not be seen with the naked eye. This suggests that pin hole leaks could occur and not be visually detected in control rods and fuel elements examined underwater. A review of the rod calibrations showed that the leak had not caused a loss in rod worth. Slides will be presented showing the bubbles observed during the inspection, together with an unmagnified and magnified view of the pin hole. (author)

  2. Detection of a leaking boron-carbide control rod in a TRIGA Mark I reactor

    International Nuclear Information System (INIS)

    Blotcky, A.J.; Arsenault, L.J.

    1974-01-01

    During a routine quarterly inspection of the boron-carbide control rods of the Omaha Veterans Administration Hospital 18 kW Triga Mark I reactor, a pin hole leak was detected approximately 3 mm from the chamfered edge. The leak was found by observing bubbles when the rod was withdrawn from the reactor tank for visual observation, and could not be seen with the naked eye. This suggests that pin hole leaks could occur and not be visually detected in control rods and fuel elements examined underwater. A review of the rod calibrations showed that the leak had not caused a loss in rod worth. Slides will be presented showing the bubbles observed during the inspection, together with an unmagnified and magnified view of the pin hole. (author)

  3. Process for manufacturing boron carbide pellets that can be used for the realization, of the control rods of water reactors

    International Nuclear Information System (INIS)

    Ballagny, Alain; Brie, Michel.

    1982-01-01

    The subject of the invention is a process for manufacturing boron carbide pellets with a boron carbide content of not less than 68% by volume and having an open porosity. This process consists in (a) preparing a mix comprising boron carbide powder of which at least 90% of the particles are under 3 μ in size, and an organic binder that can be transformed into carbon by thermal treatment, (b) compressing the hot mix thus obtained to form unbaked pellets, under a pressure of 1000 to 6000 bars, at a temperature of 80 to 250 0 C and (c) submitting the unbaked pellets thus obtained to vacuum thermal treatment to transform this binder into porous carbon. The finished pellets are used in the control rods of water reactors [fr

  4. Design, Fabrication and Performance of Boron-Carbide Control Elements

    International Nuclear Information System (INIS)

    Brammer, H.A.; Jacobson, J.

    1964-01-01

    A control blade design, incorporating boron-carbide (B 4 C) in stainless-steel tubes, was introduced into service in boiling water reactors in April 1961. Since that time this blade has become the standard reference control element in General Electric boiling-water reactors, replacing the 2% boron-stainless-steel blades previously used. The blades consist of a sheathed, cruciform array of small vertical stainless-steel tubes filled with compácted boron-carbide powder. The boron-carbide powder is confined longitudinally into several independent compartments by swaging over ball bearings located inside the tubes. The development and use of boron-carbide control rods is discussed in five phases: 1. Summary of experience with boron-steel blades and reasons for transition to boron-carbide control; 2. Design of the boron-carbide blade, beginning with developmental experiments, including early measurements performed in the AEC ''Control Rod Material and Development Program'' at the Vallecitos Atomic Laboratory, through a description of the final control blade configuration; 3. Fabrication of the blades and quality control procedures; 4. Results of confirmatory pre-operational mechanical and reactivity testing; and 5. Post-operational experience with the blades, including information on the results of mechanical inspection and reactivity testing after two years of reactor service. (author) [fr

  5. CODEX-B4C experiment. Core degradation test with boron carbide control rod

    International Nuclear Information System (INIS)

    Hozer, Z.; Nagy, I.; Windberg, P.; Balasko, M.; Matus, L.; Prokopiev, O.; Pinter, A.; Horvath, M.; Gyenes, Gy.; Czitrovszky, A.; Nagy, A.; Jani, P.

    2003-11-01

    The CODEX-B4C bundle test has been successfully performed on 25 th May 2001 in the framework of the COLOSS project of the EU 5 th FWP. The high temperature degradation of a VVER-1000 type bundle with B 4 C control rod was investigated with electrically heated fuel rods. The experiment was carried out according to a scenario selected in favour of methane formation. Degradation of control rod and fuel bundle took place at temperatures ∼2000 deg C, cooling down of the bundle was performed in steam atmosphere. The gas composition measurement indicated no methane production during the experiment. High release of aerosols was detected in the high temperature oxidation phase. The on-line measured data are collected into a database and are available for code validation and development. (author)

  6. CODEX-B4C experiment. Core degradation test with boron carbide control rod

    Energy Technology Data Exchange (ETDEWEB)

    Hozer, Z; Nagy, I; Windberg, P; Balasko, M; Matus, L; Prokopiev, O; Pinter, A; Horvath, M; Gyenes, Gy [KFKI Atomic Energy Research Institute, Budapest (Hungary); Czitrovszky, A; Nagy, A; Jani, P [Research Institute for Solid State Physics and Optics, Budapest (Hungary)

    2003-11-01

    The CODEX-B4C bundle test has been successfully performed on 25{sup th} May 2001 in the framework of the COLOSS project of the EU 5{sup th} FWP. The high temperature degradation of a VVER-1000 type bundle with B{sub 4}C control rod was investigated with electrically heated fuel rods. The experiment was carried out according to a scenario selected in favour of methane formation. Degradation of control rod and fuel bundle took place at temperatures {approx}2000 deg C, cooling down of the bundle was performed in steam atmosphere. The gas composition measurement indicated no methane production during the experiment. High release of aerosols was detected in the high temperature oxidation phase. The on-line measured data are collected into a database and are available for code validation and development. (author)

  7. Development of a process to recover boron carbide from nuclear reactor absorber rods

    International Nuclear Information System (INIS)

    Roth, C.; Lehnert, T.

    1991-01-01

    Boron carbide enriched with 10 B is used as a control rod in reactor engineering. At present spent rods are disposed of, although major amounts of 10 B are still 'unused'. The objective was to recover 10 B from the control rods by an energy and cost saving method in order to use it for making new control rods, thus saving raw materials and minimizing the radioactive waste volume. For this purpose, the well-known pyrohydrolysis process was taken and analysed for possible improvements. By mixing boron carbide with CO 2 as an oxidation-supporting agent, a lowering of the reaction temperature by 300deg C, and an increase in the oxidation speed by 350% were achieved. Since C0 2 is not consumed and can be circulated, the method for reprocessing spent control rods presented in this paper is both an economy-priced an energy-saving one. (orig.) With 98 refs., 9 tabs., 14 figs [de

  8. Helium diffusion in irradiated boron carbide

    International Nuclear Information System (INIS)

    Hollenberg, G.W.

    1981-03-01

    Boron carbide has been internationally adopted as the neutron absorber material in the control and safety rods of large fast breeder reactors. Its relatively large neutron capture cross section at high neutron energies provides sufficient reactivity worth with a minimum of core space. In addition, the commercial availability of boron carbide makes it attractive from a fabrication standpoint. Instrumented irradiation experiments in EBR-II have provided continuous helium release data on boron carbide at a variety of operating temperatures. Although some microstructural and compositional variations were examined in these experiments most of the boron carbide was prototypic of that used in the Fast Flux Test Facility. The density of the boron carbide pellets was approximately 92% of theoretical. The boron carbide pellets were approximately 1.0 cm in diameter and possessed average grain sizes that varied from 8 to 30 μm. Pellet centerline temperatures were continually measured during the irradiation experiments

  9. 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.)

  10. 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

  11. Shock Response of Boron Carbide

    National Research Council Canada - National Science Library

    Dandekar, D. P. (Dattatraya Purushottam)

    2001-01-01

    .... The present work was undertaken to determine tensile/spall strength of boron carbide under plane shock wave loading and to analyze all available shock compression data on boron carbide materials...

  12. 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)

  13. Thermal expansion measurements on boron carbide and europium sesquioxide by laser interferometry

    International Nuclear Information System (INIS)

    Preston, S.D.

    1980-01-01

    A laser interferometer technique for measuring the absolute linear thermal expansion of small annular specimens is described. Results are presented for unirradiated boron carbide (B 4 C) and europia (Eu 2 O 3 ) up to 1000 0 C. Both compounds are neutron-absorbing materials of potential use in fast-reactor control rods and data on their thermophysical properties, in particular linear thermal expansion, are essential to the control rod designers. (author)

  14. Electron microscopy study of radiation effects in boron carbide

    International Nuclear Information System (INIS)

    Stoto, T.

    1987-03-01

    Boron carbide is a disordered non-stoechiometric material with a strongly microtwinned polycristallyne microstructure. This ceramic is among the candidate materials for the first wall coating in fusion reactor and is used as a neutron absorber in the control rods of fast breeder reactors. The present work deals with the nature of radiation damage in this solid. Because of helium internal production, neutron irradiated boron carbide is affected by swelling and by a strong microcracking which can break up a pellet in fine powder. These processes are rather intensitive to the irradiation parameters (temperature, flux and even neutron spectrum). Transmission electron microscopy of samples irradiated by the fast neutrons of a reactor, the electrons of a high voltage electron microscope and of samples implanted with helium ions was used to understand the respective roles of helium and point defects in the processes of swelling and microcracking. The design of an irradiation chamber for helium implantation at controlled temperature from 600 to 1700 0 C was an important technical part of this work [fr

  15. Oxidation of boron carbide at high temperatures

    International Nuclear Information System (INIS)

    Steinbrueck, Martin

    2005-01-01

    The oxidation kinetics of various types of boron carbides (pellets, powder) were investigated in the temperature range between 1073 and 1873 K. Oxidation rates were measured in transient and isothermal tests by means of mass spectrometric gas analysis. Oxidation of boron carbide is controlled by the formation of superficial liquid boron oxide and its loss due to the reaction with surplus steam to volatile boric acids and/or direct evaporation at temperatures above 1770 K. The overall reaction kinetics is paralinear. Linear oxidation kinetics established soon after the initiation of oxidation under the test conditions described in this report. Oxidation is strongly influenced by the thermohydraulic boundary conditions and in particular by the steam partial pressure and flow rate. On the other hand, the microstructure of the B 4 C samples has a limited influence on oxidation. Very low amounts of methane were produced in these tests

  16. 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

  17. New Icosahedral Boron Carbide Semiconductors

    Science.gov (United States)

    Echeverria Mora, Elena Maria

    Novel semiconductor boron carbide films and boron carbide films doped with aromatic compounds have been investigated and characterized. Most of these semiconductors were formed by plasma enhanced chemical vapor deposition. The aromatic compound additives used, in this thesis, were pyridine (Py), aniline, and diaminobenzene (DAB). As one of the key parameters for semiconducting device functionality is the metal contact and, therefore, the chemical interactions or band bending that may occur at the metal/semiconductor interface, X-ray photoemission spectroscopy has been used to investigate the interaction of gold (Au) with these novel boron carbide-based semiconductors. Both n- and p-type films have been tested and pure boron carbide devices are compared to those containing aromatic compounds. The results show that boron carbide seems to behave differently from other semiconductors, opening a way for new analysis and approaches in device's functionality. By studying the electrical and optical properties of these films, it has been found that samples containing the aromatic compound exhibit an improvement in the electron-hole separation and charge extraction, as well as a decrease in the band gap. The hole carrier lifetimes for each sample were extracted from the capacitance-voltage, C(V), and current-voltage, I(V), curves. Additionally, devices, with boron carbide with the addition of pyridine, exhibited better collection of neutron capture generated pulses at ZERO applied bias, compared to the pure boron carbide samples. This is consistent with the longer carrier lifetimes estimated for these films. The I-V curves, as a function of external magnetic field, of the pure boron carbide films and films containing DAB demonstrate that significant room temperature negative magneto-resistance (> 100% for pure samples, and > 50% for samples containing DAB) is possible in the resulting dielectric thin films. Inclusion of DAB is not essential for significant negative magneto

  18. Boron-carbide-aluminum and boron-carbide-reactive metal cermets. [B/sub 4/C-Al

    Science.gov (United States)

    Halverson, D.C.; Pyzik, A.J.; Aksay, I.A.

    1985-05-06

    Hard, tough, lighweight boron-carbide-reactive metal composites, particularly boron-carbide-aluminum composites, are produced. These composites have compositions with a plurality of phases. A method is provided, including the steps of wetting and reacting the starting materials, by which the microstructures in the resulting composites can be controllably selected. Starting compositions, reaction temperatures, reaction times, and reaction atmospheres are parameters for controlling the process and resulting compositions. The ceramic phases are homogeneously distributed in the metal phases and adhesive forces at ceramic-metal interfaces are maximized. An initial consolidated step is used to achieve fully dense composites. Microstructures of boron-carbide-aluminum cermets have been produced with modules of rupture exceeding 110 ksi and fracture toughness exceeding 12 ksi..sqrt..in. These composites and methods can be used to form a variety of structural elements.

  19. Elastic modulus and fracture of boron carbide

    International Nuclear Information System (INIS)

    Hollenberg, G.W.; Walther, G.

    1978-12-01

    The elastic modulus of hot-pressed boron carbide with 1 to 15% porosity was measured at room temperature. K/sub IC/ values were determined for the same porosity range at 500 0 C by the double torsion technique. The critical stress intensity factor of boron carbide with 8% porosity was evaluated from 25 to 1200 0 C

  20. Material operating behaviour of ABB BWR control rods

    International Nuclear Information System (INIS)

    Rebensdorff, B.; Bart, G.

    2000-01-01

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

  1. 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.)

  2. Control rod

    International Nuclear Information System (INIS)

    Kawakami, Kazuo; Shimoshige, Takanori; Nishimura, Akira

    1979-01-01

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

  3. Preparation of fiber reinforced titanium diboride and boron carbide composite bodies

    International Nuclear Information System (INIS)

    Newkirk, L.R.; Riley, R.E.; Sheinberg, H.; Valencia, F.A.; Wallace, T.C.

    1979-01-01

    A process is described for uniformly infiltrating woven carbon cloth with either titanium diboride or boron carbide at reduced pressure (15 to 25 torr). The effects of deposition temperature on the uniformity of penetration and on coating rate are described for temperatures from 750 to 1000 0 C and deposit loadings from 20 to 43 vol. %. For the boron carbides, boron composition is discussed and evidence is presented suggesting that propene is the dominant rate controlling reactant

  4. 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.)

  5. Degradation in steam of 60 cm-long B{sub 4}C control rods

    Energy Technology Data Exchange (ETDEWEB)

    Dominguez, C., E-mail: christina.dominguez@irsn.fr; Drouan, D.

    2014-08-01

    In the framework of nuclear reactor core meltdown accident studies, the degradation of boron carbide control rod segments exposed to argon/steam atmospheres was investigated up to about 2000 °C in IRSN laboratories. The sequence of the phenomena involved in the degradation has been found to take place as expected. Nevertheless, the ZrO{sub 2} oxide layer formed on the outer surface of the guide tube was very protective, significantly delaying and limiting the guide tube failure and therefore the boron carbide pellet oxidation. Contrary to what was expected, the presence of the control rod decreases the hydrogen release instead of increasing it by additional oxidation of boron compounds. Boron contents up to 20 wt.% were measured in metallic mixtures formed during degradation. It was observed that these metallic melts are able to attack the surrounding fuel rods, which could have consequences on fuel degradation and fission product release kinetics during severe accidents.

  6. Determination of free carbon content in boron carbide ceramic powders

    International Nuclear Information System (INIS)

    Castro, A.R.M. de; Lima, N.B. de; Paschoal, J.O.A.

    1990-01-01

    Boron carbide is a ceramic material of technological importance due to its hardness and high chemical and thermal stabilities. Free carbon is always found as a process dependent impurity in boron carbide. The development of procedures for its detection is required because its presence leads to a degradation of the boron carbide properties. In this work, several procedures for determining free carbon content in boron carbide specimens are reported and discussed for comparison purposes. (author) [pt

  7. The estimation of the control rods absorber burn-up during the VVER-1000 operation

    Energy Technology Data Exchange (ETDEWEB)

    Bolshagin, Sergey N.; Gorodkov, Sergey S.; Sukhino-Khomenko, Evgeniya A. [National Research Centre ' Kurchatov Institute' , Moscow (Russian Federation)

    2013-09-15

    The isotopic composition of the control rods absorber changes under the neutron flux influence, so the control rods efficiency can decrease. In the VVER-1000 control rods boron carbide and dysprosium titanate are used as absorbing materials. In boric part the efficiency decreases due to the {sup 10}B isotope burn-up. Dysprosium isotopes turn into other absorbing isotopes, so the absorbing properties of dysprosium part decrease to a lesser degree. Also the control rod's shells may be deformed as a consequence of boron carbide radiation swelling. This fact should be considered in substantiation of control rods durability. For the estimation of the control rods absorber burn-up two models are developed: VVER-1000 3-D fuel assembly with control rods partially immersed (imitation of the control rods operation in the working group) and VVER-1000 3-D fuel assembly with control rods, located at the upper limit switch (imitation of the control rods operation in groups of the emergency shutdown system). (orig.)

  8. French LMFBR's control rods experience and development

    International Nuclear Information System (INIS)

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

    1983-06-01

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

  9. Anomalous Seebeck coefficient in boron carbides

    International Nuclear Information System (INIS)

    Aselage, T.L.; Emin, D.; Wood, C.; Mackinnon, I.D.R.; Howard, I.A.

    1987-01-01

    Boron carbides exhibit an anomalously large Seebeck coefficient with a temperature coefficient that is characteristic of polaronic hopping between inequivalent sites. The inequivalence in the sites is associated with disorder in the solid. The temperature dependence of the Seebeck coefficient for materials prepared by different techniques provides insight into the nature of the disorder

  10. Joining of boron carbide using nickel interlayer

    International Nuclear Information System (INIS)

    Vosughi, A.; Hadian, A. M.

    2008-01-01

    Carbide ceramics such as boron carbide due to their unique properties such as low density, high refractoriness, and high strength to weight ratio have many applications in different industries. This study focuses on direct bonding of boron carbide for high temperature applications using nickel interlayer. The process variables such as bonding time, temperature, and pressure have been investigated. The microstructure of the joint area was studied using electron scanning microscope technique. At all the bonding temperatures ranging from 1150 to 1300 d eg C a reaction layer formed across the ceramic/metal interface. The thickness of the reaction layer increased by increasing temperature. The strength of the bonded samples was measured using shear testing method. The highest strength value obtained was about 100 MPa and belonged to the samples bonded at 1250 for 75 min bonding time. The strength of the joints decreased by increasing the bonding temperature above 1250 d eg C . The results of this study showed that direct bonding technique along with nickel interlayer can be successfully utilized for bonding boron carbide ceramic to itself. This method may be used for bonding boron carbide to metals as well.

  11. Reaction of boron carbide with molybdenum disilicide

    International Nuclear Information System (INIS)

    Novikov, A.V.; Melekhin, V.F.; Pegov, V.S.

    1989-01-01

    The investigation results of interaction in the B 4 C-MoSi 2 system during sintering in vacuum are presented. Sintering of boron carbide with molybdenum disilicide is shown to lead to the formation of MoB 2 , SiC, Mo 5 Si 3 compounds, the presence of carbon-containing covering plays an important role in sintering

  12. Study of boron carbide evolution under neutron irradiation

    International Nuclear Information System (INIS)

    Simeone, D.

    1999-01-01

    Owing to its high neutron efficiency, boron carbide (B 4 C) is used as a neutron absorber in control rods of nuclear plants. Its behaviour under irradiation has been extensively studied for many years. It now seems clear that brittleness of the material induced by the 10 B(n,α) 7 Li capture reaction is due to penny shaped helium bubbles associated to a high strain field around them. However, no model explains the behaviour of the material under neutron irradiation. In order to build such a model, this work uses different techniques: nuclear microprobe X-ray diffraction profile analysis and Raman and Nuclear Magnetic Resonance Spectroscopy to present an evolution model of B 4 C under neutron irradiation. The use of nuclear reactions produced by a nuclear microprobe such as the 7 Li(p,p'γ) 7 Li reaction, allows to measure lithium profile in B 4 C pellets irradiated either in Pressurised Water Reactors or in Fast Breeder Reactors. Examining such profiles enables us to describe the migration of lithium atoms out of B 4 C materials under neutron irradiation. The analysis of X-ray diffraction profiles of irradiated B 4 C samples allows us to quantify the concentrations of helium bubbles as well as the strain fields around such bubbles.Furthermore Raman spectroscopy studies of different B 4 C samples lead us to propose that under neutron irradiation. the CBC linear chain disappears. Such a vanishing of this CBC chain. validated by NMR analysis, may explain the penny shaped of helium bubbles inside irradiated B 4 C. (author)

  13. Compatibility of heat resistant alloys with boron carbide, 5

    International Nuclear Information System (INIS)

    Baba, Shinichi; Kurasawa, Toshimasa; Endow, Taichi; Someya, Hiroyuki; Tanaka, Isao.

    1986-08-01

    This paper includes an experimental result of out-of-pile compatibility and capsule design for irradiation test in Japan Materials Testing Reactor (JMTR). The compatibility between sheath material and neutron absorber materials for control rod devices (CRD) was examined for potential use in a very high temperature reactor (VHTR) which is under development at JAERI. The purpose of the compatibility tests are preliminary evaluation of safety prior to irradiation tests. Preliminary compatibility evaluation was concerned with three items as follows : 1) Lithium effects on the penetrating reaction of Incoloy 800H alloy in contact with a mixture of boronated graphite and lithium hydroxide powders, 2) Short term tensile properties of Incoloy 800H and Hastelloy XR alloy reacted with boronated graphite and fracture mode analysis, 3) Reaction behavior of both alloys under transient power conditions of a VHTR. It was clear that the reaction rate constant of the Incoloy 800H alloy was accelerated by doping lithium hydroxide into the boron carbide and graphite powder. The mechanical properties of Incoloy 800H and Hastelloy XR alloy reacted with boronated graphite were decreased. Ultimate tensile strength and tensile ductilities at temperatures over 850 deg C were reduced, but there was no change in the proof (yield) stress. Both alloys exhibited a brittle intergranular fracture mode during transient power conditions of a VHTR and also exhibited severe penetration. Irradiation capsules for compatibility test were designed to simulate three irradiation conditions of VHTR: 1) steady state for VHTR, 2) Transient power condition, 3) Service limited life of CRD. Capsule irradiation experiments have been carried out satisfactorily and thus confirm the validity of the capsule design procedure. (author)

  14. In search of amorphization-resistant boron carbide

    International Nuclear Information System (INIS)

    Subhash, Ghatu; Awasthi, Amnaya P.; Kunka, Cody; Jannotti, Phillip; DeVries, Matthew

    2016-01-01

    Despite its superior mechanical properties, boron carbide suffers from amorphization, a pressure-induced phenomenon that disturbs crystalline order and likely reduces shear strength. Numerous experimental and computational studies have investigated the structure and origins of amorphization, yet strategies to mitigate this deleterious phenomenon elude. However, recent investigations have revealed three new research avenues for addressing this issue. First, we identify crystallographic cage spaces that may accommodate foreign atoms to potentially prevent structural collapse. Second, we propose polymorph-level tailoring through strict control of processing conditions. Finally, we demonstrate that reducing grain size to nanometer scale increases hardness and may counter amorphization.

  15. Design, Fabrication and Performance of Boron-Carbide Control Elements; Conception, Fabrication et Comportement de Lames de Commande en Carbure de Bore; Raschety, izgotovlenie i kharakteristiki reguliruyushchikh sterzhnej. Iz karbida Bora; Proyecto, Elaboracion y Rendimiento de Elementos de Control de Carburo de Boro

    Energy Technology Data Exchange (ETDEWEB)

    Brammer, H. A.; Jacobson, J. [General Electric Company, San Jose, CA (United States)

    1964-06-15

    A control blade design, incorporating boron-carbide (B{sub 4}C) in stainless-steel tubes, was introduced into service in boiling water reactors in April 1961. Since that time this blade has become the standard reference control element in General Electric boiling-water reactors, replacing the 2% boron-stainless-steel blades previously used. The blades consist of a sheathed, cruciform array of small vertical stainless-steel tubes filled with compacted boron-carbide powder. The boron-carbide powder is confined longitudinally into several independent compartments by swaging over ball bearings located inside the tubes. The development and use of boron-carbide control rods is discussed in five phases: 1. Summary of experience with boron-steel blades and reasons for transition to boron-carbide control; 2. Design of the boron-carbide blade, beginning with developmental experiments, including early measurements performed in the AEC ''Control Rod Material and Development Program'' at the Vallecitos Atomic Laboratory, through a description of the final control blade configuration; 3. Fabrication of the blades and quality control procedures; 4. Results of confirmatory pre-operational mechanical and reactivity testing; and 5. Post-operational experience with the blades, including information on the results of mechanical inspection and reactivity testing after two years of reactor service. (author) [French] Un modele de lame de commande en carbure de bore (B{sub 4}C) a ete mis en place dans des reacteurs a eau bouillante en avril 1961. Depuis lors, cette lame est devenue l 'element de commande temoin classique dans les reacteurs a eau bouillante de la General Electric et a remplace les lames en acier inoxydable a 2% de bore utilisees auparavant. Ces lames consistent en un assemblage gaine cruciforme comprenant de petits tubes d'acier inoxydable verticaux remplis de poudre de carbure de bore agglomeree. Dans le sens de la longueur, cette poudre est enfermee dans plusieurs

  16. High resolution imaging of boron carbide microstructures

    International Nuclear Information System (INIS)

    MacKinnon, I.D.R.; Aselage, T.; Van Deusen, S.B.

    1986-01-01

    Two samples of boron carbide have been examined using high resolution transmission electron microscopy (HRTEM). A hot-pressed B 13 C 2 sample shows a high density of variable width twins normal to (10*1). Subtle shifts or offsets of lattice fringes along the twin plane and normal to approx.(10*5) were also observed. A B 4 C powder showed little evidence of stacking disorder in crystalline regions

  17. Control rod

    International Nuclear Information System (INIS)

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

    1988-01-01

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

  18. Boron Carbide: Stabilization of Highly-Loaded Aqueous Suspensions, Pressureless Sintering, and Room Temperature Injection Molding

    Science.gov (United States)

    Diaz-Cano, Andres

    Boron carbide (B4C) is the third hardest material after diamond and cubic boron nitride. It's unique combination of properties makes B4C a highly valuable material. With hardness values around 35 MPa, a high melting point, 2450°C, density of 2.52 g/cm3, and high chemical inertness, boron carbide is used in severe wear components, like cutting tools and sandblasting nozzles, nuclear reactors' control rots, and finally and most common application, armor. Production of complex-shaped ceramic component is complex and represents many challenges. Present research presents a new and novel approach to produce complex-shaped B4C components. Proposed approach allows forming to be done at room temperatures and under very low forming pressures. Additives and binder concentrations are kept as low as possible, around 5Vol%, while ceramics loadings are maximized above 50Vol%. Given that proposed approach uses water as the main solvent, pieces drying is simple and environmentally safe. Optimized formulation allows rheological properties to be tailored and adjust to multiple processing approaches, including, injection molding, casting, and additive manufacturing. Boron carbide samples then were pressureless sintered. Due to the high covalent character of boron carbide, multiples sintering aids and techniques have been proposed in order to achieve high levels of densification. However, is not possible to define a clear sintering methodology based on literature. Thus, present research developed a comprehensive study on the effect of multiple sintering aids on the densification of boron carbide when pressureless sintered. Relative densities above 90% were achieved with values above 30MPa in hardness. Current research allows extending the uses and application of boron carbide, and other ceramic systems, by providing a new approach to produce complex-shaped components with competitive properties.

  19. B4C control rod behavior during severe accident sequences

    International Nuclear Information System (INIS)

    Steinbrueck, M.

    2003-01-01

    The oxidation kinetics of various types of boron carbides (pellets, powder) as well as the degradation of B 4 C control rod segments were investigated in the temperature range between 800 and 1600 deg C. Mass spectrometric gas analysis was used to determine oxidation rates in transient and isothermal tests. The oxidation kinetics of boron carbide are determined by the formation of a liquid boron oxide layer and its loss due to the reaction with surplus steam to form volatile boric acids and at temperatures above 1500 deg C by direct evaporation. Under these test conditions linear oxidation kinetics are established soon after oxidation has initiated. The oxidation kinetics are strongly influenced by the thermal-hydraulic boundary conditions, in particular by the steam flow rate. Only very low amounts of methane were ever produced in these tests. Enhanced degradation of B 4 C control rods starts with the rapid formation of eutectic melts in the systems B 4 C-stainless steel (SS) and SS-Zircaloy at temperatures above 1250 deg C. Initially, this melt is kept within a ZrO 2 scale externally formed at the Zircaloy guide tube. The absorber melt is rapidly oxidized after failure of the oxide shell and aggressively attacks adjacent fuel claddings. (author)

  20. CONTROL ROD

    Science.gov (United States)

    Walker, D.E.; Matras, S.

    1963-04-30

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

  1. Control rods

    International Nuclear Information System (INIS)

    Koga, Isao; Masuoka, Ryuzo.

    1979-01-01

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

  2. Determination of free and combined carbon in boron carbide

    International Nuclear Information System (INIS)

    Shankaran, P.S.; Kulkarni, Amit S.; Pandey, K.L.; Ramanjaneyulu, P.S.; Yadav, C.S.; Sayi, Y.S.; Ramakumar, K.L.

    2009-01-01

    A simple, sensitive and fast method for the determination of free and combined carbon in boron carbide samples, based on combustion in presence of oxygen at different temperatures, has been developed. Method has been standardized by analyzing mixture of two different boron carbide samples. Error associated with the method in the determination of free carbon is less than 5%. (author)

  3. Accident-tolerant control rod

    International Nuclear Information System (INIS)

    Ohta, Hirokazu; Sawabe, Takashi; Ogata, Takanari

    2013-01-01

    Boron carbide (B 4 C) and hafnium (Hf) metal are used for the neutron absorber materials of control rods in BWRs, and silver-indium-cadmium (Ag-In-Cd) alloy is used in PWRs. These materials are clad with stainless steel. The eutectic point of B 4 C and iron (Fe) is about 1150 deg. C and the melting point of Ag-In-Cd alloy is about 800 deg. C, which are lower than the temperature of zircaloy - steam reaction increases rapidly (∼1200 deg. C). Accordingly, it is possible that the control rods melt and collapse before the reactor core is significantly damaged in the case of severe accidents. Since the neutron absorber would be separated from the fuels, there is a risk of re-criticality, when pure water or seawater is injected for emergency cooling. In order to ensure sub-criticality and extend options of emergency cooling in the course of severe accidents, a concept of accident-tolerant control rod (ACT) has been derived. ACT utilises a new absorber material having the following properties: - higher neutron absorption than current control rod; - higher melting or eutectic temperature than 1200 deg. C where rapid zircaloy oxidation occurs; - high miscibility with molten fuel materials. The candidate of a new absorber material for ATC includes gadolinia (Gd 2 O 3 ), samaria (Sm 2 O 3 ), europia (Eu 2 O 3 ), dysprosia (Dy 2 O 3 ), hafnia (HfO 2 ). The melting point of these materials and the liquefaction temperature with Fe are higher than the rapid zircaloy oxidation temperature. ACT will not collapse before the core melt-down. After the core melt-down, the absorber material will be mixed with molten fuel material. The current absorber materials, such as B 4 C, Hf and Ag-In-Cd, are charged at the tip of ATC in which the neutron flux is high, and a new absorber material is charged in the low-flux region. This design could minimise the degradation of a new absorber material by the neutron absorption and the influence of ATC deployment on reactor control procedure. As a

  4. Control rod

    International Nuclear Information System (INIS)

    Takeda, Toshikazu; Inoue, Kotaro.

    1979-01-01

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

  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. Method for fabricating boron carbide articles

    International Nuclear Information System (INIS)

    Ardary, Z.; Reynolds, C.

    1980-01-01

    Described is a method for fabricating an essentially uniformly dense boron carbide article of a length-to-diameter or width ratio greater than 2 to 1 comprising the steps of providing a plurality of article segments to be joined together to form the article with each of said article segments having a length-to-diameter or width ratio less than 1.5 to 1. Each segment is fabricated by hot pressing a composition consisting of boron carbide powder at a pressure and temperature effective to provide the article segment with a density greater than about 85% of theoretical density, providing each article segment with parallel planar end surfaces, placing a plurality of said article segments in a hot-pressing die in a line with the planar surfaces of adjacent article segments being disposed in intimate contact, and hot pressing the aligned article segments at a temperature and pressure effective to provide said article with a density over the length thereof in the range of about 94 to 98 percent theoretical density and greater than the density provided in the discrete hot pressing of each of the article segments and to provide a bond between adjacent article segments with said bond being at least equivalent in hardness, strength and density to a remainder of said article

  7. Graphite and boron carbide composites made by hot-pressing

    International Nuclear Information System (INIS)

    Miyazaki, K.; Hagio, T.; Kobayashi, K.

    1981-01-01

    Composites consisting of graphite and boron carbide were made by hot-pressing mixed powders of coke carbon and boron carbide. The change of relative density, mechanical strength and electrical resistivity of the composites and the X-ray parameters of coke carbon were investigated with increase of boron carbide content and hot-pressing temperature. From these experiments, it was found that boron carbide powder has a remarkable effect on sintering and graphitization of coke carbon powder above the hot-pressing temperature of 2000 0 C. At 2200 0 C, electrical resistivity of the composite and d(002) spacing of coke carbon once showed minimum values at about 5 to 10 wt% boron carbide and then increased. The strength of the composite increased with increase of boron carbide content. It was considered that some boron from boron carbide began to diffuse substitutionally into the graphite structure above 2000 0 C and densification and graphitization were promoted with the diffusion of boron. Improvements could be made to the mechanical strength, density, oxidation resistance and manufacturing methods by comparing with the properties and processes of conventional graphites. (author)

  8. Boron carbide in pile behaviour Rapsodie experience

    International Nuclear Information System (INIS)

    Kryger, B.; Colin, M.

    1983-04-01

    Results concerning boron carbide irradiation experiments performed in RAPSODIE up to 10 22 .cm - 3 capture density in the temperature range 600-1100 0 lead to the following main conclusions: initial density and grain size lowering contribute to swelling decrease but density is the major parameter for swelling limitation; swelling rate can vary in a wide range (ratio 1 to 3) according to combinations of density (1.8 to 2.3) and grain size (10 to 50 μm) values; a swelling balance reveals that the most important contribution to swelling should be a high density of helium small bubbles (<400 A); helium retention increases with density and grain size and decreases with temperature elevation. A diffusion law is proposed to describe the rate of helium release

  9. Helium behaviour in implanted boron carbide

    Directory of Open Access Journals (Sweden)

    Motte Vianney

    2015-01-01

    Full Text Available When boron carbide is used as a neutron absorber in nuclear power plants, large quantities of helium are produced. To simulate the gas behaviour, helium implantations were carried out in boron carbide. The samples were then annealed up to 1500 °C in order to observe the influence of temperature and duration of annealing. The determination of the helium diffusion coefficient was carried out using the 3He(d,p4He nuclear reaction (NRA method. From the evolution of the width of implanted 3He helium profiles (fluence 1 × 1015/cm2, 3 MeV corresponding to a maximum helium concentration of about 1020/cm3 as a function of annealing temperatures, an Arrhenius diagram was plotted and an apparent diffusion coefficient was deduced (Ea = 0.52 ± 0.11 eV/atom. The dynamic of helium clusters was observed by transmission electron microscopy (TEM of samples implanted with 1.5 × 1016/cm2, 2.8 to 3 MeV 4He ions, leading to an implanted slab about 1 μm wide with a maximum helium concentration of about 1021/cm3. After annealing at 900 °C and 1100 °C, small (5–20 nm flat oriented bubbles appeared in the grain, then at the grain boundaries. At 1500 °C, due to long-range diffusion, intra-granular bubbles were no longer observed; helium segregates at the grain boundaries, either as bubbles or inducing grain boundaries opening.

  10. Analytical chemistry methods for boron carbide absorber material. [Standard

    Energy Technology Data Exchange (ETDEWEB)

    DELVIN WL

    1977-07-01

    This standard provides analytical chemistry methods for the analysis of boron carbide powder and pellets for the following: total C and B, B isotopic composition, soluble C and B, fluoride, chloride, metallic impurities, gas content, water, nitrogen, and oxygen. (DLC)

  11. Control rods

    International Nuclear Information System (INIS)

    Hirukawa, Koji.

    1979-01-01

    Purpose: To ensure the fuel safety by constituting a control rod with a plurality of poison bodies suspended in a cross-like section and shorter length poison bodies made movable and engageable in the gap between each of the above poison bodies thereby maintaining the function of the shorter length poison constant. Constitution: Cross-like supports are secured to the upper and lower parts of a driving shaft journaled in a sheath and poison bodies composed of neutron absorber poisons of a large thermal neutron absorption cross section and neutron absorber poison tubes for containing them are suspended from the supports. A movable cross-like support is mounted slidably at its base to the lower part of the driving shaft and poison bodies shorter than the above poison bodies and composed of neutron absorber poisons having a greater absorption cross section at the neutron energy region higher than thermal neutron region and neutron poison tubes for containing them are suspended to the movable support at the position capable of engaging in the gap between each of the poison bodies. (Kawakami, Y.)

  12. Application of hafnium hydride control rod to large sodium cooled fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ikeda, Kazumi, E-mail: kazumi_ikeda@mfbr.mhi.co.jp [Mitsubishi FBR Systems, Inc., 34-17, Jingumae 2-Chome, Shibuya-ku, Tokyo 150-0001 (Japan); Moriwaki, Hiroyuki, E-mail: hiroyuki_moriwaki@mfbr.mhi.co.jp [Mitsubishi FBR Systems, Inc., 34-17, Jingumae 2-Chome, Shibuya-ku, Tokyo 150-0001 (Japan); Ohkubo, Yoshiyuki, E-mail: yoshiyuki_okubo@mfbr.mhi.co.jp [Mitsubishi FBR Systems, Inc., 34-17, Jingumae 2-Chome, Shibuya-ku, Tokyo 150-0001 (Japan); Iwasaki, Tomohiko, E-mail: tomohiko.iwasaki@qse.tohoku.ac.jp [Department of Quantum Science and Energy Engineering, Tohoku University, Aoba, Aramaki, Aoba-ku, Sendai-shi, Miyagi-ken 980-8579 (Japan); Konashi, Kenji, E-mail: konashi@imr.tohoku.ac.jp [Institute for Materials Research, Tohoku University, Narita-cho, Oarai-machi, Higashi-Ibaraki-gun, Ibaraki-ken 311-1313 (Japan)

    2014-10-15

    Highlights: • Application of hafnium hydride control rod to large sodium cooled fast breeder reactor. • This paper treats application of an innovative hafnium hydride control rod to a large sodium cooled fast breeder reactor. • Hydrogen absorption triples the reactivity worth by neutron spectrum shift at H/Hf ratio of 1.3. • Lifetime of the control rod quadruples because produced daughters of hafnium isotopes are absorbers. • Nuclear and thermal hydraulic characteristics of the reactor are as good as or better than B-10 enriched boron carbide. - Abstract: This study treats the feasibility of long-lived hafnium hydride control rod in a large sodium-cooled fast breeder reactor by nuclear and thermal analyses. According to the nuclear calculations, it is found that hydrogen absorption of hafnium triples the reactivity by the neutron spectrum shift at the H/Hf ratio of 1.3, and a hafnium transmutation mechanism that produced daughters are absorbers quadruples the lifetime due to a low incineration rate of absorbing nuclides under irradiation. That is to say, the control rod can function well for a long time because an irradiation of 2400 EFPD reduces the reactivity by only 4%. The calculation also reveals that the hafnium hydride control rod can apply to the reactor in that nuclear and thermal characteristics become as good as or better than 80% B-10 enriched boron carbide. For example, the maximum linear heat rate becomes 3% lower. Owing to the better power distribution, the required flow rate decreases approximately by 1%. Consequently, it is concluded on desk analyses that the long lived hafnium hydride control rod is feasible in the large sodium-cooled fast breeder reactor.

  13. Highly thermal conductive carbon fiber/boron carbide composite material

    International Nuclear Information System (INIS)

    Chiba, Akio; Suzuki, Yasutaka; Goto, Sumitaka; Saito, Yukio; Jinbo, Ryutaro; Ogiwara, Norio; Saido, Masahiro.

    1996-01-01

    In a composite member for use in walls of a thermonuclear reactor, if carbon fibers and boron carbide are mixed, since they are brought into contact with each other directly, boron is reacted with the carbon fibers to form boron carbide to lower thermal conductivity of the carbon fibers. Then, in the present invention, graphite or amorphous carbon is filled between the carbon fibers to provide a fiber bundle of not less than 500 carbon fibers. Further, the surface of the fiber bundle is coated with graphite or amorphous carbon to suppress diffusion or solid solubilization of boron to carbon fibers or reaction of them. Then, lowering of thermal conductivity of the carbon fibers is prevented, as well as the mixing amount of the carbon fiber bundles with boron carbide, a sintering temperature and orientation of carbon fiber bundles are optimized to provide a highly thermal conductive carbon fiber/boron carbide composite material. In addition, carbide or boride type short fibers, spherical graphite, and amorphous carbon are mixed in the boron carbide to prevent development of cracks. Diffusion or solid solubilization of boron to carbon fibers is reduced or reaction of them if the carbon fibers are bundled. (N.H.)

  14. Synergistic methods for the production of high-strength and low-cost boron carbide

    Science.gov (United States)

    Wiley, Charles Schenck

    2011-12-01

    Boron carbide (B4C) is a non-oxide ceramic in the same class of nonmetallic hard materials as silicon carbide and diamond. The high hardness, high elastic modulus and low density of B4C make it a nearly ideal material for personnel and vehicular armor. B4C plates formed via hot-pressing are currently issued to U.S. soldiers and have exhibited excellent performance; however, hot-pressed articles contain inherent processing defects and are limited to simple geometries such as low-curvature plates. Recent advances in the pressureless sintering of B4C have produced theoretically-dense and complex-shape articles that also exhibit superior ballistic performance. However, the cost of this material is currently high due to the powder shape, size, and size distribution that are required, which limits the economic feasibility of producing such a product. Additionally, the low fracture toughness of pure boron carbide may have resulted in historically lower transition velocities (the projectile velocity range at which armor begins to fail) than competing silicon carbide ceramics in high-velocity long-rod tungsten penetrator tests. Lower fracture toughness also limits multi-hit protection capability. Consequently, these requirements motivated research into methods for improving the densification and fracture toughness of inexpensive boron carbide composites that could result in the development of a superior armor material that would also be cost-competitive with other high-performance ceramics. The primary objective of this research was to study the effect of titanium and carbon additives on the sintering and mechanical properties of inexpensive B4C powders. The boron carbide powder examined in this study was a sub-micron (0.6 mum median particle size) boron carbide powder produced by H.C. Starck GmbH via a jet milling process. A carbon source in the form of phenolic resin, and titanium additives in the form of 32 nm and 0.9 mum TiO2 powders were selected. Parametric studies of

  15. Estimation of dose rate around the spent control rods of a BWR

    International Nuclear Information System (INIS)

    Cancino P, G.

    2016-01-01

    The energy can come from fossil renewable sources (solar (natural gas, oil), wind, hydro, tidal, geothermal, biomass, bio energy and nuclear. Nuclear power can be obtained by fission reactions and fusion (still under investigation) atomic nuclei. Fission, is a partition of a very heavy nucleus (Uranium 235, for example) into two lighter nuclei. Much of the world's electric power is generated from the energy released by fission processes. In a nuclear power reactor, light water as the BWR, there are many important elements that allow safe driving operation, one of them are the elements or control systems, the burnable poison or neutron absorber inherently allow control power reactor. The control rods, which consist mostly of stainless steel and absorbing elements (such as boron carbide, hafnium, cadmium, among others) of thermal neutrons is able to initiate, regulate or stop the reactor power. These, due to the use of depleted burned or absorbing material and therefore reach their lifespan, which can be 15 years or have other values depending on the manufacturer. Control rods worn should be removed, stored or confined in expressly places. Precisely at this stage arises the importance of knowing their radiological condition to manipulate safely and without incident to the people health responsible for conducting these proceedings state arises. This thesis consists in the estimation of the dose rate in spent control rod made of boron carbide, from a typical BWR reactor. It will be estimated by direct radiation measurements with measurement equipment for radiotherapy ionization chamber, in six spent control rods, which were taken at different reactor operating cycles and are in a spent fuel pool. Using bracket electromechanical and electronic equipment for positioning and lifting equipment for radiation measurement around the control rod in the axial and radial arrangement for proper scanning. Finally will be presented a graphic corresponding to the dose

  16. 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.)

  17. Control rod assembly

    International Nuclear Information System (INIS)

    Takahashi, Akio.

    1982-01-01

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

  18. 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

  19. Electron microscopy of boron carbide before and after electron irradiation

    International Nuclear Information System (INIS)

    Stoto, T.; Zuppiroli, L.; Beauvy, M.; Athanassiadis, T.

    1984-06-01

    The microstructure of boron carbide has been studied by electron microscopy and related to the composition of the material. After electron irradiations in an usual transmission electron microscope and in a high voltage electron microscope at different temperatures and fluxes no change of these microstructures have been observed but a sputtering of the surface of the samples, which has been studied quantitatively [fr

  20. Production of boron carbide powder by carbothermal synthesis of ...

    Indian Academy of Sciences (India)

    TECS

    weight armour plates etc (Alizadeh et al 2004). It can also be used as a reinforcing material for ceramic matrix composites. It is an excellent neutron absorption material in nuclear industry due to its high neutron absorption co- efficient (Sinha et al 2002). Boron carbide can be prepared by reaction of elemental boron and ...

  1. Functionalization and cellular uptake of boron carbide nanoparticles

    DEFF Research Database (Denmark)

    Mortensen, M. W.; Björkdahl, O.; Sørensen, P. G.

    2006-01-01

    In this paper we present surface modification strategies of boron carbide nanoparticles, which allow for bioconjugation of the transacting transcriptional activator (TAT) peptide and fluorescent dyes. Coated nanoparticles can be translocated into murine EL4 thymoma cells and B16 F10 malignant...

  2. 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.)

  3. Structure and single-phase regime of boron carbides

    International Nuclear Information System (INIS)

    Emin, D.

    1988-01-01

    The boron carbides are composed of twelve-atom icosahedral clusters which are linked by direct covalent bonds and through three-atom intericosahedral chains. The boron carbides are known to exist as a single phase with carbon concentrations from about 8 to about 20 at. %. This range of carbon concentrations is made possible by the substitution of boron and carbon atoms for one another within both the icosahedra and intericosahedral chains. The most widely accepted structural model for B 4 C (the boron carbide with nominally 20% carbon) has B/sub 11/C icosahedra with C-B-C intericosahedral chains. Here, the free energy of the boron carbides is studied as a function of carbon concentration by considering the effects of replacing carbon atoms within B 4 C with boron atoms. It is concluded that entropic and energetic considerations both favor the replacement of carbon atoms with boron atoms within the intericosahedral chains, C-B-C→C-B-B. Once the carbon concentration is so low that the vast majority of the chains are C-B-B chains, near B/sub 13/C 2 , subsequent substitutions of carbon atoms with boron atoms occur within the icosahedra, B/sub 11/C→B/sub 12/. Maxima of the free energy occur at the most ordered compositions: B 4 C,B/sub 13/C 2 ,B/sub 14/C. This structural model, determined by studying the free energy, agrees with that previously suggested by analysis of electronic and thermal transport data. These considerations also provide an explanation for the wide single-phase regime found for boron carbides

  4. Neutronic analysis of absorbing materials for the control rod system in reactor ALLEGRO

    Energy Technology Data Exchange (ETDEWEB)

    Cajko, Frantisek; Secansky, Michal; Chrebet, Tomas; Zajac, Radoslav; Darilek, Petr [VUJE, a.s., Trnava (Slovakia)

    2016-09-15

    Experimental reactor ALLEGRO is a gas cooled fast reactor in the design stage. The current design of its reactivity control system is based on control rods filled with boron carbide as the absorber. Because of disadvantages connected to high boron enrichment a possibility of using other absorbent materials was explored to lower the boron enrichment and increase the worth of the control rods. The results of neutronic Monte-Carlo analyses in a computational supercell are presented in this paper. Three absorbent materials most suitable for a use in reactor ALLEGRO (B{sub 4}C, EuB{sub 6} and ReB{sub 2}) have been analysed also in a full core model. A possible benefit of a neutron trap concept is explored as well but materials with satisfactory neutronic properties proved to be not suitable for expected high temperatures in the reactor.

  5. Control rod drive mechanism

    International Nuclear Information System (INIS)

    Nakamura, Akira.

    1981-01-01

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

  6. Control rod drive

    International Nuclear Information System (INIS)

    Hawke, B.C.

    1986-01-01

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

  7. Evaluation of Aluminum-Boron Carbide Neutron Absorbing Materials for Interim Storage of Used Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Lumin [Univ. of Michigan, Ann Arbor, MI (United States). Department of Nuclear Engineering and Radiological Science; Wierschke, Jonathan Brett [Univ. of Michigan, Ann Arbor, MI (United States). Department of Nuclear Engineering and Radiological Science

    2015-04-08

    The objective of this work was to understand the corrosion behavior of Boral® and Bortec® neutron absorbers over long-term deployment in a used nuclear fuel dry cask storage environment. Corrosion effects were accelerated by flowing humidified argon through an autoclave at temperatures up to 570°C. Test results show little corrosion of the aluminum matrix but that boron is leaching out of the samples. Initial tests performed at 400 and 570°C were hampered by reduced flow caused by the rapid build-up of solid deposits in the outlet lines. Analysis of the deposits by XRD shows that the deposits are comprised of boron trioxide and sassolite (H3BO3). The collection of boron- containing compounds in the outlet lines indicated that boron was being released from the samples. Observation of the exposed samples using SEM and optical microscopy show the growth of new phases in the samples. These phases were most prominent in Bortec® samples exposed at 570°C. Samples of Boral® exposed at 570°C showed minimal new phase formation but showed nearly the complete loss of boron carbide particles. Boron carbide loss was also significant in Boral samples at 400°C. However, at 400°C phases similar to those found in Bortec® were observed. The rapid loss of the boron carbide particles in the Boral® is suspected to inhibit the formation of the new secondary phases. However, Material samples in an actual dry cask environment would be exposed to temperatures closer to 300°C and less water than the lowest test. The results from this study conclude that at the temperature and humidity levels present in a dry cask environment, corrosion and boron leaching will have no effect on the performance of Boral® and Bortec® to maintain criticality control.

  8. Evaluation of Aluminum-Boron Carbide Neutron Absorbing Materials for Interim Storage of Used Nuclear Fuel

    International Nuclear Information System (INIS)

    Wang, Lumin; Wierschke, Jonathan Brett

    2015-01-01

    The objective of this work was to understand the corrosion behavior of Boral® and Bortec® neutron absorbers over long-term deployment in a used nuclear fuel dry cask storage environment. Corrosion effects were accelerated by flowing humidified argon through an autoclave at temperatures up to 570°C. Test results show little corrosion of the aluminum matrix but that boron is leaching out of the samples. Initial tests performed at 400 and 570°C were hampered by reduced flow caused by the rapid build-up of solid deposits in the outlet lines. Analysis of the deposits by XRD shows that the deposits are comprised of boron trioxide and sassolite (H 3 BO 3 ). The collection of boron- containing compounds in the outlet lines indicated that boron was being released from the samples. Observation of the exposed samples using SEM and optical microscopy show the growth of new phases in the samples. These phases were most prominent in Bortec® samples exposed at 570°C. Samples of Boral® exposed at 570°C showed minimal new phase formation but showed nearly the complete loss of boron carbide particles. Boron carbide loss was also significant in Boral samples at 400°C. However, at 400°C phases similar to those found in Bortec® were observed. The rapid loss of the boron carbide particles in the Boral® is suspected to inhibit the formation of the new secondary phases. However, Material samples in an actual dry cask environment would be exposed to temperatures closer to 300°C and less water than the lowest test. The results from this study conclude that at the temperature and humidity levels present in a dry cask environment, corrosion and boron leaching will have no effect on the performance of Boral® and Bortec® to maintain criticality control.

  9. Lattice dynamics of α boron and of boron carbide

    International Nuclear Information System (INIS)

    Vast, N.

    1999-01-01

    The atomic structure and the lattice dynamics of α boron and of B 4 C boron carbide have been studied by Density Functional Theory (D.F.T.) and Density Functional Perturbation Theory (D.F.P.T.). The bulk moduli of the unit-cell and of the icosahedron have been investigated, and the equation of state at zero temperature has been determined. In α boron, Raman diffusion and infrared absorption have been studied under pressure, and the theoretical and experimental Grueneisen coefficients have been compared. In boron carbide, inspection of the theoretical and experimental vibrational spectra has led to the determination of the atomic structure of B 4 C. Finally, the effects of isotopic disorder have been modeled by an exact method beyond the mean-field approximation, and the effects onto the Raman lines has been investigated. The method has been applied to isotopic alloys of diamond and germanium. (author)

  10. Epithermal neutron activation analysis using a boron carbide irradiation filter

    International Nuclear Information System (INIS)

    Ehmann, W.D.; Brueckner, J.

    1980-01-01

    The use of boron carbide as a thermal neutron filter in epithermal neutron activation (ENAA) analysis has been investigated. As compared to the use of a cadmium filter, boron provides a greater reduction of activities from elements relatively abundant in terrestrial rocks and fossil fuels, such as Na, La, Sc and Fe. These elements have excitation functions which follow the 1/v law in the 1 to 10 eV lower epithermal region. This enhances the sensitivity of ENAA for elements such as U, Th, Ba and etc. which have strong resonances in the higher epithermal region above 10 eV. In addition, a boron carbide filter has the advantages over cadmium of acquiring a relatively low level of induced activity which poses minimal radiation safety problems, when used for ENAA. (author)

  11. Electronic and vibrational hopping transport in boron carbides

    International Nuclear Information System (INIS)

    Emin, D.

    1991-01-01

    General concepts of hopping-type transport and localization are reviewed. Disorder, electronic correlations and atomic displacements, effects ignored in electronic band structure calculations, foster localization of electronic charge carriers. Examples are given that illustrate the efficacy of these effects in producing localization. This introduction is followed by a brief discussion of the relation between hopping-type transport and localization. The fundamentals of the formation, localization, and hopping transport of small polarons and/or bipolarons is then described. Electronic transport in boron carbides is presented as an example of the adiabatic hopping of small bipolarons. Finally, the notion of vibrational hopping is introduced. The high-temperature thermal diffusion in boron carbides is presented as a potential application of this idea

  12. Determination of soluble carbon in nuclear grade boron carbide

    International Nuclear Information System (INIS)

    Vega Bustillos, J.O.; Gomes, R.; Camaro, J.; Zorzetto, F.; Domingues, P.; Riella, H.

    1990-05-01

    The present work describes two different techniques (manometric and wet chemical) for the soluble carbon determination in nuclear grade boron carbide. The techniques are based on the reaction of the boron carbide with a sulfocromic mixture, generating CO 2 . The techniques differ on the mode they do the measurement of CO 2 produced. By wet chemical technique the CO 2 is absorved in a barium hydroxide solution and is determinated by titration. In the manometric technique the CO 2 gas is measured using a McLeod gauge. The gas produced by the latter technique is analysed by mass spectrometry. The details of the analytical technique and the data obtained are discussed. (author) [pt

  13. Fission reactor control rod

    International Nuclear Information System (INIS)

    Irie, Tomoo.

    1991-01-01

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

  14. Sodium erosion of boron carbide from breached absorber pins

    International Nuclear Information System (INIS)

    Basmajian, J.A.; Baker, D.E.

    1981-03-01

    The purpose of the irradiation experiment was to provide an engineering demonstration of the irradiation behavior of breached boron carbide absorber pins. By building defects into the cladding of prototypic absorber pins, and performing the irradiation under typical FFTF operating conditions, a qualitative assessment of the consequences of a breach was achieved. Additionally, a direct comparison of pin behavior with that of the ex-reactor test could be made

  15. Correlation for boron carbide helium release in fast reactors

    International Nuclear Information System (INIS)

    Basmajian, J.A.; Pitner, A.L.

    1977-04-01

    An empirical helium correlation for the helium release from boron carbide has been developed. The correlation provides a good fit to the experimental data in the temperature range from 800 to 1350 0 K, and burnup levels up to 80 x 10 20 captures/cm 3 . The correlation has the capability of extrapolation to 2200 0 K (3500 0 F) and 200 x 10 20 captures/cm 3 . In this range the helium release rate will not exceed the generation rate

  16. Disorder and defects are not intrinsic to boron carbide

    Science.gov (United States)

    Mondal, Swastik; Bykova, Elena; Dey, Somnath; Ali, Sk Imran; Dubrovinskaia, Natalia; Dubrovinsky, Leonid; Parakhonskiy, Gleb; van Smaalen, Sander

    2016-01-01

    A unique combination of useful properties in boron-carbide, such as extreme hardness, excellent fracture toughness, a low density, a high melting point, thermoelectricity, semi-conducting behavior, catalytic activity and a remarkably good chemical stability, makes it an ideal material for a wide range of technological applications. Explaining these properties in terms of chemical bonding has remained a major challenge in boron chemistry. Here we report the synthesis of fully ordered, stoichiometric boron-carbide B13C2 by high-pressure-high-temperature techniques. Our experimental electron-density study using high-resolution single-crystal synchrotron X-ray diffraction data conclusively demonstrates that disorder and defects are not intrinsic to boron carbide, contrary to what was hitherto supposed. A detailed analysis of the electron density distribution reveals charge transfer between structural units in B13C2 and a new type of electron-deficient bond with formally unpaired electrons on the C-B-C group in B13C2. Unprecedented bonding features contribute to the fundamental chemistry and materials science of boron compounds that is of great interest for understanding structure-property relationships and development of novel functional materials.

  17. Electrical Characterization of Irradiated Semiconducting Amorphous Hydrogenated Boron Carbide

    Science.gov (United States)

    Peterson, George Glenn

    Semiconducting amorphous partially dehydrogenated boron carbide has been explored as a neutron voltaic for operation in radiation harsh environments, such as on deep space satellites/probes. A neutron voltaic device could also be used as a solid state neutron radiation detector to provide immediate alerts for radiation workers/students, as opposed to the passive dosimetry badges utilized today. Understanding how the irradiation environment effects the electrical properties of semiconducting amorphous partially dehydrogenated boron carbide is important to predicting the stability of these devices in operation. p-n heterojunction diodes were formed from the synthesis of semiconducting amorphous partially dehydrogenated boron carbide on silicon substrates through the use of plasma enhanced chemical vapor deposition (PECVD). Many forms of structural and electrical measurements and analysis have been performed on the p-n heterojunction devices as a function of both He+ ion and neutron irradiation including: transmission electron microscopy (TEM), selected area electron diffraction (SAED), current versus voltage I(V), capacitance versus voltage C(V), conductance versus frequency G(f), and charge carrier lifetime (tau). In stark contrast to nearly all other electronic devices, the electrical performance of these p-n heterojunction diodes improved with irradiation. This is most likely the result of bond defect passivation and resolution of degraded icosahedral based carborane structures (icosahedral molecules missing a B, C, or H atom(s)).

  18. 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.)

  19. Status of control assembly materials in Indian water reactors

    International Nuclear Information System (INIS)

    Date, V.G.; Kulkarni, P.G.

    2000-01-01

    India's present operating water cooled power reactors comprise boiling water reactors of Tarapur Atomic Power Station (TAPS) and pressurized heavy water reactors (PHWRs) at Kota (RAPS), Kalpakkam (MAPS), Narora (NAPS) and Kakrapara (KAPS). Boiling water reactors of TAPS use boron carbide control blades for control of power as well as for shut down (scram). PHWRs use boron steel and cobalt absorber rods for power control and Cd sandwiched shut off rods (primary shut down system) and liquid poison rods (secondary shut down system) for shut down. In TAPS, Gadolinium rods (burnable poison rods) are also incorporated in fuel assembly for flux flattening. Boron carbide control blades and Gadolinium rods for TAPS, cobalt absorber rods and shut down assemblies for PHWRs are fabricated indigenously. Considerable development work was carried out for evolving material specifications, component and assembly drawings, and fabrication processes. Details of various control and shut off assemblies being fabricated currently are highlighted in the paper. (author)

  20. Control rod drives

    International Nuclear Information System (INIS)

    Futatsugi, Masao.

    1980-01-01

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

  1. 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.)

  2. 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.)

  3. Effects of heat treatment on the microstructure of amorphous boron carbide coating deposited on graphite substrates by chemical vapor deposition

    International Nuclear Information System (INIS)

    Li Siwei; Zeng Bin; Feng Zude; Liu Yongsheng; Yang Wenbin; Cheng Laifei; Zhang Litong

    2010-01-01

    A two-layer boron carbide coating is deposited on a graphite substrate by chemical vapor deposition from a CH 4 /BCl 3 /H 2 precursor mixture at a low temperature of 950 o C and a reduced pressure of 10 KPa. Coated substrates are annealed at 1600 o C, 1700 o C, 1800 o C, 1900 o C and 2000 o C in high purity argon for 2 h, respectively. Structural evolution of the coatings is explored by electron microscopy and spectroscopy. Results demonstrate that the as-deposited coating is composed of pyrolytic carbon and amorphous boron carbide. A composition gradient of B and C is induced in each deposition. After annealing, B 4 C crystallites precipitate out of the amorphous boron carbide and grow to several hundreds nanometers by receiving B and C from boron-doped pyrolytic carbon. Energy-dispersive spectroscopy proves that the crystallization is controlled by element diffusion activated by high temperature annealing, after that a larger concentration gradient of B and C is induced in the coating. Quantified Raman spectrum identifies a graphitization enhancement of pyrolytic carbon. Transmission electron microscopy exhibits an epitaxial growth of B 4 C at layer/layer interface of the annealed coatings. Mechanism concerning the structural evolution on the basis of the experimental results is proposed.

  4. Estimation of dose rate around the spent control rods of a BWR; Estimacion de la rapidez de dosis alrededor de las barras de control gastadas de un BWR

    Energy Technology Data Exchange (ETDEWEB)

    Cancino P, G.

    2016-10-01

    The energy can come from fossil renewable sources (solar (natural gas, oil), wind, hydro, tidal, geothermal, biomass, bio energy and nuclear. Nuclear power can be obtained by fission reactions and fusion (still under investigation) atomic nuclei. Fission, is a partition of a very heavy nucleus (Uranium 235, for example) into two lighter nuclei. Much of the world's electric power is generated from the energy released by fission processes. In a nuclear power reactor, light water as the BWR, there are many important elements that allow safe driving operation, one of them are the elements or control systems, the burnable poison or neutron absorber inherently allow control power reactor. The control rods, which consist mostly of stainless steel and absorbing elements (such as boron carbide, hafnium, cadmium, among others) of thermal neutrons is able to initiate, regulate or stop the reactor power. These, due to the use of depleted burned or absorbing material and therefore reach their lifespan, which can be 15 years or have other values depending on the manufacturer. Control rods worn should be removed, stored or confined in expressly places. Precisely at this stage arises the importance of knowing their radiological condition to manipulate safely and without incident to the people health responsible for conducting these proceedings state arises. This thesis consists in the estimation of the dose rate in spent control rod made of boron carbide, from a typical BWR reactor. It will be estimated by direct radiation measurements with measurement equipment for radiotherapy ionization chamber, in six spent control rods, which were taken at different reactor operating cycles and are in a spent fuel pool. Using bracket electromechanical and electronic equipment for positioning and lifting equipment for radiation measurement around the control rod in the axial and radial arrangement for proper scanning. Finally will be presented a graphic corresponding to the dose

  5. Evidence for multiple polytypes of semiconducting boron carbide (C2B10) from electronic structure

    International Nuclear Information System (INIS)

    Lunca-Popa, Petru; Brand, J I; Balaz, Snjezana; Rosa, Luis G; Boag, N M; Bai Mengjun; Robertson, B W; Dowben, P A

    2005-01-01

    Boron carbides fabricated via plasma enhanced chemical vapour deposition from different isomeric source compounds with the same C 2 B 10 H 12 closo-icosahedral structure result in materials with very different direct (optical) band gaps. This provides compelling evidence for the existence of multiple polytypes of C 2 B 10 boron carbide and is consistent with electron diffraction results

  6. Characterization of plastic and boron carbide additive manufactured neutron collimators

    Science.gov (United States)

    Stone, M. B.; Siddel, D. H.; Elliott, A. M.; Anderson, D.; Abernathy, D. L.

    2017-12-01

    Additive manufacturing techniques allow for the production of materials with complicated geometries with reduced costs and production time over traditional methods. We have applied this technique to the production of neutron collimators for use in thermal and cold neutron scattering instrumentation directly out of boron carbide. We discuss the design and generation of these collimators. We also provide measurements at neutron scattering beamlines which serve to characterize the performance of these collimators. Additive manufacturing of parts using neutron absorbing material may also find applications in radiography and neutron moderation.

  7. Kerma factors in interaction of neutrons with boron carbide

    International Nuclear Information System (INIS)

    Bondarenko, I.M.

    1979-01-01

    Heat generation in neutron interactions with boron carbide B 10 ; B 11 and 12 C is calculated. Kerma-factors (kerma-kinetic energy released in materials) were calculated for neutron energies between 10 -4 eV and 15 MeV. No major simplifying assumptions are introduced, and the accuracy of the calculated kerma-factors depends only on availability and accuracy of the basic nuclear data. The ENDF/B-4 data and recent experimental information are used for the calculation of kerma-factors. Plots of these kerma-factors are presented in units of eVxb/atom and wtxsec/(cmxn) as a function of neutron energy

  8. Boron carbide synthesis by carbothermic reduction of boron oxide

    International Nuclear Information System (INIS)

    Castro, A.R.M. de; Paschoal, J.O.A.

    1988-01-01

    Boron carbide (B 4 C) is a ceramic material of technological applications due to its extreme hardness and high chemical as well as thermal stability. Some parameters of the process for obtaining B 4 C by carbothermic reduction of B 2 O 3 have been determined. The starting powders and the final products have been analysed by chemical, spectrographic and X-ray diffraction methods. The results show that the B 4 C obtained by the carbothermic reduction process is suitable for applications with a definite determination of the free carbon content. (author) [pt

  9. 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

  10. Control rod drive

    International Nuclear Information System (INIS)

    Okutani, Tetsuro.

    1988-01-01

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

  11. 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

  12. 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

  13. Control rod drives

    International Nuclear Information System (INIS)

    Hayakawa, Hiroyasu; Kawamura, Atsuo.

    1979-01-01

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

  14. Control rod driving mechanism

    International Nuclear Information System (INIS)

    Ooshima, Yoshio.

    1983-01-01

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

  15. Control rod position control device

    International Nuclear Information System (INIS)

    Ubukata, Shinji.

    1997-01-01

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

  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. Control rod drives

    International Nuclear Information System (INIS)

    Asano, Hiromitsu.

    1979-01-01

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

  18. Size-scaling of tensile failure stress in boron carbide

    Energy Technology Data Exchange (ETDEWEB)

    Wereszczak, Andrew A [ORNL; Kirkland, Timothy Philip [ORNL; Strong, Kevin T [ORNL; Jadaan, Osama M. [University of Wisconsin, Platteville; Thompson, G. A. [U.S. Army Dental and Trauma Research Detachment, Greak Lakes

    2010-01-01

    Weibull strength-size-scaling in a rotary-ground, hot-pressed boron carbide is described when strength test coupons sampled effective areas from the very small (~ 0.001 square millimeters) to the very large (~ 40,000 square millimeters). Equibiaxial flexure and Hertzian testing were used for the strength testing. Characteristic strengths for several different specimen geometries are analyzed as a function of effective area. Characteristic strength was found to substantially increase with decreased effective area, and exhibited a bilinear relationship. Machining damage limited strength as measured with equibiaxial flexure testing for effective areas greater than ~ 1 mm2 and microstructural-scale flaws limited strength for effective areas less than 0.1 mm2 for the Hertzian testing. The selections of a ceramic strength to account for ballistically-induced tile deflection and to account for expanding cavity modeling are considered in context with the measured strength-size-scaling.

  19. Structure and properties of hot-pressed boron carbide ceramics

    Energy Technology Data Exchange (ETDEWEB)

    Koval' chenko, M S; Tkachenko, IU G; Koval' chuk, V V; Iurchenko, D Z; Satanin, S V [Institut Problem Materialovedeniia, Kiev (Ukrainian SSR)

    1990-07-01

    The microstructure and strength of B4C-TiB2-TiO{sub 2} ceramics samples, hot-compacted from a mixture of two types of B4C-TiO2-C powder, are examined. The two types are obtained by combining boric acid with either sucrose or carbon black. The grain-sizes of the two powders are found to be distinctly different from one another both before and after the grinding procedure and the degree of dispersion is not high. The strength tests show 600 MPa, the Vicker's hardness is 34.5 GPa, and the crack resistance coefficient of ceramics containing 15 percent TiB2 by mass is 5 MPa m exp 1/2. The use of soluble boron carbide powder helps achieve higher levels of strength and crack resistance. 5 refs.

  20. Frequency mixing in boron carbide laser ablation plasmas

    Science.gov (United States)

    Oujja, M.; Benítez-Cañete, A.; Sanz, M.; Lopez-Quintas, I.; Martín, M.; de Nalda, R.; Castillejo, M.

    2015-05-01

    Nonlinear frequency mixing induced by a bichromatic field (1064 nm + 532 nm obtained from a Q-switched Nd:YAG laser) in a boron carbide (B4C) plasma generated through laser ablation under vacuum is explored. A UV beam at the frequency of the fourth harmonic of the fundamental frequency (266 nm) was generated. The dependence of the efficiency of the process as function of the intensities of the driving lasers differs from the expected behavior for four-wave mixing, and point toward a six-wave mixing process. The frequency mixing process was strongly favored for parallel polarizations of the two driving beams. Through spatiotemporal mapping, the conditions for maximum efficiency were found for a significant delay from the ablation event (200 ns), when the medium is expected to be a low-ionized plasma. No late components of the harmonic signal were detected, indicating a largely atomized medium.

  1. Kerma factors in interaction of neutrons with boron carbide

    International Nuclear Information System (INIS)

    Bondarenko, I.M.

    1986-03-01

    Heat generation in neutron interactions with boron carbide B 10 ; B 11 and 12 C is calculated. Kerma-factors (kerma-kinetic energy released in materials) were calculated for neutron energies between 10 -4 eV and 15 MeV. No major simplifying assumptions are introduced, and the accuracy of the calculated kerma-factors depends only on availability and accuracy of the basic nuclear data. The ENDF/B-4 data and recent experimental information are used for the calculation of kerma-factors. Plots of these kerma-factors are presented in units of eVxb/atom and wtxsec/(cmxn) as a function of neutron energy [fr

  2. Structural phase transitions in boron carbide under stress

    International Nuclear Information System (INIS)

    Korotaev, P; Pokatashkin, P; Yanilkin, A

    2016-01-01

    Structural transitions in boron carbide B 4 C under stress were studied by means of first-principles molecular dynamics in the framework of density functional theory. The behavior depends strongly on degree of non-hydrostatic stress. Under hydrostatic stress continuous bending of the three-atom C–B–C chain was observed up to 70 GPa. The presence of non-hydrostatic stress activates abrupt reversible chain bending, which is displacement of the central boron atom in the chain with the formation of weak bonds between this atom and atoms in the nearby icosahedra. Such structural change can describe a possible reversible phase transition in dynamical loading experiments. High non-hydrostatic stress achieved in uniaxial loading leads to disordering of the initial structure. The formation of carbon chains is observed as one possible transition route. (paper)

  3. Control rod drive mechanism

    International Nuclear Information System (INIS)

    Mizuno, Katsuyuki.

    1976-01-01

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

  4. Investigations on the conditions for obtaining high density boron carbide by sintering

    International Nuclear Information System (INIS)

    Kislyj, P.S.; Grabtschuk, B.L.

    1975-01-01

    The results of investigations on kinetics of condensation and mechanisms of mass transfer in the process of sintering of technical, chemically pure and synthesized boron carbide are generalized. Laws on boron carbide densification depending upon temperature, time of isothermic endurance, thermal speed, size of powder particles and variable composition in homogeneity are determined. From the results obtained on condensation kinetics and special experiments on studying the changes in properties after heating under different conditions, the role of dislocation and diffusion processes in mass transfer during boron carbide sintering is exposed. The properties of sintered boron carbide are 15-20% lower than the properties of high-pressed one, that is conditioned by intercrystallite distortion of the first one and transcrystallite of the second one

  5. 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

  6. Pulverization of boron element and proportions of boron carbide in boron

    International Nuclear Information System (INIS)

    Lang, F.M.; Finck, C.

    1956-01-01

    It is possible to reduce boron element into fine powder by means of a mortar and pestle made of sintered boron carbide, the ratio of boron carbide introduced being less than one per cent. Boron element at our disposal is made of sharp edged, dark brown, little grains of average size greater than 5 μ. Grain sizes smaller than 1μ are required for applying thin layers of such boron. (author) [fr

  7. Control rod drives

    International Nuclear Information System (INIS)

    Ikakura, Hiroaki.

    1986-01-01

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

  8. 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.)

  9. Control rod drive mechanism

    International Nuclear Information System (INIS)

    Futatsugi, Masao; Goto, Mikihiko.

    1976-01-01

    Purpose: To provide a control rod drive mechanism using water as an operating source, which prevents a phenomenon for forming two-layers of water in the neighbourhood of a return nozzle in a reactor to limit formation of excessive thermal stress to improve a safety. Constitution: In the control rod drive mechanism of the present invention, a heating device is installed in the neighbourhood of a pressure container for a reactor. This heating device is provided to heat return water in the reactor to a level equal to the temperature of reactor water thereby preventing a phenomenon for forming two-layers of water in the reactor. This limits formation of thermal stress in the return nozzle in the reactor. Accordingly, it is possible to minimize damages in the return nozzle portion and yet a possibility of failure in reactor water. (Kawakami, Y.)

  10. Control rod drives

    International Nuclear Information System (INIS)

    Yamanaka, Toshikatsu.

    1979-01-01

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

  11. 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)

  12. Control rod drive shaft latch

    International Nuclear Information System (INIS)

    Thorp, A.G. II.

    1976-01-01

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

  13. 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.)

  14. Amorphisation of boron carbide under slow heavy ion irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Gosset, D., E-mail: Dominique.gosset@cea.fr [CEA Saclay, DEN, DANS, DMN, SRMA, LA2M, Université Paris-Saclay, 91191, Gif/Yvette (France); Miro, S. [CEA Saclay, DEN, DANS, DMN, SRMP, Laboratoire JANNUS, Université Paris-Saclay, 91191, Gif/Yvette (France); Doriot, S. [CEA Saclay, DEN, DANS, DMN, SRMA, LA2M, Université Paris-Saclay, 91191, Gif/Yvette (France); Moncoffre, N. [CNRS/IN2P3/IPNL, 69622, Villeurbanne (France)

    2016-08-01

    Boron carbide B{sub 4}C is widely used as a neutron absorber in nuclear plants. Most of the post-irradiation examinations have shown that the structure of the material remains crystalline, in spite of very high atomic displacement rates. Here, we have irradiated B{sub 4}C samples with 4 MeV Au ions with different fluences at room temperature. Transmission electron microscopy (TEM) and Raman spectroscopy have been performed. The Raman analyses show a high structural disorder at low fluence, around 10{sup −2} displacements per atoms (dpa). However, the TEM observations show that the material remains crystalline up to a few dpa. At high fluence, small amorphous areas a few nanometers large appear in the damaged zone but the long range order is preserved. Moreover, the size and density of the amorphous zones do not significantly grow when the damage increases. On the other hand, full amorphisation is observed in the implanted zone at a Au concentration of about 0.0005. It can be inferred from those results that short range and long range damages arise at highly different fluences, that heavy ions implantation has drastic effects on the structure stability and that in this material self-healing mechanisms are active in the damaged zone.

  15. Boron carbide-carbon composites and composites for cryogenic applications

    International Nuclear Information System (INIS)

    Sheinberg, H.

    1979-01-01

    Because of its neutronic properties, high hardness, and high melting temperature, boron carbide (B 4 C) is widely used at the Los Alamos Scientific Laboratory. However because of its hardness and mode of manufacture, it is expensive to machine finish to tight dimensional specifictions. For some neutronic applications, a density considerably below the theoretical 2.52 Mg/m 3 was acceptable, and this relaxation in density specification permitted addition of carbon as a second phase to reduce machining costs. We conducted an experimental program to prepare 50.8-mm-diam by 34.8-mm-thick cylinders of B 4 C and B 4 C-C composites with concentrations of carbon varying from 5.5 to 30 volume percent. Additionally we used three forms of carbon, natural flake graphite, synthetic graphite flour, and a fine furnace black as the source of the second phase. We determined the sound velocity, compressive strength, coefficient of thermal expansion, electrical resistivity, and microstructure as functions of composition. Additionally, an enriched boron ( 10 B)-carbon composite was studied as an alternate material

  16. Study of rare earth elements as material for control rods

    International Nuclear Information System (INIS)

    1975-03-01

    The properties of rare earth elements as the material for control rods were studied. The rare earth elements, especially europium oxide, has the nuclear property corresponding to boron carbide, and its neutron absorption process does not emit alpha particles. The elements produced as a result of neutron capture also have large capture cross sections. This paper presents survey report on the properties and nuclear properties of rare earth elements, and comparison with other materials. Preliminary experiment was performed to make the pellets of europium oxide, and is described in this paper. Because of large density, the crystal form to be made was monoclinic system. Europium hydroxide was decomposed at 1000 0 C and 10 -5 torr. The obtained powder was dipped into benzene, and dryed in the air at 450 0 C. This powder was pressed and sintered in the air for one hour at 1500 0 C. The density of the obtained pellets was 97.0% of the theoretical density. The cross section of europium for fast neutron absorption is not yet accurately obtained, and is in the range between 4.65 and 8.5 barn for 151 Eu(n,γ) reaction. Since chain absorption reaction is caused in Eu, the overall capability of neutron absorption is not much changed by the loss of original material due to absorption. The pellets of europium oxide may be handled in air, but must be kept in dry atmosphere. The reactions of europium oxide with various metals were also investigated. The characteristic behavior in case of irradiation depends on the amount of silicon contained, and it was very good if the amount was less than 0.03%. (Kato, T.)

  17. Boron carbide nanostructures: A prospective material as an additive in concrete

    Science.gov (United States)

    Singh, Paviter; Kaur, Gurpreet; Kumar, Rohit; Kumar, Umesh; Singh, Kulwinder; Kumar, Manjeet; Bala, Rajni; Meena, Ramovatar; Kumar, Akshay

    2018-05-01

    In recent decades, manufacture and ingestion of concrete have increased particularly in developing countries. Due to its low cost, safety and strength, concrete have become an economical choice for protection of radiation shielding material in nuclear reactors. As boron carbide has been known as a neutron absorber material makes it a great candidate as an additive in concrete for shielding radiation. This paper presents the synthesis of boron carbide nanostructures by using ball milling method. The X-ray diffraction pattern, Fourier Transform Infrared Spectroscopy (FTIR) and Scanning Electron Microscope analysis confirms the formation of boron carbide nanostructures. The effect of boron carbide nanostructures on the strength of concrete samples was demonstrated. The compressive strength tests of concrete cube B4C powder additives for 0 % and 5 % of total weight of cement was compared for different curing time period such as 7, 14, 21 and 28 days. The high compressive strength was observed when 5 wt % boron carbide nanostructures were used as an additive in concrete samples after 28 days curing time and showed significant improvement in strength.

  18. Method of fabricating a poision tube for reactor control rods

    Energy Technology Data Exchange (ETDEWEB)

    Mori, Yasuhiko; Yoshida, Toshimi; Masaoka, Isao; Naruse, Akisuke

    1983-04-28

    A method to unify the neutron absorbing performance, enhance the workability in the insertion of neutron absorber tube and further decrease the stresses acting on the neutron absorber coating tube is described. The neutron absorber coated rod comprising neutron absorbing substance and a metal pipe is fabricated by compressing a metal pipe filled with the neutron absorber. Specifically, neutron absorbing substance such as boron carbide powder or the like is filled in a metal pipe such as made of stainless steel tube by way of vibration packing or the like. Then, after heating the metal pipe, it is applied with compression working such as swaging into a fine tube to increase the packing density of the absorbing substance filled in the pipe to greater than 60% of the theoretical density and completely contacted closely to the inner wall of the pipe. The neutron absorber coated rod thus fabricated can be inserted to an external coating tube with ease at a predetermined gap.

  19. Effects of B4C control rod degradation under severe accident

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Si-Won; Park, Sang-Gil; Han, Sang-Ku [Atomic Creative Technology Co., Daejeon (Korea, Republic of)

    2016-10-15

    Boron carbide (B4C) is widely used as absorber material in western boiling water reactor (BWR), some PWR, EPR and Russian RBMK and VVERs. B4C oxidation is one of the important phenomena of in-vessel. In the present paper, the main results and knowledge gained regarding the B4C control rod degradation from above mentioned experiments are reviewed and arranged to inform its significance on the severe accident consequences. In this paper, the role of B4C control rod oxidation and the subsequent degradation on the severe accident consequences is reviewed with available literature and report of previous experimental program regarding the B4C oxidation. From this review, it seems that the contribution of this B4C oxidation on the accident progression to the further severe accident situation is not negligible. For the future work, the extensive experimental data interpretation will be performed to assess quantitatively the effect of B4C oxidation and degradation on the various postulated severe accident conditions.

  20. 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.)

  1. Study and optimization of the carbothermic reduction process for obtaining boron carbide

    International Nuclear Information System (INIS)

    Castro, A.R.M. de.

    1989-01-01

    Boron carbide - B sub(4)C - is a ceramic material of technological importance due to its hardness and high chemical and thermal stabilities. Moreover, its high neutron capture cross section makes it suitable for application as neutron absorber in nuclear technology. The process for obtaining carbothermally derived boron carbide has been studied in two steps: firstly, the parameters of the boric acid → boron oxide dehydration reaction have been defined; secondly, the optimization of the carbothermal reduction reaction using boron oxide has been undertaken looking for boron carbide having low level of free carbon. The starting materials as well as the main products have been studied by chemical and spectrographic analyses, X-ray diffractometry, granulometric classification and scanning electron microscopy. The optimization of the carbothermic reduction process allowed for the development and set up of a fabrication procedure yielding high quality B sub(4) C powders, starting from low cost and easily available (in the Brazilian market) raw materials. (author)

  2. The determination of boron and carbon in reactor grade boron carbide

    International Nuclear Information System (INIS)

    Crossley, D.; Wood, A.J.; McInnes, C.A.J.; Jones, I.G.

    1978-09-01

    The sealed tube method of dissolution at high temperature and pressure has been successfully applied in the analysis of reactor grade boron carbide for the determination of boron. A 50 mg sample of boron carbide is completely dissolved by heating with concentrated nitric acid in a sealed tube at 300 0 C. The boron content of the resultant sample solution is determined by the mannitol potentiometric titration method. The precision of the method for the determination of 2.5 mg of boron using the Harwell automatic potentiometric titrator is 0.2% (coefficient of variation). The carbon content of a boron carbide sample is determined by combustion of the sample at 1050 0 C in a stream of oxygen using vanadium pentoxide to ensure the complete oxidation of the sample. The carbon dioxide produced from the sample is measured manometrically and the precision of the method for the determination of 4 mg of carbon is 0.4% (coefficient of variation). (author)

  3. Evaluation of mechanical properties of aluminium alloy–alumina–boron carbide metal matrix composites

    International Nuclear Information System (INIS)

    Vijaya Ramnath, B.; Elanchezhian, C.; Jaivignesh, M.; Rajesh, S.; Parswajinan, C.; Siddique Ahmed Ghias, A.

    2014-01-01

    Highlights: • Fabrication of MMC with aluminium alloy–alumina–boron carbide is done. • Different proportions of reinforcements are added. • The effects of varying proportions are studied. • Investigation on mechanical properties above composites is performed. • Failure morphology analysis is done using SEM. - Abstract: This paper deals with the fabrication and mechanical investigation of aluminium alloy, alumina (Al 2 O 3 ) and boron carbide metal matrix composites. Aluminium is the matrix metal having properties like light weight, high strength and ease of machinability. Alumina which has better wear resistance, high strength, hardness and boron carbide which has excellent hardness and fracture toughness are added as reinforcements. Here, the fabrication is done by stir casting which involves mixing the required quantities of additives into stirred molten aluminium. After solidification, the samples are prepared and tested to find the various mechanical properties like tensile, flexural, impact and hardness. The internal structure of the composite is observed using Scanning Electron Microscope (SEM)

  4. Control-rod driving mechanism

    International Nuclear Information System (INIS)

    Jodoi, Takashi.

    1976-01-01

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

  5. Control rod drives

    International Nuclear Information System (INIS)

    Furumitsu, Yutaka.

    1981-01-01

    Purpose: To improve the reliability of a device for driving an LMFBR type reactor control rod by providing a buffer unit having a stationary electromagnetic coil and a movable electromagnetic coil in the device to thereby avord impact stress at scram time and to simplify the structure of the buffer unit. Constitution: A non-contact type buffer unit is constructed with a stationary electromagnetic coil, a cable for the stationary coil, a movable electromagnetic coil, a spring cable for the movable coil, and a backup coil spring or the like. Force produced at scram time is delivered without impact by the attracting or repelling force between the stationary coil and the movable coil of the buffer unit. Accordingly, since the buffer unit is of a non-contact type, there is no mechanical impact and thus no large impact stress, and as it has simple configuration, the reliability is improved and the maintenance can be conducted more easily. (Yoshihara, H.)

  6. Irradiation damage in boron carbide: point defects, clusters and helium bubbles

    International Nuclear Information System (INIS)

    Stoto, T.; Zuppiroli, L.

    1986-06-01

    Boron carbide is a refractory hard and light material of interest in nuclear technology (fission and also fusion). Transmission electron microscopy was used to examine the properties of radiation induced damage. Firstly, the production of point defects and their clustering was studied in samples irradiated by 1 MeV electron in a high voltage electron microscope at selected temperatures from 12 K to 1000 K. Secondly, conventional transmission electron microscopy was used to understand the production of helium bubbles in neutron irradiated boron carbide and their role in the generation of microcracks. Finally, the interaction between point defects and bubbles was also examined

  7. 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.)

  8. Reconstitutable control rod spider assembly

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  9. 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

  10. Control rod drive

    International Nuclear Information System (INIS)

    Kojima, Akira.

    1989-01-01

    In the control rod drive for a BWR type reactor, etc., according to this invention, the lower limit flow rate is set so as to keep the restriction for stability upon spectral shift operation. The setting condition for keeping the restriction is the lowest pump speed and the lower limit for the automatic control of the flow rate, which are considered to be important in view of the stablility from the actual power state. In view of the above, it is possible to keep the reactor core stably even in a case where such a transient phenomenon occurs that the recycling flow rate has to be run back to the lowest pump speed during spectral shift opeeration or in a case where the load demand is reduced and the flow rate is decreased by an automatic mode as in night operation. Accordingly, in the case of conducting the spectral shift operation according to this invention, the operation region capable of keeping the reactor core state stably during operation can be extended. (I.S.)

  11. 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

  12. 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.)

  13. Evidence of amorphisation of B{sub 4}C boron carbide under slow, heavy ion irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Gosset, D., E-mail: dominique.gosset@cea.fr [CEA, DEN, DMN-SRMA-LA2M, F-91191 Gif/Yvette (France); Miro, S. [CEA, DEN, DMN-SRMP-JANNUS, F-91191 Gif/Yvette (France); Doriot, S. [CEA, DEN, DMN-SRMA-LA2M, F-91191 Gif/Yvette (France); Victor, G. [CNRS-IN2P3-IPNL, F-69622 Villeurbanne (France); Motte, V. [CEA, DEN, DMN-SRMA-LA2M, F-91191 Gif/Yvette (France)

    2015-12-15

    Boron carbide is widely used either as armor-plate or neutron absorber. In both cases, a good structural stability is required. However, a few studies have shown amorphisation may occur in severe conditions. Hard impacts lead to the formation of amorphous bands. Some irradiations in electronic regime with H or He ions have also shown amorphisation of the material. Most authors however consider the structure is not drastically affected by irradiations in the ballistic regime. Here, we have irradiated at room temperature dense boron carbide pellets with Au 4 MeV ions, for which most of the damage is in the ballistic regime. This study is part of a program devoted to the behavior of boron carbide under irradiation. Raman observations have been performed after the irradiations together with transmission electron microscopy (TEM). Raman observations show a strong structural damage at moderate fluences (10{sup 14}/cm{sup 2}, about 0.1 dpa), in agreement with previous studies. On the other hand, TEM shows the structure remains crystalline up to 10{sup 15}/cm{sup 2} then partially amorphises. The amorphisation is heterogeneous, with the formation of nanometric amorphous zones with increasing density. It then appears short range and long range disorder occurs at quite different damage levels. Further experiments are in progress aiming at studying the structural stability of boron carbide and isostructural materials (α-B, B{sub 6}Si,…).

  14. 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

  15. 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.)

  16. Study of boron carbide evolution under neutron irradiation; Contribution a l'etude de l'evolution du carbure de bore sous irradiation neutronique

    Energy Technology Data Exchange (ETDEWEB)

    Simeone, D. [CEA/Saclay, Dept. de Mecanique et de Technologie (DMT), 91 - Gif-sur-Yvette (France)]|[Universite Blaise Pascal, Clermont-Ferrand II, (CNRS), 63 - Aubiere (France)

    1999-07-01

    Owing to its high neutron efficiency, boron carbide (B{sub 4}C) is used as a neutron absorber in control rods of nuclear plants. Its behaviour under irradiation has been extensively studied for many years. It now seems clear that brittleness of the material induced by the {sup 10}B(n,{alpha}){sup 7}Li capture reaction is due to penny shaped helium bubbles associated to a high strain field around them. However, no model explains the behaviour of the material under neutron irradiation. In order to build such a model, this work uses different techniques: nuclear microprobe X-ray diffraction profile analysis and Raman and Nuclear Magnetic Resonance Spectroscopy to present an evolution model of B{sub 4}C under neutron irradiation. The use of nuclear reactions produced by a nuclear microprobe such as the {sup 7}Li(p,p'{gamma}){sup 7}Li reaction, allows to measure lithium profile in B{sub 4}C pellets irradiated either in Pressurised Water Reactors or in Fast Breeder Reactors. Examining such profiles enables us to describe the migration of lithium atoms out of B{sub 4}C materials under neutron irradiation. The analysis of X-ray diffraction profiles of irradiated B{sub 4}C samples allows us to quantify the concentrations of helium bubbles as well as the strain fields around such bubbles.Furthermore Raman spectroscopy studies of different B{sub 4}C samples lead us to propose that under neutron irradiation. the CBC linear chain disappears. Such a vanishing of this CBC chain. validated by NMR analysis, may explain the penny shaped of helium bubbles inside irradiated B{sub 4}C. (author)

  17. Equations of state and melting curve of boron carbide in the high-pressure range of shock compression

    Energy Technology Data Exchange (ETDEWEB)

    Molodets, A. M., E-mail: molodets@icp.ac.ru; Golyshev, A. A.; Shakhrai, D. V. [Russian Academy of Sciences, Institute for Problems in Chemical Physics (Russian Federation)

    2017-03-15

    We have constructed the equations of state for crystalline boron carbide B{sub 11}C (C–B–C) and its melt under high dynamic and static pressures. A kink on the shock adiabat for boron carbide has been revealed in the pressure range near 100 GPa, and the melting curve with negative curvature in the pressure range 0–120 GPa has been calculated. The results have been used for interpreting the kinks on the shock adiabat for boron carbide in the pressure range of 0–400 GPa.

  18. Control rod calibration including the rod coupling effect

    International Nuclear Information System (INIS)

    Szilard, R.; Nelson, G.W.

    1984-01-01

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

  19. Approach to criticality in cold core of the Unit 1 of the cycle 10 of the Laguna Verde power plant with control rods similar in design to the original ones and the advanced ones

    International Nuclear Information System (INIS)

    Perusquia, R.; Ramirez, J.R.; Ortiz, J.J.; Hernandez, J.L.; Montes, J.L.

    2004-01-01

    In the Laguna Verde power plant (CLV) worn out control bars have been replacing for control rods of advanced design. At the moment of the 109 rods of similar design to the original of the Unit 1 of the CLV, 14 of them have been substituted with two types of different advanced control rods. The same as the original rods the advanced rods use boron carbide as control element for neutrons, but additionally they use hafnium, to achieve a longer useful half life. The rods of similar design to the original are homogeneous axially on the other hand the advanced presenting several zones. In principle the advanced rods were designed to be neutronically compatible to the original rods. However the differences in the design and of constituents assumed to consider a different performance during the operation. Since the one numbers of advanced bars is growing with the number of cycles of the CLV reactor has requested to the National Institute of Nuclear Research (ININ) to evaluate in detailed form the performance of the control rods. Previously the calculations and analysis of cycle design that that were carried out assumed that control rods should be of similar design as the original. Using the HELIOS/CM-LENDS system, the effect of the inclusion of advanced rods was analyzed in the approximation of the criticality in cold (20 C) in the cycle 10 of the Unit 1. Four scenarios were poposed, where the type of the rod of substitution control is varied. The value of reactivity was calculated for the rods similar in design to the original and for the advanced one. By comparison of these values, it was found that the value of reactivity of the control rods of the advanced design was superior to the value for rods similar to the original ones by 7.6% for one of the advanced designs and of 13.6% for the other one. (Author)

  20. 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.)

  1. Development of high temperature metallic melting processes related to detritiation of exhausted control rods

    International Nuclear Information System (INIS)

    Dworschak, H.; Mannone, F.; Modica, G.

    1994-01-01

    A rather critical problem to be faced in developing a safe strategy for the management of tritiated solid wastes is dealing with the outgassing property of tritium. Releases of tritium under elemental or oxide form may occur from waste items at different temperatures and rates depending upon the nature of tritium bonds into the waste matrix as well as on its 'contamination history'. Apart from the commercial value of tritium, its release from waste packages anyhow represents a risk of tritium exposure that cannot be accepted by skippers, by store and disposal site operators as well as by the general public. Consequently it is mandatory to carry out the detritiation of such wastes before their packaging and storage or disposal. In the boron carbide control rods from the Lingen BWR after about three years of operation, tritium generated by neutron reaction was essentially retained in the B 4 C matrix. The objectives of the study are to demonstrate the feasibility of two processes aimed at reducing to the maximum practicable extent the level of tritium contamination in such waste management are facilitated

  2. Control rod cluster with removable rods for nuclear fuel assembly

    International Nuclear Information System (INIS)

    Denizou, J.P.

    1989-01-01

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

  3. Method of accurate thickness measurement of boron carbide coating on copper foil

    Science.gov (United States)

    Lacy, Jeffrey L.; Regmi, Murari

    2017-11-07

    A method is disclosed of measuring the thickness of a thin coating on a substrate comprising dissolving the coating and substrate in a reagent and using the post-dissolution concentration of the coating in the reagent to calculate an effective thickness of the coating. The preferred method includes measuring non-conducting films on flexible and rough substrates, but other kinds of thin films can be measure by matching a reliable film-substrate dissolution technique. One preferred method includes determining the thickness of Boron Carbide films deposited on copper foil. The preferred method uses a standard technique known as inductively coupled plasma optical emission spectroscopy (ICPOES) to measure boron concentration in a liquid sample prepared by dissolving boron carbide films and the Copper substrates, preferably using a chemical etch known as ceric ammonium nitrate (CAN). Measured boron concentration values can then be calculated.

  4. Role of intericosahedral chains on the hardness of sputtered boron carbide films

    International Nuclear Information System (INIS)

    Jacobsohn, L.G.; Averitt, R.D.; Wetteland, C.J.; Schulze, R.K.; Nastasi, M.; Daemen, L.L.; Jenei, Z.; Asoka-Kumar, P.

    2004-01-01

    The relationship between the structure and mechanical properties of sputter-deposited boron carbide films was investigated. Changes in the structure induced by annealing were characterized in terms of chemical composition, chemical bonding, and concentrations of defects and trapped impurities. The creation of intericosahedral chains for higher annealing temperatures was revealed by infrared and Raman measurements, and the intensity of the infrared band at 1500 cm-1 was found to be related to the hardness. The presence of residual trapped Ar atoms and of open-volume defects is insensitive to relatively high annealing temperatures and does not influence the recovery of the hardness. Our results suggest postdeposition annealing as a pathway to enhance the mechanical properties of boron carbide films

  5. 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.)

  6. Advanced gray rod control assembly

    Science.gov (United States)

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

    2013-09-17

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

  7. The BWR Hybrid 4 control rod

    International Nuclear Information System (INIS)

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

    1988-01-01

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

  8. Optimizing Grain Boundary Complexions to Produce Dense Pressure-Less Sintered Boron Carbide (B4C)

    Science.gov (United States)

    2008-11-14

    discontinuous distribution of the yttria. At this stage it is difficult to determine if the discontinuity is genuine or results from dewetting upon cooling...sample. However, the tendency of the film to form beads indicates a dewetting behavior. The weak interface between the yttria and the boron carbide...conform to the dewetting behavior. There is a possibility of a complexion transition as the sample is cooled down in the furnace. At high temperature the

  9. Advanced Boron Carbide-Based Visual Obscurants for Military Smoke Grenades

    Science.gov (United States)

    2014-07-13

    components on grenade burning time must be accounted for. Also, how these factors affect smoke cloud thickness and perceived visible obscuration...of Boron Carbide-Based Smoke Compositions. Propellants, Explos., Pyrotech. 2013, 38, 622-628. (30) Commission Internationale de l’Éclairage ( CIE ...volume, gives equation 2. Since αm is only a characteristic of the aerosol, it does not account for

  10. Compatibility of heat resistant alloys with boron carbide, (4)

    International Nuclear Information System (INIS)

    Baba, Sinichi; Saruta, Toru; Ooka, Kiichi; Tanaka, Isao; Aoyama, Isao

    1985-07-01

    This paper relates to the compatibility test of control rod sheath (Hastelloy XR alloy) and neutron absorber (boronated graphite) for the VHTR, which has been researched and developed by JAERI. The irradiation was conducted by using the OGL-1 irradiation facility in the JMTR in order to study reaction behaviour between Hastelloy XR alloy and boronated graphite as well as to determine a reaction barrier performance of refractory metal foils Nb, Mo, W and Re. Irradiation conditions were as follows. Neutron dose : 4.05 x 10 22 m -2 (E 18 m -2 (E > 0.16 pJ, 1 Mev). Helium coolant : Average temperature 855 0 C, Pressure 2.94 MPa, Total impurity concentration 400 kBq/m 3 . Irradiation time : 5.0 Ms (1390 hours). Post-irradiation examinations i.e. visual inspection, dimensional inspection, weight measurement, metallography, hardness test, morphological observations by SEM and analysis of element distributions by EPMA were carried out. In the result, reaction products of Hastelloy XR alloy were observed in the ellipsoidal form locally. These results were same as those of the out-of-pile tests. Obvious irradiation effects were not detectable but a little accelarated increase in reaction depth of Hastelloy XR alloy by heat effect of specimens was observed. The refractory metal foils had a good performance of reaction barrier between Hastelloy XR alloy and boronated graphite. Furthermore, movement of Ni, Fe and Cr in the reaction area of Hastelloy XR alloy, difference in the reaction depth of B and C, irradiation effects on diffusion coefficient, lithium production and heat effect are discussed. (author)

  11. Structural modifications induced by ion irradiation and temperature in boron carbide B{sub 4}C

    Energy Technology Data Exchange (ETDEWEB)

    Victor, G., E-mail: g.victor@ipnl.in2p3.fr [Institut de Physique Nucléaire de Lyon (IPNL), Université Lyon 1, CNRS/IN2P3, 4 rue Enrico Fermi, 69622 Villeurbanne Cedex (France); Pipon, Y.; Bérerd, N. [Institut de Physique Nucléaire de Lyon (IPNL), Université Lyon 1, CNRS/IN2P3, 4 rue Enrico Fermi, 69622 Villeurbanne Cedex (France); Institut Universitaire de Technologie (IUT) Lyon-1, Université Claude Bernard Lyon 1, 69622 Villeurbanne Cedex (France); Toulhoat, N. [Institut de Physique Nucléaire de Lyon (IPNL), Université Lyon 1, CNRS/IN2P3, 4 rue Enrico Fermi, 69622 Villeurbanne Cedex (France); CEA-DEN, Saclay, 91191 Gif-sur-Yvette (France); Moncoffre, N. [Institut de Physique Nucléaire de Lyon (IPNL), Université Lyon 1, CNRS/IN2P3, 4 rue Enrico Fermi, 69622 Villeurbanne Cedex (France); Djourelov, N. [Institute for Nuclear Research and Nuclear Energy, Bulgarian Academy of Sciences, 72 Tzarigradsko chaussee blvd, BG-1784 Sofia (Bulgaria); ELI-NP, IFIN-HH, 30 Reactorului Str, MG-6 Bucharest-Magurele (Romania); Miro, S. [CEA-DEN, Service de Recherches de Métallurgie Physique, Laboratoire JANNUS, F-91191 Gif-sur-Yvette (France); Baillet, J. [Institut de Physique Nucléaire de Lyon (IPNL), Université Lyon 1, CNRS/IN2P3, 4 rue Enrico Fermi, 69622 Villeurbanne Cedex (France); Pradeilles, N.; Rapaud, O.; Maître, A. [SPCTS, UMR CNRS 7315, Centre Européen de la céramique, University of Limoges (France); Gosset, D. [CEA, Saclay, DMN-SRMA-LA2M, 91191 Gif-sur-Yvette (France)

    2015-12-15

    Already used as neutron absorber in the current French nuclear reactors, boron carbide (B{sub 4}C) is also considered in the future Sodium Fast Reactors of the next generation (Gen IV). Due to severe irradiation conditions occurring in these reactors, it is of primary importance that this material presents a high structural resistance under irradiation, both in the ballistic and electronic damage regimes. Previous works have shown an important structural resistance of boron carbide even at high neutron fluences. Nevertheless, the structural modification mechanisms due to irradiation are not well understood. Therefore the aim of this paper is to study structural modifications induced in B{sub 4}C samples in different damage regimes. The boron carbide pellets were shaped and sintered by using spark plasma sintering method. They were then irradiated in several conditions at room temperature or 800 °C, either by favoring the creation of ballistic damage (between 1 and 3 dpa), or by favoring the electronic excitations using 100 MeV swift iodine ions (S{sub e} ≈ 15 keV/nm). Ex situ micro-Raman spectroscopy and Doppler broadening of annihilation radiation technique with variable energy slow positrons were coupled to follow the evolution of the B{sub 4}C structure under irradiation.

  12. Structural modifications induced by ion irradiation and temperature in boron carbide B4C

    Science.gov (United States)

    Victor, G.; Pipon, Y.; Bérerd, N.; Toulhoat, N.; Moncoffre, N.; Djourelov, N.; Miro, S.; Baillet, J.; Pradeilles, N.; Rapaud, O.; Maître, A.; Gosset, D.

    2015-12-01

    Already used as neutron absorber in the current French nuclear reactors, boron carbide (B4C) is also considered in the future Sodium Fast Reactors of the next generation (Gen IV). Due to severe irradiation conditions occurring in these reactors, it is of primary importance that this material presents a high structural resistance under irradiation, both in the ballistic and electronic damage regimes. Previous works have shown an important structural resistance of boron carbide even at high neutron fluences. Nevertheless, the structural modification mechanisms due to irradiation are not well understood. Therefore the aim of this paper is to study structural modifications induced in B4C samples in different damage regimes. The boron carbide pellets were shaped and sintered by using spark plasma sintering method. They were then irradiated in several conditions at room temperature or 800 °C, either by favoring the creation of ballistic damage (between 1 and 3 dpa), or by favoring the electronic excitations using 100 MeV swift iodine ions (Se ≈ 15 keV/nm). Ex situ micro-Raman spectroscopy and Doppler broadening of annihilation radiation technique with variable energy slow positrons were coupled to follow the evolution of the B4C structure under irradiation.

  13. Erosion wear of boron carbide ceramic nozzles by abrasive air-jets

    International Nuclear Information System (INIS)

    Deng Jianxin

    2005-01-01

    Boron carbide nozzles were produced by hot pressing. The erosion wear of this nozzle caused by abrasive particle impact was investigated by abrasive air-jets. Silica, silicon carbide and alumina powders with different hardness were used as the erodent abrasive particles. Results showed that the hardness of the erodent particles played an important role with respect to the erosion wear of the boron carbide nozzles. As the hardness of the erodent particles increases, there is a dramatic increase in erosion rate of the nozzles. The nozzle entrance area suffered from severe abrasive impact under large impact angles, and generated maximum tensile stresses. The wear mechanisms of boron carbide nozzle at this area appeared to be entirely brittle in nature with the evidence of large scale-chipping, and exhibited a brittle fracture induced removal process. While at the nozzle center wall section, most of the particles traveled parallel to the nozzle wall, and showed minimum tensile stresses. The wear mode in this area of the nozzle changed from impact to sliding erosion, and the wear mechanisms appeared to be the lateral cracking owing to a surface fatigue fracture mechanism

  14. The diffusion bonding of silicon carbide and boron carbide using refractory metals

    International Nuclear Information System (INIS)

    Cockeram, B.V.

    1999-01-01

    Joining is an enabling technology for the application of structural ceramics at high temperatures. Metal foil diffusion bonding is a simple process for joining silicon carbide or boron carbide by solid-state, diffusive conversion of the metal foil into carbide and silicide compounds that produce bonding. Metal diffusion bonding trials were performed using thin foils (5 microm to 100 microm) of refractory metals (niobium, titanium, tungsten, and molybdenum) with plates of silicon carbide (both α-SiC and β-SiC) or boron carbide that were lapped flat prior to bonding. The influence of bonding temperature, bonding pressure, and foil thickness on bond quality was determined from metallographic inspection of the bonds. The microstructure and phases in the joint region of the diffusion bonds were evaluated using SEM, microprobe, and AES analysis. The use of molybdenum foil appeared to result in the highest quality bond of the metal foils evaluated for the diffusion bonding of silicon carbide and boron carbide. Bonding pressure appeared to have little influence on bond quality. The use of a thinner metal foil improved the bond quality. The microstructure of the bond region produced with either the α-SiC and β-SiC polytypes were similar

  15. The All Boron Carbide Diode Neutron Detector: Experiment and Modeling Approach

    International Nuclear Information System (INIS)

    Sabirianov, Ildar F.; Brand, Jennifer I.; Fairchild, Robert W.

    2008-01-01

    Boron carbide diode detectors, fabricated from two different polytypes of semiconducting boron carbide, will detect neutrons in reasonable agreement with theoretical expectations. The performance of the all boron carbide neutron detector differs, as expected, from devices where a boron rich neutron capture layer is distinct from the diode charge collection region (i.e. a conversion layer solid state detector). Diodes were fabricated from natural abundance boron (20% 10 B and 80% 11 B.) directly on the metal substrates and metal contacts applied to the films as grown. The total boron depth was on the order of 2 microns. This is clearly not a conversion-layer configuration. The diodes were exposed to thermal neutrons generated from a paraffin moderated plutonium-beryllium source in moderated and un-moderated, as well as shielded and unshielded experimental configurations, where the expected energy peaks at at 2.31 MeV and 2.8 MeV were clearly observed, albeit with some incomplete charge collection typical of thinner diode structures. The results are compared with other boron based thin film detectors and literature models. (authors)

  16. Boron carbide-coated carbon material, manufacturing method therefor and plasma facing material

    International Nuclear Information System (INIS)

    Suzuki, Takayuki; Kikuchi, Yoshihiro; Hyakki, Yasuo.

    1997-01-01

    The present invention concerns a plasma facing material suitable to a thermonuclear device. The material comprises a carbon material formed by converting the surface of a carbon fiber-reinforced carbon material comprising a carbon matrix and carbon fibers to a boron carbide, the material has a surface comprising vertically or substantially vertically oriented carbon fibers, and the thickness of the surface converted to boron carbide is reduced in the carbon fiber portion than in the carbon matrix portion. Alternatively, a carbon fiber-reinforced carbon material containing carbon fibers having a higher graphitizing degree than the carbon matrix is converted to boron carbide on the surface where the carbon fibers are oriented vertically or substantially vertically. The carbon fiber-reinforced material is used as a base material, and a resin material impregnated into a shaped carbon fiber product is carbonized or thermally decomposed carbon is filled as a matrix. The material of the present invention has high heat conduction and excellent in heat resistance thereby being suitable to a plasma facing material for a thermonuclear device. Electric specific resistivity of the entire coating layer can be lowered, occurrence of arc discharge is prevented and melting can be prevented. (N.H.)

  17. Control rod driving hydraulic device

    International Nuclear Information System (INIS)

    Sugano, Hiroshi.

    1993-01-01

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

  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. Control rod guide tube assemblies

    International Nuclear Information System (INIS)

    Jabsen, F.S.

    1979-01-01

    A nuclear fuel assembly including sleeves telescoped over end portions of control rod guide tubes which bear against internal shoulders of the sleeves. Upper ends of the sleeves protrude beyond a control rod guide tube spider and are locked in place by means of a resilient cellular lattice or lock that is seated in mating grooves in the outer surfaces of the sleeves. A grapple is provided for disengaging the entire lock structure spider and associated washers, springs and a grill from the end of the fuel assembly in order to enable these components to be removed and subsequently replaced on the fuel assembly after inspection and repair. (UK)

  20. Damage analysis of ceramic boron absorber materials in boiling water reactors and initial model for an optimum control rod management

    International Nuclear Information System (INIS)

    Schulz, W.

    2000-01-01

    Operating experience has proved so far that BWR control rods cannot be used for the total reactor life time as originally presumed, but instead has to be considered as a consumable article. After only few operating cycles, the mechanism of absorber failure has been shown to be neutron induced boron carbide swelling and stress cracking of the absorber tubes, followed by erosion of the absorber material. In the case that operation of such a control rod is continued in control cells, this can lead to an increase of the local power density distribution in the core and, under certain conditions, can even cause fuel rod damage. A non destructive testing method has been developed called 'UNDERWATER NEUTRON RADIOGRAPHY' applicable for any BWR control rod. 'Lead-control rods' being radiographed are used to evaluate their actual nuclear worth by the help of a special analytical procedure developed and verified by the author. Nuclear worth data plotted against bum up history data will allow to create an 'EMPIRIC MODEL'. This model includes the basic idea of operating control rods of a certain design first in a control position up to a target fluence limited to an amount just below the appearance of control rod washout. Afterwards they have to be moved in a shut down position to work therefor the total remaining holding period. The initial model is applicable to any CR-design as long as sufficient measuring-data and thus data about the nuclear worth are available. The results of these experiences are extrapolated to the whole reactor holding period. After modelling no further measurements of this particular control rod type are necessary in any reactor. The second focal point is to provide an APPROXIMATION EQUATION. By knowing the absorber radius, B 4 C density and absorber enclosure data an engineer will calculate reliably the working life of any control rod design on control position. indicated as maximum allowable neutron fluence margin until absorber wash-out starts. This

  1. NEUTRONIC REACTOR CONTROL ROD DRIVE APPARATUS

    Science.gov (United States)

    Oakes, L.C.; Walker, C.S.

    1959-12-15

    ABS>A suspension mechanism between a vertically movable nuclear reactor control rod and a rod extension, which also provides information for the operator or an automatic control signal, is described. A spring connects the rod extension to a drive shift. The extension of the spring indicates whether (1) the rod is at rest on the reactor, (2) the rod and extension are suspended, or (3) the extension alone is suspended, the spring controlling a 3-position electrical switch.

  2. Measuring device for control rod driving time

    International Nuclear Information System (INIS)

    Tanaka, Kazuhiko; Hanabusa, Masatoshi.

    1993-01-01

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

  3. Control rod drive for vertical movement

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  4. Managing the reactivity excess of the gas turbine-modular helium reactor by burnable poison and control rods

    International Nuclear Information System (INIS)

    Talamo, Alberto

    2006-01-01

    The gas turbine-modular helium reactor coupled to the deep burn in-core fuel management strategy offers the extraordinary capability to incinerate over 50% of the initial inventory of fissile material. This extraordinary feature, coming from an advanced and well tested fuel element design, which takes advantage of the TRISO particles technology, is maintained while the reactor is loaded with the most different types of fuels. In the present work, we assumed the reactor operating at the equilibrium of the fuel composition, obtained by a 6 years irradiation of light water reactor waste, and we investigated the effects of the introduction of the burnable poison and the control rods; we equipped the core with all the three types of control rods: operational, startup and shutdown ones. We employed as burnable poison natural erbium, due to the 167 Er increasing neutron capture microscopic cross-section in the energy range where the neutron spectrum exhibits the thermal peak; in addition, we utilized boron carbide, with 90% enrichment in 1 B, as the absorption material of the control rods. Concerning the burnable poison studies, we focused on the k eff value, the 167 Er mass during burnup, the influence of modifying the radius of the BISO particles kernel and the fuel and moderator coefficients of temperature. Concerning the control rods studies, we investigated the reactivity worth, the changes in the neutron flux profile due to a partial insertion, the influence of modifying the radius of the BISO particles kernel and the β eff , at the beginning of the operation

  5. Managing the reactivity excess of the gas turbine-modular helium reactor by burnable poison and control rods

    Energy Technology Data Exchange (ETDEWEB)

    Talamo, Alberto [Department of Nuclear and Reactor Physics, Royal Institute of Technology, Roslagstullsbacken 21, S-10691, Stockholm (Sweden)]. E-mail: alby@neutron.kth.se

    2006-01-15

    The gas turbine-modular helium reactor coupled to the deep burn in-core fuel management strategy offers the extraordinary capability to incinerate over 50% of the initial inventory of fissile material. This extraordinary feature, coming from an advanced and well tested fuel element design, which takes advantage of the TRISO particles technology, is maintained while the reactor is loaded with the most different types of fuels. In the present work, we assumed the reactor operating at the equilibrium of the fuel composition, obtained by a 6 years irradiation of light water reactor waste, and we investigated the effects of the introduction of the burnable poison and the control rods; we equipped the core with all the three types of control rods: operational, startup and shutdown ones. We employed as burnable poison natural erbium, due to the {sup 167}Er increasing neutron capture microscopic cross-section in the energy range where the neutron spectrum exhibits the thermal peak; in addition, we utilized boron carbide, with 90% enrichment in {sup 1}B, as the absorption material of the control rods. Concerning the burnable poison studies, we focused on the k {sub eff} value, the {sup 167}Er mass during burnup, the influence of modifying the radius of the BISO particles kernel and the fuel and moderator coefficients of temperature. Concerning the control rods studies, we investigated the reactivity worth, the changes in the neutron flux profile due to a partial insertion, the influence of modifying the radius of the BISO particles kernel and the {beta} {sub eff}, at the beginning of the operation.

  6. Approach to criticality in cold core of the Unit 1 of the cycle 10 of the Laguna Verde power plant with control rods similar in design to the original ones and the advanced ones; Aproximacion a criticidad en frio de la Unidad 1 del ciclo 10 de la CLV con barras de control de diseno similar al original y avanzadas

    Energy Technology Data Exchange (ETDEWEB)

    Perusquia, R.; Ramirez, J.R.; Ortiz, J.J.; Hernandez, J.L.; Montes, J.L. [ININ, A.P. 18-1027, 11801 Mexico D.F. (Mexico)]. E-mail: rpc@nuclear.inin.mx

    2004-07-01

    In the Laguna Verde power plant (CLV) worn out control bars have been replacing for control rods of advanced design. At the moment of the 109 rods of similar design to the original of the Unit 1 of the CLV, 14 of them have been substituted with two types of different advanced control rods. The same as the original rods the advanced rods use boron carbide as control element for neutrons, but additionally they use hafnium, to achieve a longer useful half life. The rods of similar design to the original are homogeneous axially on the other hand the advanced presenting several zones. In principle the advanced rods were designed to be neutronically compatible to the original rods. However the differences in the design and of constituents assumed to consider a different performance during the operation. Since the one numbers of advanced bars is growing with the number of cycles of the CLV reactor has requested to the National Institute of Nuclear Research (ININ) to evaluate in detailed form the performance of the control rods. Previously the calculations and analysis of cycle design that that were carried out assumed that control rods should be of similar design as the original. Using the HELIOS/CM-LENDS system, the effect of the inclusion of advanced rods was analyzed in the approximation of the criticality in cold (20 C) in the cycle 10 of the Unit 1. Four scenarios were poposed, where the type of the rod of substitution control is varied. The value of reactivity was calculated for the rods similar in design to the original and for the advanced one. By comparison of these values, it was found that the value of reactivity of the control rods of the advanced design was superior to the value for rods similar to the original ones by 7.6% for one of the advanced designs and of 13.6% for the other one. (Author)

  7. 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.)

  8. 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.)

  9. Evaluation of the Westinghouse 10B depletion for BWR control rods

    International Nuclear Information System (INIS)

    Vallgren, Christina

    2008-03-01

    The aim of this work was to establish the 10 B depletion model for CR 99 control rods by using the latest version of POLCA7. In order to obtain an understanding of the differences between the currently used 10 B depletion models implemented in POLCA4 at O3 and in SIMULATE-3 at OL1, and the latest improved model implemented in the latest POLCA7, this work has been performed in three different parts. The first part of the work was to find out how large differences there exist in 10 B depletion between the calculated data by using the latest core monitoring system (POLCA7 version 4.10.0) and the measured data obtained in the hot-cell laboratory in Studsvik. It was found that the 10 B depletion computed by the latest POLCA7 version is in good agreement with the measured data from Studsvik. A poor agreement with a conservative overestimation in 10 B depletion was also found between the old model and the measured data. The aim of the second part of the work was to compare the calculated 10 B depletion values for two CR 99 rods from Olkiluoto 1 with the calculated 10 B depletion value for a CR 99 rod from Oskarshamn 3, by using the new 10 B depletion model implemented in the latest POLCA7 version. Swelling measurements of the boron carbide pins, used as absorber material, have indicated that the 10 B depletion should be of similar magnitude for the rods in Olkiluoto 1 and the rod in Oskarshamn 3, whereas the calculated values by using the earlier 10 B depletion models on the process computers showed a difference of about 20 %. By using the new 10 B depletion model m POLCA7, it was found that the 10 B depletion in the two studied cases was similar to each other and, thus, the hypothesis of a linear relationship between B 4 C swelling and thermal neutron fluence was supported. This third part of the work was carried out at KKL, Switzerland, and focused on comparing the B depletion models used in Westinghouse/POLCA7 and KKL/PRESTO-2. It was found that there is a slight

  10. Evaluation of the Westinghouse 10B depletion for BWR control rods

    Energy Technology Data Exchange (ETDEWEB)

    Vallgren, Christina

    2008-03-15

    The aim of this work was to establish the 10B depletion model for CR 99 control rods by using the latest version of POLCA7. In order to obtain an understanding of the differences between the currently used 10B depletion models implemented in POLCA4 at O3 and in SIMULATE-3 at OL1, and the latest improved model implemented in the latest POLCA7, this work has been performed in three different parts. The first part of the work was to find out how large differences there exist in 10B depletion between the calculated data by using the latest core monitoring system (POLCA7 version 4.10.0) and the measured data obtained in the hot-cell laboratory in Studsvik. It was found that the 10B depletion computed by the latest POLCA7 version is in good agreement with the measured data from Studsvik. A poor agreement with a conservative overestimation in 10B depletion was also found between the old model and the measured data. The aim of the second part of the work was to compare the calculated 10B depletion values for two CR 99 rods from Olkiluoto 1 with the calculated 10B depletion value for a CR 99 rod from Oskarshamn 3, by using the new 10B depletion model implemented in the latest POLCA7 version. Swelling measurements of the boron carbide pins, used as absorber material, have indicated that the 10B depletion should be of similar magnitude for the rods in Olkiluoto 1 and the rod in Oskarshamn 3, whereas the calculated values by using the earlier 10B depletion models on the process computers showed a difference of about 20 %. By using the new 10B depletion model m POLCA7, it was found that the 10B depletion in the two studied cases was similar to each other and, thus, the hypothesis of a linear relationship between B{sub 4}C swelling and thermal neutron fluence was supported. This third part of the work was carried out at KKL, Switzerland, and focused on comparing the B depletion models used in Westinghouse/POLCA7 and KKL/PRESTO-2. It was found that there is a slight difference in

  11. 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

  12. Snubber assembly for a control rod drive

    International Nuclear Information System (INIS)

    1976-01-01

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

  13. Comparative studies of electrochemical properties of carbon nanotubes and nanostructured boron carbide

    Science.gov (United States)

    Singh, Paviter; Kaur, Gurpreet; Singh, Kulwinder; Singh, Bikramjeet; Kaur, Manjot; Kumar, Manjeet; Bala, Rajni; Kumar, Akshay

    2018-05-01

    Boron carbide (B4C) and carbon nanotubes (CNTs) have the potential to act as electrocatalyst as these material show bifunctional behavior. B4C and CNTs were synthesized using solvothermal method. B4C display great catalytic activity as compared to CNTs. Raman spectra confirmed the formation of nanostructured carbon nanotubes. The observed onset potential was smaller 1.58 V in case of B4C as compared to CNTs i.e. 1.96 V in cyclic voltammetry. B4C material can emerge as a promising bifunctional electrocatalyst for battery applications.

  14. Determination of nitrogen in boron carbide with the Leco UO-14 Nitrogen Determinator

    International Nuclear Information System (INIS)

    Gardner, R.D.; Ashley, W.H.; Henicksman, A.L.

    1977-11-01

    Use of various metals as fluxes for releasing nitrogen from boron carbide in the Leco Nitrogen Determinator was investigated. Metals such as iron, chromium, and molybdenum that wet the graphite crucible all promoted nitrogen release. Tin, copper, aluminum, and platinum did not wet the graphite and were of no value as fluxes. A procedure for sample handling and the resulting performance of the method are described. The precision at 0.06 to 0.6 percent nitrogen averaged 4 percent relative standard deviation

  15. Effect of material variables on the irradiation performance of boron carbide

    International Nuclear Information System (INIS)

    Basmajian, J.A.; Hollenberg, G.W.

    1980-01-01

    Boron carbide pellets were fabricated with variations in material parameters. These pellets were irradiated in the Experimental Breeder Reactor-II (EBR-II) to determine the effect of these variations on the performance. Helium release from the material and swelling of the pellets are the primary measures of performance. It was determined that material with a smaller grain size released more helium and swelled less. The pellets with boron-to-carbon ratios greater than 4 to 1 did not perform well. Iron additions improved the performance of the material while density variations had little effect

  16. Fuel followed control rod installation at AFRRI

    International Nuclear Information System (INIS)

    Moore, Mark; Owens, Chris; Forsbacka, Matt

    1992-01-01

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

  17. Monitoring device for withdrawing control rods

    International Nuclear Information System (INIS)

    Higashigawa, Yuichi.

    1985-01-01

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

  18. Control rods in LMFBRs: a physics assessment

    International Nuclear Information System (INIS)

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

    1982-08-01

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

  19. Rope wind-up type control rod

    International Nuclear Information System (INIS)

    Tsuji, Teruaki; Watanabe, Shigeru.

    1979-01-01

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

  20. Manufacturing method for boron carbide/carbon composite neutron shielding material

    International Nuclear Information System (INIS)

    Inoue, Takenori; Ukai, Shigeharu; Maruyama, Tadashi; Suya, Kiyoshi; Sunami, Yoshihiko.

    1994-01-01

    A less volatile binder pitch which is melted upon heating is used as a binder. Raw materials mainly comprising 60 to 85% by volume of a boron carbide powder and 15 to 40% by volume of a binder pitch are mixed, molded under pressure and heating at 480 to 600degC, then baked under non-pressurization, further impregnated with pitch under a reduced pressure and then baked again. The volume percentage of each of the materials is calculated based on the volume obtained by dividing the blending weight for each of raw materials with the intrinsic density respectively. The binding property relative to the boron carbide powder is improved by using a pitch having satisfactory melting performance and reduction of strength is decreased. Moreover, if the binder pitch is baked at about 2,000degC, it is easily converted into a graphitized tissues to have excellent slidability and fabricability. With such procedures, high bending strength and high heat conductivity can be ensured while keeping high boron content and neutron absorbing performance. (T.M.)

  1. The role of free carbon in the transport and magnetic properties of boron carbide

    International Nuclear Information System (INIS)

    Bandyopadhyay, A.K.; Beuneu, F.; Zuppiroli, L.; Beauvy, M.

    1984-01-01

    Boron carbide is a ceramic which has a wide field of application because of its mechanical and nuclear properties. This material is difficult to characterise due to the presence of different levels of disorder and inhomogeneities which are found in the usual available samples. The transport and magnetic properties of several samples of boron carbide have been measured from liquid helium to room temperature as a function of temperature and composition. We have attempted to attribute the different features of these properties to the different levels of disorder. The role of free carbon, in form of thin layers of graphite within the disordered semi-conducting matrix, was investigated in particular details, because it was either ignored or neglected by others. Free carbon is found to dominate the D.C. transport when its concentration is larger than 5%; while the principal features of the electron spin resonance (E.S.R.) line show a dominance of free carbon when the concentration is larger than 3.5%. Below these concentrations conductivities as well as spin relaxation rates do not depend very much on free carbon; neither these have been found to be correlated in a simple way to the stoichiometry. (author)

  2. Dynamic Failure and Fragmentation of a Hot-Pressed Boron Carbide

    Science.gov (United States)

    Sano, Tomoko; Vargas-Gonzalez, Lionel; LaSalvia, Jerry; Hogan, James David

    2017-12-01

    This study investigates the failure and fragmentation of a hot-pressed boron carbide during high rate impact experiments. Four impact experiments are performed using a composite-backed target configuration at similar velocities, where two of the impact experiments resulted in complete target penetration and two resulted in partial penetration. This paper seeks to evaluate and understand the dynamic behavior of the ceramic that led to either the complete or partial penetration cases, focusing on: (1) surface and internal failure features of fragments using optical, scanning electron, and transmission electron microscopy, and (2) fragment size analysis using state-of-the-art particle-sizing technology that informs about the consequences of failure. Detailed characterization of the mechanical properties and the microstructure is also performed. Results indicate that transgranular fracture was the primary mode of failure in this boron carbide material, and no stress-induced amorphization features were observed. Analysis of the fragment sizes for the partial and completely penetrated experiments revealed a possible correlation between larger fragment sizes and impact performance. The results will add insight into designing improved advanced ceramics for impact protection applications.

  3. Tritium release from fast neutron irradiated boron carbide

    International Nuclear Information System (INIS)

    Hollenberg, G.W.

    1977-01-01

    A high-energy neutron reaction with boron produces tritium. In the LMFBR control material, B 4 C, most of the tritium that is generated remains in the pellets. Potential retention mechanisms are discussed. 5 figures

  4. 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.)

  5. 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.)

  6. Studies on the influence of surface pre-treatments on electroless copper coating of boron carbide particles

    International Nuclear Information System (INIS)

    Deepa, J.P.; Resmi, V.G.; Rajan, T.P.D.; Pavithran, C.; Pai, B.C.

    2011-01-01

    Boron carbide is one of the hard ceramic particles which find application as structural materials and neutron shielding material due to its high neutron capture cross section. Copper coating on boron carbide particle is essential for the synthesis of metal-ceramic composites with enhanced sinterability and dispersibility. Surface characteristics of the substrate and the coating parameters play a foremost role in the formation of effective electroless coating. The effect of surface pre-treatment conditions and pH on electroless copper coating of boron carbide particles has been studied. Surface pre-treatement of B 4 C when compared to acid treated and alkali treated particles were carried out. Uniform copper coating was observed at pH 12 in alkali treated particles when compared to others due to the effective removal of inevitable impurities during the production and processing of commercially available B 4 C. A threshold pH 11 was required for initiation of copper coating on boron carbide particles. The growth pattern of the copper coating also varies depending on the surface conditions from acicular to spherical morphology.

  7. 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.)

  8. Structural models of increasing complexity for icosahedral boron carbide with compositions throughout the single-phase region from first principles

    Science.gov (United States)

    Ektarawong, A.; Simak, S. I.; Alling, B.

    2018-05-01

    We perform first-principles calculations to investigate the phase stability of boron carbide, concentrating on the recently proposed alternative structural models composed not only of the regularly studied B11Cp (CBC) and B12(CBC), but also of B12(CBCB) and B12( B4 ). We find that a combination of the four structural motifs can result in low-energy electron precise configurations of boron carbide. Among several considered configurations within the composition range of B10.5C and B4C , we identify in addition to the regularly studied B11Cp (CBC) at the composition of B4C two low-energy configurations, resulting in a new view of the B-C convex hull. Those are [B12 (CBC)]0.67[B12(B4)] 0.33 and [B12 (CBC)]0.67[ B12 (CBCB)]0.33, corresponding to compositions of B10.5C and B6.67C , respectively. As a consequence, B12(CBC) at the composition of B6.5C , previously suggested in the literature as a stable configuration of boron carbide, is no longer part of the B -C convex hull. By inspecting the electronic density of states as well as the elastic moduli, we find that the alternative models of boron carbide can provide a reasonably good description for electronic and elastic properties of the material in comparison with the experiments, highlighting the importance of considering B12(CBCB) and B12( B4 ), together with the previously proposed B11Cp (CBC) and B12(CBC), as the crucial ingredients for modeling boron carbide with compositions throughout the single-phase region.

  9. Control rod guide tube assembly

    International Nuclear Information System (INIS)

    Jabsen, F.S.

    1982-01-01

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

  10. 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.)

  11. Thick boron carbide coatings for protection of tokamak first wall and divertor

    International Nuclear Information System (INIS)

    Buzhinskij, O.I.; Semenets, Yu.M.

    1999-01-01

    A review of characteristics of various types of boron carbide coatings considered as candidate materials for protection of tokamak inner surfaces against high energy heat fluxes is presented. Such coatings are produced by various methods: chemical vapor deposition by means of chloride and fluoride techniques, gas conversion, plasma spray and reaction-sintering. Contrary to pure carbon materials, B 4 C has much lower chemical and high-temperature sputtering, is capable to oxygen gettering and lower hydrogen recycling. In contrast to thin boronization films, the thick coatings can resist high heat fluxes such as in tokamak divertors. Comparative analysis shows that coatings produced by the diffusion methods, such as fluoride CVD and gas conversion, are more resistent to heat loads, and one of the most promising candidates are the fluoride CVD coatings. (orig.)

  12. Novel semiconducting boron carbide/pyridine polymers for neutron detection at zero bias

    Energy Technology Data Exchange (ETDEWEB)

    Echeverria, Elena; Enders, A.; Dowben, P.A. [University of Nebraska-Lincoln, Department of Physics and Astronomy, Lincoln, NE (United States); James, Robinson; Chiluwal, Umesh; Gapfizi, Richard; Tae, Jae-Do; Driver, M. Sky; Kelber, Jeffry A. [University of North Texas, Department of Chemistry, Denton, TX (United States); Pasquale, Frank L. [University of North Texas, Department of Chemistry, Denton, TX (United States); Lam Research Corporation, PECVD Business Unit, Tualatin, OR (United States); Colon Santana, Juan A. [Center for Energy Sciences Research, Lincoln, NE (United States)

    2014-09-19

    Thin films containing aromatic pyridine moieties bonded to boron, in the partially dehydrogenated boron-rich icosahedra (B{sub 10}C{sub 2}H{sub X}), prove to be an effective material for neutron detection applications when deposited on n-doped (100) silicon substrates. The characteristic I-V curves for the heterojunction diodes exhibit strong rectification and largely unperturbed normalized reverse bias leakage currents with increasing pyridine content. The neutron capture generated pulses from these heterojunction diodes were obtained at zero bias voltage although without the signatures of complete electron-hole collection. These results suggest that modifications to boron carbide may result in better neutron voltaic materials. (orig.)

  13. Structural stability of boron carbide under pressure proven by spectroscopic studies up to 73 GPa

    Energy Technology Data Exchange (ETDEWEB)

    Chuvashova, Irina [Material Physics and Technology at Extreme Conditions, Laboratory of Crystallography, University of Bayreuth (Germany); Bayerisches Geoinstitut, University of Bayreuth (Germany); Gasharova, Biliana; Mathis, Yves-Laurent [IBPT, Karlsruhe Institute of Technology, Karlsruhe (Germany); Dubrovinsky, Leonid [Bayerisches Geoinstitut, University of Bayreuth (Germany); Dubrovinskaia, Natalia [Material Physics and Technology at Extreme Conditions, Laboratory of Crystallography, University of Bayreuth (Germany)

    2017-11-17

    Being a material of choice for lightweight armor applications, boron carbide has been intensively studied. Its behavior under pressure was investigated using both theoretical and experimental methods, such as powder X-ray diffraction and vibrational spectroscopy. As there is a discrepancy in experimental observations, in the presented work we studied vibrational properties of commercially available, ''nearly stoichiometric'' B{sub 4}C using IR and Raman spectroscopy up to 73 GPa. No phase transitions were found in the entire pressure range. Our results are at odds with the recent report of a phase transition in B{sub 4.3}C at about 40 GPa. (copyright 2017 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  14. Microscopic origin of the composition-dependent change of the thermal conductivity in boron carbides

    International Nuclear Information System (INIS)

    Emin, D.; Howard, I.A.; Green, T.A.; Beckel, C.L.

    1987-01-01

    Large grain polycrystalline boron carbides have a high-temperature thermal conductivity which changes from being characteristic of a crystal to being glass-like as the carbon content is reduced from its maximal value. We relate this phenomenon, to compositional changes within the three-atom intericosahedral chains. With a reduction of the carbon concentration from its maximal concentration (20%), a carbon atom within some of the three-atoms (CBC) intericosahedral chains is replaced by a boron atom, thereby producing CBB chains. We estimate that the CBB chains are significantly softer than the CBC chains. Thus, with this reduction of carbon content the intericosahedral chains are inhomogeneously softened. This suppresses the coherent transport of heat through the chains. The remaining thermal transport occurs incoherently through vibrationally inequivalent structural units, i.e. ''phonon hopping.''

  15. The irradiation behaviour of boron carbide/graphite between 800 and 1,1000C

    International Nuclear Information System (INIS)

    Hattenbach, K.; Hilgendorff, W.; Weiler, K.; Zimmermann, H.U.

    1975-01-01

    64 samples of boron carbide/graphite, a material used as burnable poison in high temperature reactors, were irradiated at temperatures between 800 and 1,100 0 C up to a fluence of 1-2 x 10 20 nvt. The following post-investigations were extended to dimensional measurements to determime a possible swelling or shrinking of the pellet, corrosion tests in completely desalinated water at 300 0 C, preparation of metallographic microsections to check for crack formation, determination of the helium hold back power and the thus involved gas chromatic analysis, as well as burn-up determinations by determining the boron 10/boron 11 ratio and the lithium concentration. (orig./LN) [de

  16. Defects in boron carbide: First-principles calculations and CALPHAD modeling

    International Nuclear Information System (INIS)

    Saengdeejing, Arkapol; Saal, James E.; Manga, Venkateswara Rao; Liu Zikui

    2012-01-01

    The energetics of defects in B 4+x C boron carbide and β-boron are studied through first-principles calculations, the supercell phonon approach and the Debye–Grüneisen model. It is found that suitable sublattice models for β-boron and B 4+x C are B 101 (B,C) 4 and B 11 (B,C) (B,C,Va) (B,Va) (B,C,Va), respectively. The thermodynamic properties of B 4+x C, β-boron, liquid and graphite are modeled using the CALPHAD approach based on the thermochemical data from first-principles calculations and experimental phase equilibrium data in the literature. The concentrations of various defects are then predicted as a function of carbon composition and temperature.

  17. Standard test methods for chemical, mass spectrometric, and spectrochemical analysis of nuclear-grade boron carbide

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2004-01-01

    1.1 These test methods cover procedures for the chemical, mass spectrometric, and spectrochemical analysis of nuclear-grade boron carbide powder and pellets to determine compliance with specifications. 1.2 The analytical procedures appear in the following order: Sections Total Carbon by Combustion and Gravimetry 7-17 Total Boron by Titrimetry 18-28 Isotopic Composition by Mass Spectrometry 29-38 Chloride and Fluoride Separation by Pyrohydrolysis 39-45 Chloride by Constant-Current Coulometry 46-54 Fluoride by Ion-Selective Electrode 55-63 Water by Constant-Voltage Coulometry 64-72 Impurities by Spectrochemical Analysis 73-81 Soluble Boron by Titrimetry 82-95 Soluble Carbon by a Manometric Measurement 96-105 Metallic Impurities by a Direct Reader Spectrometric Method 106-114

  18. Seismic scrammability of HTTR control rods

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  19. Development of a control rod drive

    International Nuclear Information System (INIS)

    1991-01-01

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

  20. 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.)

  1. Control rod for HTGR type reactor

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  2. Detection device for control rod interference

    International Nuclear Information System (INIS)

    Saito, Noboru.

    1984-01-01

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

  3. Position control device for a control rod

    International Nuclear Information System (INIS)

    Ono, Takehiko; Kusaka, Shuji.

    1976-01-01

    Purpose: To reliably prevent dangerous operation in the control of the position of the control rod by checking for abnormal pulse motor coil excitation voltage and, at the time of occurrence of abnormality, immediately holding the control rod stationary lest it should be moved to an unsafe position, this being accomplished excitation from a compensating excitation system. Constitution: In an FBR reactor, a circuit for memorizing the correct output states of individual drive signals at arbitrary instants and consequtively producing the memorized results is provided, and the output of the circuit and the actual drive signal are compared at all times to discriminate whether the drive signal being compared is normal or not. When the actual drive signal is abnormal, a series signal varying after a predetermined pattern is shifted to enable replacement of the actual drive signal, so that irrespective of whether the problem drive signal is ''on'' or ''off'', a drive signal of the correct pattern may be supplied to the pulse motor to hold the control rod and prevent it from being moved toward the dangerous side due to its own weight or other causes. (Horiuchi, T.)

  4. RODMOD: a code for control rod positioning

    International Nuclear Information System (INIS)

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

    1978-11-01

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

  5. Microcomputer system for controlling fuel rod length

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  6. Design of boron carbide-shielded irradiation channel of the outer irradiation channel of the Ghana Research Reactor-1 using MCNP.

    Science.gov (United States)

    Abrefah, R G; Sogbadji, R B M; Ampomah-Amoako, E; Birikorang, S A; Odoi, H C; Nyarko, B J B

    2011-01-01

    The MCNP model for the Ghana Research Reactor-1 was redesigned to incorporate a boron carbide-shielded irradiation channel in one of the outer irradiation channels. Extensive investigations were made before arriving at the final design of only one boron carbide covered outer irradiation channel; as all the other designs that were considered did not give desirable results of neutronic performance. The concept of redesigning a new MCNP model, which has a boron carbide-shielded channel is to equip the Ghana Research Reactor-1 with the means of performing efficient epithermal neutron activation analysis. After the simulation, a comparison of the results from the original MCNP model for the Ghana Research Reactor-1 and the new redesigned model of the boron carbide shielded channel was made. The final effective criticality of the original MCNP model for the GHARR-1 was recorded as 1.00402 while that of the new boron carbide designed model was recorded as 1.00282. Also, a final prompt neutron lifetime of 1.5245 × 10(-4)s was recorded for the new boron carbide designed model while a value of 1.5571 × 10(-7)s was recorded for the original MCNP design of the GHARR-1. Copyright © 2010 Elsevier Ltd. All rights reserved.

  7. 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)

  8. 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.

  9. 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

  10. Control rod housing alignment and repair method

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  11. Effect of the reinforced boron carbide particulate content of AA6061 alloy on formation of the passive film in seawater

    International Nuclear Information System (INIS)

    Katkar, V.A.; Gunasekaran, G.; Rao, A.G.; Koli, P.M.

    2011-01-01

    Highlights: → Presence of boron carbide increases the corrosion rate of A6061 alloy in seawater. → Increasing the B 4 C content decreases passive layer thickness. → Passive films formed on A6061 and its B 4 C composites are n-type semiconductors. - Abstract: The effect of boron carbide (B 4 C) reinforcement on the corrosion of AA6061 alloy was studied by investigating passive films formed in seawater. The higher passive current and its potential-dependence for these composites indicated formation of porous passive film. Electrochemical impedance spectroscopy (EIS) graph suggests that the alloy surface is partly or totally active. The formed passive film is n-type semiconductor junction in nature. The difference between corrosion potential (E corr ) and potential at zero charge (PZC) suggests that the chloride ions responsible for film breakdown exist within the passive film. A suitable mechanism is proposed for the passive film breakdown.

  12. Lattice dynamics of {alpha} boron and of boron carbide; Proprietes vibrationnelles du bore {alpha} et du carbure de bore

    Energy Technology Data Exchange (ETDEWEB)

    Vast, N

    1999-07-01

    The atomic structure and the lattice dynamics of {alpha} boron and of B{sub 4}C boron carbide have been studied by Density Functional Theory (D.F.T.) and Density Functional Perturbation Theory (D.F.P.T.). The bulk moduli of the unit-cell and of the icosahedron have been investigated, and the equation of state at zero temperature has been determined. In {alpha} boron, Raman diffusion and infrared absorption have been studied under pressure, and the theoretical and experimental Grueneisen coefficients have been compared. In boron carbide, inspection of the theoretical and experimental vibrational spectra has led to the determination of the atomic structure of B{sub 4}C. Finally, the effects of isotopic disorder have been modeled by an exact method beyond the mean-field approximation, and the effects onto the Raman lines has been investigated. The method has been applied to isotopic alloys of diamond and germanium. (author)

  13. 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

  14. Determination of isotopic composition of boron in boron carbide by TIMS and PIGE: an inter-comparison study

    International Nuclear Information System (INIS)

    Sasibhushan, K.; Rao, R.M.; Parab, A.R.; Alamelu, D.; Aggarwal, S.K.; Acharya, R.; Chhillar, S.; Pujari, P.K.

    2015-01-01

    The paper reports a comparison of results on the determination of isotopic composition of boron in boron carbide (B 4 C) samples by Thermal Ionisation Mass Spectrometry (TIMS) and Particle Induced Gamma ray Spectrometry (PIGE). B 4 C samples having varying boron isotopic composition (natural, enriched with respect to 10 B) and their synthetic mixtures) have been analysed by both the techniques. The 10 B atom% was found to be in the range of 20-67%. (author)

  15. 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)

  16. 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

  17. Control rod position detector for nuclear reactor

    International Nuclear Information System (INIS)

    Kudo, Mitsuru; Fujiwara, Hiroshi.

    1981-01-01

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

  18. Control rod for a nuclear reactor

    Science.gov (United States)

    Roman, Walter G.; Sutton, Jr., Harry G.

    1979-01-01

    A control rod assembly for a nuclear reactor is disclosed having a remotely disengageable coupling between the control rod and the control rod drive shaft. The coupling is actuated by first lowering then raising the drive shaft. The described motion causes axial repositioning of a pin in a grooved rotatable cylinder, each being attached to different parts of the drive shaft which are axially movable relative to each other. In one embodiment, the relative axial motion of the parts of the drive shaft is used either to couple or to uncouple the connection by forcing resilient members attached to the drive shaft into or out of shouldered engagement, respectively, with an indentation formed in the control rod.

  19. 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

  20. Detection device for control rod scram

    International Nuclear Information System (INIS)

    Ishiyama, Satoshi.

    1989-01-01

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

  1. 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.

  2. 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.)

  3. Control rod supporting device in reactor

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  4. Fabrication Of Control Rod System Of The RSG-GAS

    International Nuclear Information System (INIS)

    Sudirdjo, Hari; Setyono; Prasetya, Hendra

    2001-01-01

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

  5. Porosity determination of alumina and boron carbide ceramic samples by gamma ray transmission

    International Nuclear Information System (INIS)

    Moreira, Anderson Camargo; Appoloni, Carlos Roberto

    2009-01-01

    The aim of this work is to apply the Gamma Ray Transmission (GRT), a non destructive technique, for structural characterization of ceramic samples. With this technique, the porosity of Alumina (Al 2 O 3 ) and Boron Carbide (B 4 C) ceramic samples, in tablet format, was determined. The equipment employed is constituted by a 241 Am gamma ray source (59.6 keV and 100mCi), a 2''x2'' diameter NaI (Tl) scintillation detector coupled to a standard gamma ray transmission electronic and a micrometric and automated table for sample movement. The porosity profile of the samples shows a homogeneous porosity distribution, within the spatial resolution of the employed transmission system. The mean porosity determined for Al 2 O 3 and B 4 C were 17.8±1.3% and 3.87±0.43%, respectively. A statistical treatment of these results was performed and showed that the mean porosity values determinate by the GRT are the same as those supplied by the manufacturer. (author)

  6. Boron carbide reinforced aluminium matrix composite: Physical, mechanical characterization and mathematical modelling

    International Nuclear Information System (INIS)

    Shirvanimoghaddam, K.; Khayyam, H.; Abdizadeh, H.; Karbalaei Akbari, M.; Pakseresht, A.H.; Ghasali, E.; Naebe, M.

    2016-01-01

    This paper investigates the manufacturing of aluminium–boron carbide composites using the stir casting method. Mechanical and physical properties tests to obtain hardness, ultimate tensile strength (UTS) and density are performed after solidification of specimens. The results show that hardness and tensile strength of aluminium based composite are higher than monolithic metal. Increasing the volume fraction of B_4C, enhances the tensile strength and hardness of the composite; however over-loading of B_4C caused particle agglomeration, rejection from molten metal and migration to slag. This phenomenon decreases the tensile strength and hardness of the aluminium based composite samples cast at 800 °C. For Al-15 vol% B_4C samples, the ultimate tensile strength and Vickers hardness of the samples that were cast at 1000 °C, are the highest among all composites. To predict the mechanical properties of aluminium matrix composites, two key prediction modelling methods including Neural Network learned by Levenberg–Marquardt Algorithm (NN-LMA) and Thin Plate Spline (TPS) models are constructed based on experimental data. Although the results revealed that both mathematical models of mechanical properties of Al–B_4C are reliable with a high level of accuracy, the TPS models predict the hardness and tensile strength values with less error compared to NN-LMA models.

  7. Characterizing the Effect of Laser Power on Laser Metal Deposited Titanium Alloy and Boron Carbide

    Science.gov (United States)

    Akinlabi, E. T.; Erinosho, M. F.

    2017-11-01

    Titanium alloy has gained acceptance in the aerospace, marine, chemical, and other related industries due to its excellent combination of mechanical and corrosion properties. In order to augment its properties, a hard ceramic, boron carbide has been laser cladded with it at varying laser powers between 0.8 and 2.4 kW. This paper presents the effect of laser power on the laser deposited Ti6Al4V-B4C composites through the evolving microstructures and microhardness. The microstructures of the composites exhibit the formation of α-Ti phase and β-Ti phase and were elongated towards the heat affected zone. These phases were terminated at the fusion zone and globular microstructures were found growing epitaxially just immediately after the fusion zone. Good bondings were formed in all the deposited composites. Sample A1 deposited at a laser power of 0.8 kW and scanning speed of 1 m/min exhibits the highest hardness of HV 432 ± 27, while sample A4 deposited at a laser power of 2.0 kW and scanning speed of 1 m/min displays the lowest hardness of HV 360 ± 18. From the hardness results obtained, ceramic B4C has improved the mechanical properties of the primary alloy.

  8. High pressure synthesis and investigations of properties of boron allotropes and boron carbide

    International Nuclear Information System (INIS)

    Chuvashova, Irina

    2017-01-01

    This work aimed at the development of the high-pressure high-temperature (HPHT) synthesis of single crystals of boron allotropes and boron-rich compounds, which could be used further for precise investigations of their structures, properties, and behavior at extreme conditions. To summarize, the present work resulted in the HPHT synthesis of the first previously unknown non-icosahedral boron allotrope ζ-B. This finding confirmed earlier theoretical predictions, which stayed unproven for decades because of experimental challenges which couldn't be overcome until recently. Structural stability of α-B and β-B in the Mbar pressure range and B 13 C 2 up to 68 GPa was experimentally proven. Accurate measurements of the unit cell and B 12 icosahedra volumes of the stoichiometric boron carbide B 13 C 2 as a function of pressure led to conclusion that they undergo a similar reduction upon compression that is typical for covalently bonded solids. Neither 'molecular-like' nor 'inversed molecular-like' solid behavior upon compression was detected that has closed a long-standing scientific dispute. A comparison of the compressional behavior of B 13 C 2 with that of α-B and γ-B allotropes and B 4 C showed that it is determined by the types of bonding involved in the course of compression.

  9. Neutron shielding behavior of thermoplastic natural rubber/boron carbide composites

    Science.gov (United States)

    Mat Zali, Nurazila; Yazid, Hafizal; Megat Ahmad, Megat Harun Al Rashid

    2018-01-01

    Many shielding materials have been designed against the harm of different types of radiation to the human body. Today, polymer-based lightweight composites have been chosen by the radiation protection industry. In the present study, thermoplastic natural rubber (TPNR) composites with different weight percent of boron carbide (B4C) fillers (0% to 30%) were fabricated as neutron shielding through melt blending method. Neutron attenuation properties of TPNR/B4C composites have been investigated. The macroscopic cross section (Σ), half value layer (HVL) and mean free path length (λ) of the composites have been calculated and the transmission curves have been plotted. The obtained results show that Σ, HVL and λ greatly depend on the B4C content. Addition of B4C fillers into TPNR matrix were found to enhance the macroscopic cross section values thus decrease the mean free path length (λ) and half value layer (HVL) of the composites. The transmission curves exhibited that the neutron transmission of the composites decreased with increasing shielding thickness. These results showed that TPNR/B4C composites have high potential for neutron shielding applications.

  10. Effect of mechanical and thermal loading on boron carbide particles reinforced Al-6061 alloy

    International Nuclear Information System (INIS)

    Manjunatha, B.; Niranjan, H.B.; Satyanarayana, K.G.

    2015-01-01

    Metal Matrix Composites (MMC) considered as one of the ‘advanced materials’ have evoked growing interest during the last three decades due to their high performance and applications in strategic sectors. These composites exhibit unique and attractive properties over the monolithic alloys, but suffer from low ductility, which makes them not so attractive for some of the applications where high toughness is one of the design criteria. This limitation of MMCs has been overcome by resorting to various treatments such as mechanical and thermal loading. Considering very limited reports available on Al alloy reinforced with boron carbide (B 4 C) particles, this paper presents (i) preparation of Al-6061 alloy reinforced with 1.5–10 wt% B 4 C, (ii) subjecting them to mechanical and thermal treatments and (iii) characterization of all the above samples. Specific ultimate tensile strength and hardness of all the composites were higher than those of matrix. Also, these values increased with increasing amount of particles, with composites containing 8 wt% B 4 C showing the maximum values in all the three conditions. These observations are supported by the uniform distribution of particles in the matrix as observed in their microstructure

  11. High pressure synthesis and investigations of properties of boron allotropes and boron carbide

    Energy Technology Data Exchange (ETDEWEB)

    Chuvashova, Irina

    2017-06-12

    This work aimed at the development of the high-pressure high-temperature (HPHT) synthesis of single crystals of boron allotropes and boron-rich compounds, which could be used further for precise investigations of their structures, properties, and behavior at extreme conditions. To summarize, the present work resulted in the HPHT synthesis of the first previously unknown non-icosahedral boron allotrope ζ-B. This finding confirmed earlier theoretical predictions, which stayed unproven for decades because of experimental challenges which couldn't be overcome until recently. Structural stability of α-B and β-B in the Mbar pressure range and B{sub 13}C{sub 2} up to 68 GPa was experimentally proven. Accurate measurements of the unit cell and B{sub 12} icosahedra volumes of the stoichiometric boron carbide B{sub 13}C{sub 2} as a function of pressure led to conclusion that they undergo a similar reduction upon compression that is typical for covalently bonded solids. Neither 'molecular-like' nor 'inversed molecular-like' solid behavior upon compression was detected that has closed a long-standing scientific dispute. A comparison of the compressional behavior of B{sub 13}C{sub 2} with that of α-B and γ-B allotropes and B{sub 4}C showed that it is determined by the types of bonding involved in the course of compression.

  12. Stabilizing device for control rod tip

    International Nuclear Information System (INIS)

    Verdone, G.F.

    1982-01-01

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

  13. Control rod for FBR type reactor

    International Nuclear Information System (INIS)

    Nakai, Koichi.

    1993-01-01

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

  14. Control rod drives for HTGR type reactor

    International Nuclear Information System (INIS)

    Nishiguchi, Isoharu; Katagiri, Shigeo.

    1991-01-01

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

  15. Control rod drives for FBR type reactor

    International Nuclear Information System (INIS)

    Ikakura, Hiroaki.

    1990-01-01

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

  16. Protective guide structure for reactor control rod

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  17. Regulatory perspective on incomplete control rod insertions

    International Nuclear Information System (INIS)

    Chatterton, M.

    1997-01-01

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

  18. Method of driving control rod in reactor

    International Nuclear Information System (INIS)

    Osa, Hirotaka.

    1986-01-01

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

  19. Radiological characterization of spent control rod assemblies

    International Nuclear Information System (INIS)

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

    1995-10-01

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

  20. Overview of Japanese control rods development program

    International Nuclear Information System (INIS)

    Koyama, M.

    1984-01-01

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

  1. Characterization of boron carbide particulate reinforced in situ copper surface composites synthesized using friction stir processing

    Energy Technology Data Exchange (ETDEWEB)

    Sathiskumar, R., E-mail: sathiscit2011@gmail.com [Department of Mechanical Engineering, Coimbatore Institute of Technology, Coimbatore, 641 014 Tamil Nadu (India); Murugan, N., E-mail: murugan@cit.edu.in [Department of Mechanical Engineering, Coimbatore Institute of Technology, Coimbatore, 641 014 Tamil Nadu (India); Dinaharan, I., E-mail: dinaweld2009@gmail.com [Department of Mechanical Engineering, V V College of Engineering, Tisaiyanvilai, 627 657 Tamil Nadu (India); Vijay, S.J., E-mail: vijayjoseph@karunya.edu [Centre for Research in Metallurgy (CRM), School of Mechanical Sciences, Karunya University, Coimbatore, 641 114 Tamil Nadu (India)

    2013-10-15

    Friction stir processing has evolved as a novel solid state technique to fabricate surface composites. The objective of this work is to apply the friction stir processing technique to fabricate boron carbide particulate reinforced copper surface composites and investigate the effect of B{sub 4}C particles and its volume fraction on microstructure and sliding wear behavior of the same. A groove was prepared on 6 mm thick copper plates and packed with B{sub 4}C particles. The dimensions of the groove was varied to result in five different volume fractions of B{sub 4}C particles (0, 6, 12, 18 and 24 vol.%). A single pass friction stir processing was done using a tool rotational speed of 1000 rpm, travel speed of 40 mm/min and an axial force of 10 kN. Metallurgical characterization of the Cu/B{sub 4}C surface composites was carried out using optical microscope and scanning electron microscope. The sliding wear behavior was evaluated using a pin-on-disk apparatus. Results indicated that the B{sub 4}C particles significantly influenced the area, dispersion, grain size, microhardness and sliding wear behavior of the Cu/B{sub 4}C surface composites. When the volume fraction of B{sub 4}C was increased, the wear mode changed from microcutting to abrasive wear and wear debris was found to be finer. Highlights: • Fabrication of Cu/B{sub 4}C surface composite by friction stir processing • Analyzing the effect of B{sub 4}C particles on the properties of Cu/B4C surface composite • Increased volume fraction of B{sub 4}C particles reduced the area of surface composite. • Increased volume fraction of B{sub 4}C particles enhanced the microhardness and wear rate. • B{sub 4}C particles altered the wear mode from microcutting to abrasive.

  2. Surface impurity removal from DIII-D graphite tiles by boron carbide grit blasting

    International Nuclear Information System (INIS)

    Lee, R.L.; Hollerbach, M.A.; Holtrop, K.L.; Kellman, A.G.; Taylor, P.L.; West, W.P.

    1993-11-01

    During the latter half of 1992, the DIII-D tokamak at General Atomics (GA) underwent several modifications of its interior. One of the major tasks involved the removal of accumulated metallic impurities from the surface of the graphite tiles used to line the plasma facing surfaces inside of the tokamak. Approximately 1500 graphite tiles and 100 boron nitride tiles from the tokamak were cleaned to remove the metallic impurities. The cleaning process consisted of several steps: the removed graphite tiles were permanently marked, surface blasted using boron carbide (B 4 C) grit media (approximately 37 μm. diam.), ultrasonically cleaned in ethanol to remove loose dust, and outgassed at 1000 degrees C. Tests were done using, graphite samples and different grit blaster settings to determine the optimum propellant and abrasive media pressures to remove a graphite layer approximately 40-50 μm deep and yet produce a reasonably smooth finish. EDX measurements revealed that the blasting technique reduced the surface Ni, Cr, and Fe impurity levels to those of virgin graphite. In addition to the surface impurity removal, tritium monitoring was performed throughout the cleaning process. A bubbler system was set up to monitor the tritium level in the exhaust gas from the grit blaster unit. Surface wipes were also performed on over 10% of the tiles. Typical surface tritium concentrations of the tiles were reduced from about 500 dpm/100 cm 2 to less than 80 dpm/100 cm 2 following the cleaning. This tile conditioning, and the installation of additional graphite tiles to cover a high fraction of the metallic plasma facing surfaces, has substantially reduced metallic impurities in the plasma discharges which has allowed rapid recovery from a seven-month machine opening and regimes of enhanced plasma energy confinement to be more readily obtained. Safety issues concerning blaster operator exposure to carcinogenic metals and radioactive tritium will also be addressed

  3. A nuclear reactor with buffered control rods

    International Nuclear Information System (INIS)

    Bevilacqua, F.

    1974-01-01

    The control rods for, e.g., water-cooled reactors are fastened as units on common crossbars in vertical downward direction. The fastening on the crossbar is achieved by means of cross-shaped parts, e.g., in the shape of a double 'H'. A cylinder connected with a drive rod in normal operation is joined to each of the crossbars. In an emergency shut-down, this connection is interrupted and the control rod unit drops into the core through the action of gravity. Its fall is slowed down by a cushion or shock absorbing unit. For this purpose a piston is provided mounted on the supporting plate below the cylinder and guided within it. In the cylinder, the coolant is contained as damping medium. An upper opening in the cylinder serves as a ventilation hole. The movement of the piston is limited by a stopping part within the cylinder and slowed down by a spiral spring. (DG) [de

  4. 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.)

  5. Influencing factors and kinetics analysis on the leaching of iron from boron carbide waste-scrap with ultrasound-assisted method.

    Science.gov (United States)

    Li, Xin; Xing, Pengfei; Du, Xinghong; Gao, Shuaibo; Chen, Chen

    2017-09-01

    In this paper, the ultrasound-assisted leaching of iron from boron carbide waste-scrap was investigated and the optimization of different influencing factors had also been performed. The factors investigated were acid concentration, liquid-solid ratio, leaching temperature, ultrasonic power and frequency. The leaching of iron with conventional method at various temperatures was also performed. The results show the maximum iron leaching ratios are 87.4%, 94.5% for 80min-leaching with conventional method and 50min-leaching with ultrasound assistance, respectively. The leaching of waste-scrap with conventional method fits the chemical reaction-controlled model. The leaching with ultrasound assistance fits chemical reaction-controlled model, diffusion-controlled model for the first stage and second stage, respectively. The assistance of ultrasound can greatly improve the iron leaching ratio, accelerate the leaching rate, shorten leaching time and lower the residual iron, comparing with conventional method. The advantages of ultrasound-assisted leaching were also confirmed by the SEM-EDS analysis and elemental analysis of the raw material and leached solid samples. Copyright © 2017 Elsevier B.V. All rights reserved.

  6. 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.)

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

    International Nuclear Information System (INIS)

    Baron, Clemens

    2008-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-07-01

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

  9. Hydrothermal development and characterization of the wear-resistant boron carbide from Pandanus: a natural carbon precursor

    Science.gov (United States)

    Saritha Devi, H. V.; Swapna, M. S.; Ambadas, G.; Sankararaman, S.

    2018-04-01

    Boron carbide (B4C) is a prominent semiconducting material that finds applications in the field of science and technology. The excellent physical, thermal and electronic properties make it suitable as ceramic armor, wear-resistant, lens polisher and neutron absorber in the nuclear industry. The existing methods of synthesis of boron carbide involve the use of toxic chemicals that adversely affect the environment. In the present work, we report for the first time the use of the hydrothermal method, for converting the cellulose from Pandanus leaves as the carbon precursor for the synthesis of B4C. The carbon precursor is changed into porous functionalized carbon by treating with sodium borohydride (NaBH4), followed by treating with boric acid to obtain B4C. The samples are characterized by scanning electron microscopy, X-ray diffraction, Fourier transform infrared, Raman, photoluminescent and Ultraviolet-Visible absorption spectroscopy. The formation of B4C from natural carbon source— Pandanus presents an eco-friendly, economic and non-toxic approach for the synthesis of refractory carbides.

  10. Friction and wear performance of diamond-like carbon, boron carbide, and titanium carbide coatings against glass

    International Nuclear Information System (INIS)

    Daniels, B.K.; Brown, D.W.; Kimock, F.M.

    1997-01-01

    Protection of glass substrates by direct ion beam deposited diamond-like carbon (DLC) coatings was observed using a commercial pin-on-disk instrument at ambient conditions without lubrication. Ion beam sputter-deposited titanium carbide and boron carbide coatings reduced sliding friction, and provided tribological protection of silicon substrates, but the improvement factor was less than that found for DLC. Observations of unlubricated sliding of hemispherical glass pins at ambient conditions on uncoated glass and silicon substrates, and ion beam deposited coatings showed decreased wear in the order: uncoated glass>uncoated silicon>boron carbide>titanium carbide>DLC>uncoated sapphire. Failure mechanisms varied widely and are discussed. Generally, the amount of wear decreased as the sliding friction decreased, with the exception of uncoated sapphire substrates, for which the wear was low despite very high friction. There is clear evidence that DLC coatings continue to protect the underlying substrate long after the damage first penetrates through the coating. The test results correlate with field use data on commercial products which have shown that the DLC coatings provide substantial extension of the useful lifetime of glass and other substrates. copyright 1997 Materials Research Society

  11. Fabrication and tribological response of aluminium 6061 hybrid composite reinforced with bamboo char and boron carbide micro-fillers

    Science.gov (United States)

    Chethan, K. N.; Pai, Anand; Keni, Laxmikant G.; Singhal, Ashish; Sinha, Shubham

    2018-02-01

    Metal matrix composites (MMCs) have a wide scope of industrial applications and triumph over conventional materials due to their light weight, higher specific strength, good wear resistance and lower coefficient of thermal expansion. The present study aims at establishing the feasibility of using Bamboo charcoal particulate and boron carbide as reinforcements in Al-6061 alloy matrix and to investigate their effect on the wear of composites taking into consideration the interfacial adhesion of the reinforcements in the alloy. Al-6061 alloy was chosen as a base metallic alloy matrix. Sun-dried bamboo canes were used for charcoal preparation with the aid of a muffle furnace. The carbon content in the charcoal samples was determined by EDS (energy dispersive spectroscopy). In present study, stir casting technique was used to prepare the samples with 1%, 2%, and 3% weight of bamboo charcoal and boron carbide with Al-6061. The fabricated composites were homogenised at 570°C for 6 hours and cooled at room temperature. Wear studies were carried out on the specimens with different speed and loads. It was found that wear rate and coefficient of friction decreased with increase in the reinforcement content.

  12. Effect of sintering temperature and boron carbide content on the wear behavior of hot pressed diamond cutting segments

    Directory of Open Access Journals (Sweden)

    Islak S.

    2015-01-01

    Full Text Available The aim of this study was to investigate the effect of sintering temperature and boron carbide content on wear behavior of diamond cutting segments. For this purpose, the segments contained 2, 5 and 10 wt.% B4C were prepared by hot pressing process carried out under a pressure of 35 MPa, at 600, 650 and 700 °C for 3 minutes. The transverse rupture strength (TRS of the segments was assessed using a three-point bending test. Ankara andesite stone was cut to examine the wear behavior of segments with boron carbide. Microstructure, surfaces of wear and fracture of segments were determined by scanning electron microscopy (SEM-EDS, and X-ray diffraction (XRD analysis. As a result, the wear rate decreased significantly in the 0-5 wt.% B4C contents, while it increased in the 5-10 wt.% B4C contents. With increase in sintering temperature, the wear rate decreased due to the hard matrix.

  13. Control rod drives for nuclear reactors

    International Nuclear Information System (INIS)

    Okada, Shige.

    1981-01-01

    Purpose: To enable the detection for the raptures of the bellows in control rod drives in LMFBR type reactor, by recycling seal gases inside the bellows and measuring the radioactivity in the recycling passage. Constitution: In the control drives, outer extension pipe is surrounded by the bellows, which is put between the cylindrical biological shieldings around the upper potion of an upper guide tube and the disk-like seal members provided at the lower flange of the outer extension pipe. Thus, the inside device of control rod is isolated from the coolants and the cover gases. The outer extension pipe is provided with a suction channel and a return channel. These channels are connected to a seal gas recycling pipeway, a pump and a radioactivity detector, where the seal gases in the bellows is recycling. If failures should occur in the bellows, cover gas leaks into the seal gas and recycles, whereby radioactivity is detected and alarmed. (J.P.N.)

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

    International Nuclear Information System (INIS)

    Magnusson, Per

    2008-01-01

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

  15. Control rod for nuclear reactor

    International Nuclear Information System (INIS)

    Tada, Kaoru; Kawano, Shohei

    1998-01-01

    A guide roller is prepared by forming an oxide membrane on the surface of a molded roller product comprising, as a material, a deposition-reinforced type nickel-based alloy reinforced by deposition of fine particles by applying a heat treatment to a nickel-based alloy. When the guide roller is used in reactor water, since the roller has an oxide membrane on the surface, leaching of nickel to reactor water is reduced, and radioactive corrosive products including cobalt 58 are reduced to decrease an operator's exposure dose upon periodical inspections of a plant. The oxide membrane is formed by applying heat treatment under an oxidative atmosphere. Then, the amount of abrasion of pins and rollers in association with start-up or shut down of a reactor and control of the power can be reduced thereby enabling to suppress increase of radiation dose due to cobalt 60 and cobalt 58. (N.H.)

  16. Control rod guide tube cleaning device

    International Nuclear Information System (INIS)

    Tsuji, Tadashi; Shiota, Yoshiaki.

    1990-01-01

    Since there was no exclusive device for cleaning control rods, no effective cleaning could not be conducted and there was a possibility that obstacles may not be recovered. Then, there are disposed a first pump for supplying pressurized water, a spray nozzle for forming a swirling flow in a control rod guide tube, a second pump for pressurizing water introduced by a sucking pipeline and a collecting device for recovering obstacles intruding to water from the second pump. The pressurized water supplied from the first pump is introduced to a head passing through a blowing pipe and jetted from the spray nozzle to the control rod guide tube. In this case, a swirling stream occurs and obstacles in the control guide tube are mixed into water. The water containing the obstacles passes from the sucking port through a pipeline, introduced to the second pump and recovered to the collecting device. Since there is no water staying portion upon cleaning operation, the obstacles accumulating over the entire region of the bottom of the guide tube can be recovered reliably and efficiently. (N.H.)

  17. The lifetime of the control rod drives

    International Nuclear Information System (INIS)

    Avet, B.; Cauquelin, C.

    1989-01-01

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

  18. Passive cooling of control rod drive mechanisms

    International Nuclear Information System (INIS)

    Hankinson, M.F.; Schwirian, R.E.

    1992-01-01

    A method and apparatus are provided for passively cooling the control rod drive mechanisms (CRDMs) in the reactor vessel of a nuclear power plant. Passive cooling is achieved by dispersing a plurality of chimneys within the CRDM array in positions where a control rod is not required. The chimneys induce convective air currents which cause ambient air from within the containment to flow over the CRDM coils. The air heated by the coils is guided into inlets in the chimneys by baffles. The chimney is insulated and extends through the seismic support platform and missile shield disposed above the closure head. A collar of adjustable height mates with plate elements formed at the distal end of the CRDM pressure housings by an interlocking arrangement so that the seismic support platform provides lateral restraint for the chimneys. (Author)

  19. 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)

  20. Magnetic switch for reactor control rod

    International Nuclear Information System (INIS)

    Germer, J.H.

    1986-01-01

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

  1. 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

  2. Verification test of control rod system for HTR-10

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  3. Characterization and evaluation of boron carbide for plate-impact conditions

    International Nuclear Information System (INIS)

    Holmquist, T. J.; Johnson, G. R.

    2006-01-01

    This article addresses the response of boron carbide (B 4 C) to high-velocity impact. The authors previously characterized this material in 1999, using the Johnson-Holmquist [AIP Conf. Proc. 309, 981 (1994)] (JH-2) model. Since then, there have been additional experimental data presented in the literature that better describe the hydrostatic pressure (including a phase change). In addition, a series of plate-impact experiments (one-dimensional, uniaxial strain) that used configurations that produced either a shock, a shock release, or a shock reshock was performed. These experiments provide material behavior regarding the damage, failed strength, and hydrostat for which previously there has been little or no data. Constitutive model constants were obtained for the Johnson-Holmquist-Beissel [J. Appl. Phys. 94, 1639 (2003)] model using some of these plate-impact experiments. Computations of all the experiments were performed and analyzed to better understand the material response. The analysis provided the following findings: (1) The material fails and loses strength when the Hugoniot elastic limit (HEL) is exceeded. (2) The material has significant strength after failure and gradually increases as the pressure increases. (3) The shear modulus does not degrade when the material fails (as has been postulated), but rather increases. (4) When the material is reloaded from an initial shocked (failed) state, the loading appears to be elastic, indicating the material is not on the yield surface after failure. To provide more insight into the behavior of B 4 C, the strength versus pressure response was compared to that of silicon carbide (SiC). The strength of SiC increases as the pressure increases beyond the HEL, probably due to pressure hardening or strain hardening. It appears that B 4 C does not experience any hardening effects and fails at the HEL. Although the HEL for B 4 C is higher than that of SiC, the hardening ability of SiC produces a similar maximum strength

  4. Installing and detaching apparatus for a control rod drive mechanism

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  5. High temperature synthesis of ceramic composition by directed reaction of molten titanium or zirconium with boron carbide

    International Nuclear Information System (INIS)

    Johnson, W.B.

    1990-01-01

    Alternative methods of producing ceramics and ceramic composites include sintering, hot pressing and more recently hot isostatic pressing (HIP) and self-propagating high temperature synthesis (SHS). Though each of these techniques has its advantages, each suffers from several restrictions as well. Sintering may require long times at high temperatures and for most materials requires sintering aids to get full density. These additives can, and generally do, change (often degrade) the properties of the ceramic. Hot pressing and hot isostatic pressing are convenient methods to quickly prepare samples of some materials to full density, but generally are expensive and may damage some types of reinforcements during densification. This paper focuses on the preparation and processing of composites prepared by the directed reaction of molten titanium or zirconium with boron carbide. Advantages and disadvantages of this approach when compared to traditional methods are discussed, with reference to specific examples. Examples of microstructure are properties of these materials are reported

  6. Phase Field Theory and Analysis of Pressure-Shear Induced Amorphization and Failure in Boron Carbide Ceramic

    Directory of Open Access Journals (Sweden)

    John D. Clayton

    2014-07-01

    Full Text Available A nonlinear continuum phase field theory is developed to describe amorphization of crystalline elastic solids under shear and/or pressure loading. An order parameter describes the local degree of crystallinity. Elastic coefficients can depend on the order parameter, inelastic volume change may accompany the transition from crystal to amorphous phase, and transitional regions parallel to bands of amorphous material are penalized by interfacial surface energy. Analytical and simple numerical solutions are obtained for an idealized isotropic version of the general theory, for an element of material subjected to compressive and/or shear loading. Solutions compare favorably with experimental evidence and atomic simulations of amorphization in boron carbide, demonstrating the tendency for structural collapse and strength loss with increasing shear deformation and superposed pressure.

  7. Safety analysis of control rod drive computers

    International Nuclear Information System (INIS)

    Ehrenberger, W.; Rauch, G.; Schmeil, U.; Maertz, J.; Mainka, E.U.; Nordland, O.; Gloee, G.

    1985-01-01

    The analysis of the most significant user programmes revealed no errors in these programmes. The evaluation of approximately 82 cumulated years of operation demonstrated that the operating system of the control rod positioning processor has a reliability that is sufficiently good for the tasks this computer has to fulfil. Computers can be used for safety relevant tasks. The experience gained with the control rod positioning processor confirms that computers are not less reliable than conventional instrumentation and control system for comparable tasks. The examination and evaluation of computers for safety relevant tasks can be done with programme analysis or statistical evaluation of the operating experience. Programme analysis is recommended for seldom used and well structured programmes. For programmes with a long, cumulated operating time a statistical evaluation is more advisable. The effort for examination and evaluation is not greater than the corresponding effort for conventional instrumentation and control systems. This project has also revealed that, where it is technologically sensible, process controlling computers or microprocessors can be qualified for safety relevant tasks without undue effort. (orig./HP) [de

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

    International Nuclear Information System (INIS)

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

    1999-10-01

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

  9. 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

  10. 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

  11. 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.)

  12. 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

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

    International Nuclear Information System (INIS)

    Becvar, J.; Saroch, V.

    1977-01-01

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

  14. Experiment on thermohydraulics of simulated control rod

    International Nuclear Information System (INIS)

    Ogawa, Masuro; Ouchi, Mitsuo; Akino, Norio; Fujimura, Kaoru; Shiina, Yasuaki; Kawamura, Hiroshi

    1984-10-01

    A thermohydraulic study of a control rod channel is required for the core design of the Very High Temperature Gas Cooled Reactor (VHTR). A non-heating experiment with air flow was performed prior to heating experiment with helium flow. Experimental results on stability of flow, flow rate distribution and pressure drop of the control rod channel are reported. In a test section of the experimental apparatus, five simulated control subrods were suspended vertically in a circular duct. Their dimension was in coincide with those of the Detailed Disign (I) of the VHTR. Air of atomospheric pressure was used as a coolant gas, which flowed in inner and outer paths of the subrods. Total flow rate ranged from 0.0011 to 0.0062 kg/s. Flow rate distribution and pressure drop were obtained for various flow rates. Velocity fluctuation in the channel was also observed using a hot wire anemometer. From these experiments, it was found that the flow rate distribution was nearly the same as a disigned value and that turbulent and laminar flows were simultaneously realized in outer and inner paths respectively. These observations supported a feasibility of the present design. (author)

  15. 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.)

  16. 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

  17. Control rod guide tube of nuclear reactor

    International Nuclear Information System (INIS)

    Suda, Yoshitaka; Ito, Kenji; Matsumoto, Kunio.

    1994-01-01

    Zr having a residual tensile stress of 3 to 10kg/mm 2 in a circumferential direction is used for the main ingredient of a control guide tube of a nuclear reactor. For this purpose, an appropriate correction method such as a roll-correction, tension-correction and press-correction method is applied to an existent Zr-base alloy tube with no substantial residual stress. If the residual tensile stress in the circumferential direction is smaller than 3kg/mm 2 , an effect sufficient to suppress irradiation growth is not obtainable, if it exceeds 10kg/mm 2 , dimensional changes, cracks or the like occurs locally since the wall thickness of the control rod guide tube is small and, accordingly, this often results in failed products as the control guide tube. (N.H.)

  18. 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

  19. Development and testing of control rod drives for ship reactors

    International Nuclear Information System (INIS)

    Bruelheide, K.; Mundt, D.; Peters, C.-H.; Manthey, H.-J.

    1978-01-01

    The following paper deals with the development and testings of a new control rod drive design for marine reactors. Starting from the good operating experience with the advanced pressurized water reactor (FDR) of the NS OTTO HAHN a control rod drive system with an hermetically sealed drive principle was developed. A prototype control rod drive system was put through extensive tests and developed ready for standard production at the 'Gesellschaft fuer Kernenergieverwertung in Schiffbau und Schiffahrt'

  20. The analytic method for calculating the control rod worth

    International Nuclear Information System (INIS)

    Kim, Han Gon; Lee, Byeong Ho; Chang, Soon Heung

    1989-01-01

    We calculated the control rod worth in this paper. To avoid complexity, we did not consider burnable poisons and soluble boron. The system was localized within one assembly. The control rod was treated as not an absorber but an another boundary. Thus all of the group constants were unchanged before and after control rod insertion. And we discussed the method for calculation of the reactivity of the whole core

  1. Analysis of control rod worth in experimental fast reactor JOYO

    International Nuclear Information System (INIS)

    Arii, Y.; Aoyama, T.; Okimoto, Y.; Yoshida, A.; Mizoo, N.

    1988-01-01

    In JOYO, the measurement of control rod worths have been carried out in the beginning of the each cycle, using both period method and neutron source multiplication method. In this paper, the calculational method of control rod worths in the design stage and the comparison with the design values and measured ones are shown. The reasons that the control rod worths change slightly in each cycle, are also investigated. (author). 13 figs, 12 tabs

  2. Analysis of control rod behavior based on numerical simulation

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  3. Pulverization of boron element and proportions of boron carbide in boron; Broyage de bore element et dosage de carbure de bore dans le bore

    Energy Technology Data Exchange (ETDEWEB)

    Lang, F M; Finck, C [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1956-07-01

    It is possible to reduce boron element into fine powder by means of a mortar and pestle made of sintered boron carbide, the ratio of boron carbide introduced being less than one per cent. Boron element at our disposal is made of sharp edged, dark brown, little grains of average size greater than 5 {mu}. Grain sizes smaller than 1{mu} are required for applying thin layers of such boron. (author) [French] Il est possible de pulveriser finement du bore element au moyen de mortier et pilon en carbure de bore fritte, le taux de carbure de bore introduit etant inferieur a 1 pour cent. Le bore element dont nous disposons est constitue de petits grains brun fonce, a aretes vives, de dimension moyenne superieure a 5 {mu}. L'application de ce bore en couches minces demande des grains de dimensions inferieures a 1 {mu}. (aute0008.

  4. Seismic analysis of hydraulic control rod driving system

    International Nuclear Information System (INIS)

    Zheng, Yanhua; Bo, Hanliang; Dong, Duo

    2002-01-01

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

  5. 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)

  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. 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)

  8. Determination of boron in graphite, boron carbide and glass by ICP-MS, ICP-OES and conventional wet chemical methods

    International Nuclear Information System (INIS)

    Venkatesh, K.; Kamble, Granthali S.; Venkatesh, Manisha; Kumar, Sanjukta A.; Reddy, A.V.R.

    2014-01-01

    Boron is an important element of interest in nuclear reactor materials due to its high neutron absorption cross section (σ 0 =3837 barns for 10 B). In the present paper, R and D work and routinely used methods have been described for the analysis of case samples (1) Graphite where boron is present at trace levels, (2) Boron Carbide having boron concentration of about 80% and (3) Glass containing 4-6 % boron. (author)

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

    International Nuclear Information System (INIS)

    Minarcin, Miroslav; Elko, Marek

    1998-01-01

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

  10. 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.)

  11. 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

  12. Hydraulic system for driving control rods

    International Nuclear Information System (INIS)

    Okuzumi, Naoaki.

    1982-01-01

    Purpose: To enable safety reactor shut down upon occurrence of an abnormal excess pressure in a hydraulic control unit. Constitution: The actuation pressure for a pressure switch that generates a scram signal is set lower than the release pressure set to a pressure release valve. Thus, if the pressure of nitrogen gas in a nitrogen container increases such as upon exposure of the hydraulic control unit to a high temperature, the pressure switch is actuated at first to generate the scram signal and a scram valve is opened to supply water at high pressure to control rod drives under the driving force of the nitrogen gas at high pressure to rapidly insert the control element into the reactor and shut down it. If the pressure of the nitrogen gas still increases after the scram, the pressure release valve is opened to release the nitrogen gas at high temperature to the atmosphere. Since the scram is attained before the actuation of the pressure release valve, safety reactor shut down can be attained and the hydraulic control unit can be protected. (Sekiya, K.)

  13. Investigation of axial power gradients near a control rod tip

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-07-15

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

  14. Investigation of axial power gradients near a control rod tip

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  15. 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.

  16. 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.

  17. BWR control rod drive scram pilot valve monitoring system

    International Nuclear Information System (INIS)

    Soden, R.A.; Kelly, V.

    1984-01-01

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

  18. BWR control rod drive scram pilot valve monitoring program

    International Nuclear Information System (INIS)

    Soden, R.A.; Kelly, V.

    1986-01-01

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

  19. 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)

  20. 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

  1. Large-area homogeneous periodic surface structures generated on the surface of sputtered boron carbide thin films by femtosecond laser processing

    Energy Technology Data Exchange (ETDEWEB)

    Serra, R., E-mail: ricardo.serra@dem.uc.pt [SEG-CEMUC, Mechanical Engineering Department, University of Coimbra, Rua Luís Reis Santos, 3030-788 Coimbra (Portugal); Oliveira, V. [ICEMS-Instituto de Ciência e Engenharia de Materiais e Superfícies, Avenida Rovisco Pais no 1, 1049-001 Lisbon (Portugal); Instituto Superior de Engenharia de Lisboa, Avenida Conselheiro Emídio Navarro no 1, 1959-007 Lisbon (Portugal); Oliveira, J.C. [SEG-CEMUC, Mechanical Engineering Department, University of Coimbra, Rua Luís Reis Santos, 3030-788 Coimbra (Portugal); Kubart, T. [The Ångström Laboratory, Solid State Electronics, P.O. Box 534, SE-751 21 Uppsala (Sweden); Vilar, R. [Instituto Superior de Engenharia de Lisboa, Avenida Conselheiro Emídio Navarro no 1, 1959-007 Lisbon (Portugal); Instituto Superior Técnico, Avenida Rovisco Pais no 1, 1049-001 Lisbon (Portugal); Cavaleiro, A. [SEG-CEMUC, Mechanical Engineering Department, University of Coimbra, Rua Luís Reis Santos, 3030-788 Coimbra (Portugal)

    2015-03-15

    Highlights: • Large-area LIPSS were formed by femtosecond laser processing B-C films surface. • The LIPSS spatial period increases with laser fluence (140–200 nm). • Stress-related sinusoidal-like undulations were formed on the B-C films surface. • The undulations amplitude (down to a few nanometres) increases with laser fluence. • Laser radiation absorption increases with surface roughness. - Abstract: Amorphous and crystalline sputtered boron carbide thin films have a very high hardness even surpassing that of bulk crystalline boron carbide (≈41 GPa). However, magnetron sputtered B-C films have high friction coefficients (C.o.F) which limit their industrial application. Nanopatterning of materials surfaces has been proposed as a solution to decrease the C.o.F. The contact area of the nanopatterned surfaces is decreased due to the nanometre size of the asperities which results in a significant reduction of adhesion and friction. In the present work, the surface of amorphous and polycrystalline B-C thin films deposited by magnetron sputtering was nanopatterned using infrared femtosecond laser radiation. Successive parallel laser tracks 10 μm apart were overlapped in order to obtain a processed area of about 3 mm{sup 2}. Sinusoidal-like undulations with the same spatial period as the laser tracks were formed on the surface of the amorphous boron carbide films after laser processing. The undulations amplitude increases with increasing laser fluence. The formation of undulations with a 10 μm period was also observed on the surface of the crystalline boron carbide film processed with a pulse energy of 72 μJ. The amplitude of the undulations is about 10 times higher than in the amorphous films processed at the same pulse energy due to the higher roughness of the films and consequent increase in laser radiation absorption. LIPSS formation on the surface of the films was achieved for the three B-C films under study. However, LIPSS are formed under

  2. Large-area homogeneous periodic surface structures generated on the surface of sputtered boron carbide thin films by femtosecond laser processing

    International Nuclear Information System (INIS)

    Serra, R.; Oliveira, V.; Oliveira, J.C.; Kubart, T.; Vilar, R.; Cavaleiro, A.

    2015-01-01

    Highlights: • Large-area LIPSS were formed by femtosecond laser processing B-C films surface. • The LIPSS spatial period increases with laser fluence (140–200 nm). • Stress-related sinusoidal-like undulations were formed on the B-C films surface. • The undulations amplitude (down to a few nanometres) increases with laser fluence. • Laser radiation absorption increases with surface roughness. - Abstract: Amorphous and crystalline sputtered boron carbide thin films have a very high hardness even surpassing that of bulk crystalline boron carbide (≈41 GPa). However, magnetron sputtered B-C films have high friction coefficients (C.o.F) which limit their industrial application. Nanopatterning of materials surfaces has been proposed as a solution to decrease the C.o.F. The contact area of the nanopatterned surfaces is decreased due to the nanometre size of the asperities which results in a significant reduction of adhesion and friction. In the present work, the surface of amorphous and polycrystalline B-C thin films deposited by magnetron sputtering was nanopatterned using infrared femtosecond laser radiation. Successive parallel laser tracks 10 μm apart were overlapped in order to obtain a processed area of about 3 mm 2 . Sinusoidal-like undulations with the same spatial period as the laser tracks were formed on the surface of the amorphous boron carbide films after laser processing. The undulations amplitude increases with increasing laser fluence. The formation of undulations with a 10 μm period was also observed on the surface of the crystalline boron carbide film processed with a pulse energy of 72 μJ. The amplitude of the undulations is about 10 times higher than in the amorphous films processed at the same pulse energy due to the higher roughness of the films and consequent increase in laser radiation absorption. LIPSS formation on the surface of the films was achieved for the three B-C films under study. However, LIPSS are formed under different

  3. 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

  4. Calculation of drop course of control rod assembly in PWR

    International Nuclear Information System (INIS)

    Zhou Xiaojia; Mao Fei; Min Peng; Lin Shaoxuan

    2013-01-01

    The validation of control rod drop performance is an important part of safety analysis of nuclear power plant. Development of computer code for calculating control rod drop course will be useful for validating and improving the design of control rod drive line. Based on structural features of the drive line, the driving force on moving assembly was analyzed and decomposed, the transient value of each component of the driving force was calculated by choosing either theoretical method or numerical method, and the simulation code for calculating rod cluster control assembly (RCCA) drop course by time step increase was achieved. The analysis results of control rod assembly drop course calculated by theoretical model and numerical method were validated by comparing with RCCA drop test data of Qinshan Phase Ⅱ 600 MW PWR. It is shown that the developed RCCA drop course calculation code is suitable for RCCA in PWR and can correctly simulate the drop course and the stress of RCCA. (authors)

  5. Basic visualization experiments on eutectic reaction of boron carbide and stainless steel under sodium-cooled fast reactor conditions

    International Nuclear Information System (INIS)

    Yamano, Hidemasa; Suzuki, Tohru; Kamiyama, Kenji; Kudo, Isamu

    2016-01-01

    This paper describes basic visualization experiments on eutectic reaction and relocation of boron carbide (B 4 C) and stainless steel (SS) under a high temperature condition exceeding 1500degC as well as the importance of such behaviors in molten core during a core disruptive accident in a Generation-IV sodium-cooled fast reactor (750 MWe class) designed in Japan. At first, a reactivity history was calculated using an exact perturbation calculation tool taking into account expected behaviors. This calculation indicated the importance of a relocation behavior of the B 4 C-SS eutectic because its behavior has a large uncertainty in the reactivity history. To clarify this behavior, basic experiments were carried out by visualizing the reaction of a B 4 C pellet contacted with molten SS in a high temperature-heating furnace. The experiments have shown the eutectic reaction visualization as well as freezing and relocation of the B 4 C-SS eutectic in upper part of the solidified test piece due to the density separation. (author)

  6. Control rod repositioning considerations in core design analysis

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  7. Expert system for control rod programming of boiling water reactors

    International Nuclear Information System (INIS)

    Fukuzaki, T.; Yoshida, K.; Kobayashi, Y.; Matsuura, H.; Hoshi, K.

    1986-01-01

    Control rod programming, one of the main tasks in reactor core management of boiling water reactors (BWRs), can be successfully accomplished by well-experienced engineers. By use of core performance evaluation codes, their knowledge plays the main role in searching through optimal control rod patterns and exposure points for adjusting notch positions and exchanging rod patterns. An expert system has been developed, based on a method of knowledge engineering, to lighten the engineer's load in control rod programming. This system utilizes an inference engine suited for planning/designing problems, and stores the knowledge of well-experienced engineers in its knowledge base. In this report, the inference engine, developed considering the characteristics of the control rod programming, is introduced. Then the constitution and function of the expert system are discussed

  8. 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

  9. 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

  10. 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

  11. 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.)

  12. Nuclear reactor internals and control rod handling device

    International Nuclear Information System (INIS)

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

    1981-01-01

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

  13. Behaviour of a VVER-1000 fuel element with boron carbide/steel absorber tested under severe fuel damage conditions in the CORA facility (Results of experiment CORA-W2)

    International Nuclear Information System (INIS)

    Hagen, S.; Hofmann, P.; Noack, V.; Schanz, G.; Schumacher, G.; Sepold, L.

    1994-10-01

    The 'Severe Fuel Damage' (SFD) experiments of the Kernforschungszentrum Karlsruhe (KfK), Federal Republic of Germany, were carried out in the out-of-pile facility 'CORA' as part of the international Severe Fuel Damage (SFD) research. The experimental program was set up to provide information on the failure mechanisms of Light Water Reactor (LWR) fuel elements in a temperature range from 1200 C to 2000 C and in few cases up to 2400 C. Between 1987 and 1992 a total of 17 CORA experiments with two different bundle configurations, i.e. PWR (Pressurized Water Reactor) and BWR (Boiling Water Reactor) bundles were performed. These assemblies represented 'Western-type' fuel elements with the pertinent materials for fuel, cladding, grid spacer, and absorber rod. At the end of the experimental program two VVER-1000 specific tests were run in the CORA facility with identical objectives but with genuine VVER-type materials. The experiments, designated CORA-W1 and CORA-W2 were conducted on February 18, 1993 and April 21, 1993, respectively. Test bundle CORA-W1 was without absorber material whereas CORA-W2 contained one absorber rod (boron carbide/steel). As in the earlier CORA tests the test bundles were subjected to temperature transients of a slow heatup rate in a steam environment. The transient phases of the tests were initiated with a temperature ramp rate of 1 K/s. With these conditions a so-called small-break LOCA was simulated. The temperature escalation due to the exothermal zircon/niobium-steam reaction started at about 1200 C, leading the bundles to maximum temperatures of approximately 1900 C. The thermal response of bundle CORA-W2 is comparable to that of CORA-W1. In test CORA-W2, however, the temperature front moved faster from the top to the bottom compared to test CORA-W1 [de

  14. Large test rigs verify Clinch River control rod reliability

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  15. Review of control rod calibration methods for irradiated AGRs

    Energy Technology Data Exchange (ETDEWEB)

    Telford, A. R.R.

    1975-10-15

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

  16. 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

  17. Cleaning and excavating tool for control rod canopy seals

    International Nuclear Information System (INIS)

    Kucera, R.A.

    1991-01-01

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

  18. Hydraulic system for the drive of control rod

    International Nuclear Information System (INIS)

    Niwano, Masao.

    1978-01-01

    Purpose: To remove thermal stress and improve safety by utilizing water discharged a driving device as a part of cooling water for the device upon driving of control rods. Constitution: A water drain valve is wholly closed and a flow stabilization valve is supplied with an amount of water necessary for driving control rods. Upon driving one control rod, an amount of water required for the driving is caused to flow to the relivant hydraulic control unit and the flow rate in the stabilization valve is reduced by an amount required for the driving to keep the flow rate constant in the flow control valve. Since Excess water conventionally returned to the pressure vessel is utilized as cooling water for the driving device of control rods, the pressure vessel nozzle can be saved. Accordingly, the thermal stress in the nozzle portion can be removed to significantly improve the safety. (Seki, T.)

  19. Absorbing rods for nuclear fast neutron reactor absorbing assembly

    International Nuclear Information System (INIS)

    Aji, M.; Ballagny, A.; Haze, R.

    1986-01-01

    The invention proposes a neutron absorber rod for neutron absorber assembly of a fast neutron reactor. The assembly comprises a bundle of vertical rods, each one comprising a stack of pellets made of a neutron absorber material contained in a long metallic casing with a certain radial play with regard to this casing; this casing includes traps for splinters from the pellets which may appear during reactor operation, at the level of contact between adjacent pellets. The present invention prevents the casing from rupture involved by the disintegration of the pellets producing pieces of boron carbide of high hardness [fr

  20. Water pressure control device for control rod drive

    International Nuclear Information System (INIS)

    Sato, Hideyuki.

    1981-01-01

    Purpose: To minimize the fluctuations in the reactor water level upon occurrence of abnormality by inputting the level signal of the reactor to an arithmetic unit for controlling the pressure of control rod drive water to thereby enable effective reactor level control. Constitution: Signal from a flow rate transmitter is inputted into an arithmetic unit to perform constant flow rate control upon normal operation. While on the other hand, if abnormality occurs such as feedwater pump trips, the arithmetic unit is switched from the constant flow rate control to the reactor water level control. Reactor water level signal is inputted into the arithmetic unit and the control valve is most suitably controlled, whereby water is fed from CST to the reactor by way of control rod drive water system to secure the reactor water level if feedwater to the reactor is interrupted by loss of coolants on the feedwater system. Since this enables to minimize the fluctuations in the reactor water level upon abnormality, the reactor water level can be controlled most suitably by the reactor water level signal. (Moriyama, K.)

  1. 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.)

  2. Synthesis, characterization and thermoelectric properties of metal borides, boron carbides and carbaborides; Synthese, Charakterisierung und thermoelektrische Eigenschaften ausgewaehlter Metallboride, Borcarbide und Carbaboride

    Energy Technology Data Exchange (ETDEWEB)

    Guersoy, Murat

    2015-07-06

    This work reports on the solid state synthesis and structural and thermoelectrical characterization of hexaborides (CaB{sub 6}, SrB{sub 6}, BaB{sub 6}, EuB{sub 6}), diboride dicarbides (CeB{sub 2}C{sub 2}, LaB{sub 2}C{sub 2}), a carbaboride (NaB{sub 5}C) and composites of boron carbide. The characterization was performed by X-ray diffraction methods and Rietveld refinements based on structure models from literature. Most of the compounds were densified by spark plasma sintering at 100 MPa. As high-temperature thermoelectric properties the Seebeck coefficients, electrical conductivities, thermal diffusivities and heat capacities were measured between room temperature and 1073 K. ZT values as high as 0.5 at 1273 K were obtained for n-type conducting EuB{sub 6}. High-temperature X-ray diffraction also confirmed its thermal stability. The solid solutions Ca{sub x}Sr{sub 1-x}B{sub 6}, Ca{sub x}Ba{sub 1-x}B{sub 6} and Sr{sub x}Ba{sub 1-x}B{sub 6} (x = 0, 0.25, 0.5, 0.75, 1) are also n-type but did not show better ZT values for the ternary compounds compared to the binaries, but for CaB{sub 6} the values of the figure of merit (ca. 0.3 at 1073 K) were significantly increased (ca. 50 %) compared to earlier investigations which is attributed to the densification process. Sodium carbaboride, NaB{sub 5}C, was found to be the first p-type thermoelectric material that crystallizes with the hexaboride-structure type. Seebeck coefficients of ca. 80 μV . K{sup -1} were obtained. Cerium diboride dicarbide, CeB{sub 2}C{sub 2}, and lanthanum diboride dicarbide, LaB{sub 2}C{sub 2}, are metallic. Both compounds were used as model compounds to develop compacting strategies for such layered borides. Densities obtained at 50 MPa were determined to be higher than 90 %. A new synthesis route using single source precursors that contain boron and carbon was developed to open the access to new metal-doped boron carbides. It was possible to obtain boron carbide, but metal-doping could not be

  3. 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

  4. 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

  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. 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

  7. Apparatus for decelerating the dropping speed of a control rod

    International Nuclear Information System (INIS)

    Shirakawa, Toshihisa.

    1975-01-01

    Object: To reduce the dropping speed (i.e. withdrawal) of a control rod of the upward insertion type in a BWR type reactor without reducing the speed of insertion. Structure: A control rod is provided with a flaring lower end so as to constitute a speed limiter which is penerated by vertically extending and upwardly open flow ducts that each have a narrow opening and flare upwardly. Thus, at the time of insertion of the control rod, the resistance offered thereto by the surrounding fluid is reduced to provide increased insertion speed. On the other hand, at the time of withdrawal the resistance offered by the fluid is increased to reduce the dropping speed of the control rod. (Ikeda, J.)

  8. 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)

  9. 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)

  10. Calculation of reactivity of control rods in graphite moderated reactors

    International Nuclear Information System (INIS)

    Nakata, H.

    1978-01-01

    A study about the method of calculation for the reactivity of control rods in graphite-moderated critical assemblies, is presented. The result of theoretical calculation, developed by super celles and Nordheim-Scalettar methods are compared with experimental results for the critical Assembly of General Atomic. The two methods are then applicable to reactivity calculation of the control rods of graphite moderated critical assemblies [pt

  11. 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

  12. 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.)

  13. 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.)

  14. Method of operating control rods for BWR type reactors

    International Nuclear Information System (INIS)

    Shirakawa, Toshihisa.

    1979-01-01

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

  15. Inlet for fuel assembly having finger control rods

    International Nuclear Information System (INIS)

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

    1975-01-01

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

  16. Rod cluster control assemblies and rod cluster control guide tubes: wear and drop time

    International Nuclear Information System (INIS)

    Zbinden, M.

    1997-01-01

    The wear of RCCAs and of RCC guide tubes is due to two quite different mechanisms and the remedies to apply for each case might lead to contradictory solutions: - the impact/sliding wear for the seldom moving RCCAs, namely the shutdown RCCAs, under flow-induced vibrations, - the axial sliding wear for the control rods subjected to the stepping movements ordered by the acting load. In this case the hydraulic sticking forces are those which produce an evolution of the surface states that may increase the drop time. The introduction, an historical survey of the encountered difficulties, is followed by short description of the components and then the paper presents contributions of EDF in the R and D field, which take place in two successive multi-annual projects. Lastly, some information is given about the recent evolutions and new problems as well for impact/sliding wear as for drop time under normal or seismic conditions. (author)

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

    International Nuclear Information System (INIS)

    Smith, J. E.

    1984-01-01

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

  18. 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.)

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

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  20. Automatic positioning control device for automatic control rod exchanger

    International Nuclear Information System (INIS)

    Nasu, Seiji; Sasaki, Masayoshi.

    1982-01-01

    Purpose: To attain accurate positioning for a control rod exchanger. Constitution: The present position for an automatic control rod exchanger is detected by a synchro generator. An aimed stopping position for the exchanger, a stop instruction range depending on the distantial operation delay in the control system and the inertia-running distance of the mechanical system, and a coincidence confirmation range depending on the required positioning accuracy are previously set. If there is a difference between the present position and the aimed stopping position, the automatic exchanger is caused to run toward the aimed stopping position. A stop instruction is generated upon arrival at the position within said stop instruction range, and a coincidence confirmation signal is generated upon arrival at the position within the coincidence confirmation range. Since uncertain factors such as operation delay in the control system and the inertia-running distance of the mechanical system that influence the positioning accuracy are made definite by the method of actual measurement or the like and the stop instruction range and the coincidence confirmation range are set based on the measured data, the accuracy for the positioning can be improved. (Ikeda, J.)

  1. 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

  2. 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

  3. Micromechanical analysis of a hybrid composite—effect of boron carbide particles on the elastic properties of basalt fiber reinforced polymer composite

    Science.gov (United States)

    Krishna Golla, Sai; Prasanthi, P.

    2016-11-01

    A fiber reinforced polymer (FRP) composite is an important material for structural application. The diversified application of FRP composites has become the center of attention for interdisciplinary research. However, improvements in the mechanical properties of this class of materials are still under research for different applications. The reinforcement of inorganic particles in a composite improves its structural properties due to their high stiffness. The present research work is focused on the prediction of the mechanical properties of the hybrid composites where continuous fibers are reinforced in a micro boron carbide particle mixed polypropylene matrix. The effectiveness of the addition of 30 wt. % of boron carbide (B4C) particle contributions regarding the longitudinal and transverse properties of the basalt fiber reinforced polymer composite at various fiber volume fractions is examined by finite element analysis (FEA). The experimental approach is the best way to determine the properties of the composite but it is expensive and time-consuming. Therefore, the finite element method (FEM) and analytical methods are the viable methods for the determination of the composite properties. The FEM results were obtained by adopting a micromechanics approach with the support of FEM. Assuming a uniform distribution of reinforcement and considering one unit-cell of the whole array, the properties of the composite materials are determined. The predicted elastic properties from FEA are compared with the analytical results. The results suggest that B4C particles are a good reinforcement for the enhancement of the transverse properties of basalt fiber reinforced polypropylene.

  4. Knowledge based system for control rod programming of BWRs

    International Nuclear Information System (INIS)

    Fukuzaki, Takaharu; Yoshida, Ken-ichi; Kobayashi, Yasuhiro

    1988-01-01

    A knowledge based system has been developed to support designers in control rod programming of BWRs. The programming searches through optimal control rod patterns to realize safe and effective burning of nuclear fuel. Knowledge of experienced designers plays the main role in minimizing the number of calculations by the core performance evaluation code. This code predicts power distibution and thermal margins of the nuclear fuel. This knowledge is transformed into 'if-then' type rules and subroutines, and is stored in a knowledge base of the knowledge based system. The system consists of working area, an inference engine and the knowledge base. The inference engine can detect those data which have to be regenerated, call those subroutine which control the user's interface and numerical computations, and store competitive sets of data in different parts of the working area. Using this system, control rod programming of a BWR plant was traced with about 500 rules and 150 subroutines. Both the generation of control rod patterns for the first calculation of the code and the modification of a control rod pattern to reflect the calculation were completed more effectively than in a conventional method. (author)

  5. 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)

  6. Method for installing a control rod driving device in a reactor

    International Nuclear Information System (INIS)

    Sato, Haruo; Watanabe, Masatoshi.

    1975-01-01

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

  7. Method of controlling a control rod drive exchange apparatus

    International Nuclear Information System (INIS)

    Kase, Keiichi; Yamazaki, Kanji; Hirano, Shigeo; Takeda, Hiroyuki; Oowada, Masataka.

    1981-01-01

    Purpose: To move the mountings means for control rod drives to an aimed position easily in a short time by alternately rotating a rotational moving means and radially moving a lateral transfer means. Constitution: Positions for a rotational moving vehicle and a lateral moving vehicle are inspected respectively by synchro generators A and B. The positional signals detected by the synchro generator A is transformed into an angle by a transducer C and the positional signals detected by the synchro generator B is transformed into a radial distance by a transducer D, whereafter each of the data is transmitted to a computer. The computer controls motors to operate the rotational moving means and the lateral moving means alternately. (Seki, T.)

  8. Seismic analysis of control and safety rod drive mechanism

    International Nuclear Information System (INIS)

    Meher Prasad, A.; Jaya, K.P.; Chellapandi, P.; Rajan Babu, V.; Selvaraj, T.

    2003-01-01

    Control rod and its driving mechanism for a Fast Breeder Reactor is to facilitate safe shutdown of the reactor in case of emergency. A theoretical study on the seismic qualification of control and safety rod driving mechanism is carried out. Earthquake excitations under Operational Basis (ORE) and Safe Shutdown condition (SSE) are considered. The time required for the control rod to reach the bottom position in order to shut down the reaction under excited condition is traced out. The maximum displaced positions and extreme stresses in various parts of the system under excitations are evaluated. The system modeled using beam elements. The connections between different parts are modeled through rigid elements. The interaction between various parts are modeled using GAP elements. (author)

  9. PWR control rods wear by vibrations induced by coolant fluid

    International Nuclear Information System (INIS)

    Reynier, R.

    1997-01-01

    Flow induced vibrations in pressurised water reactors generate the wear of control rods against their guidance systems. Alternate sliding (at 320 deg. C in water) and impact-sliding tests (at room temperature in air) were carried out on 304 L austenitic stainless steel control rods' claddings. Microstructural analysis were made on the wear scars of the tube specimen using Scanning ELectron Microscopy, microhardness measurements and X-ray diffractometry. The alternate sliding leads to an important mass loss, a strong plastic deformation due to the strain hardening of the surface layers and generates strong compressive residual stresses. These results are specific to a severe wear case. Therefore, the impact-sliding mode induces martensitic phase, a cracked oxide layer and a compressive residual stresses weaker than those created in the alternate sliding case. This type of motion leads to a milder wear of the control rods

  10. Design of controller for control rod of research reactors

    International Nuclear Information System (INIS)

    Abou-Zaid, R.M.F.M

    2008-01-01

    Designing and testing digital control system for any nuclear research reactor can be costly and time consuming. In this thesis, a rapid, low-cost proto typing and testing procedure for digital controller design is proposed using the concept of Hardware-In-The-Loop (HIL). Some of the control loop components are real hardware components and the others are simulated. First, the whole system is modeled and tested by Real-Time Simulation (RTS) using conventional simulation techniques such as MATLAB / SIMULINK. Second the Hardware-in-the-loop simulation is tested using Real-Time Windows Target in MATLAB and Visual C ++ . The control parts are included as hardware components which are the reactor control rod and its drivers. Three kinds of controllers are studied, Proportional-Derivative (PD), Proportional-Integral-Derivative (PID) and Fuzzy controller. An experimental setup for the hardware used in HIL concept for the control of the nuclear research reactor has been realized. Experimental results are obtained and compared with the simulation results. The experimental results indicate the validation of HIL method in this domain.

  11. Densification of boron carbide at relatively low temperatures by hot pressing and hot isostatic pressing

    International Nuclear Information System (INIS)

    Telle, R.

    1988-01-01

    The poor sinterability of B 4 C limits its widespread application because both high temperatures and high pressures are required for a complete densification. Moreover, B 4 C suffers from a low strength and fracture toughness, possesses, however, a high potential because of its extreme hardness. Reaction hot pressing of B 4 C-WC-TiC-Si-Co mixtures resulting in B 4 C-TiB 2 -W 2 B 5 composites of high density exhibit remarkable mechanical properties. The influence of hot isostatic pressing (HIP) on the microstructure and the mechanical properties is investigated in cooperation with participants of the COST 503 activities and related to the strengthening and toughening mechanisms. Difficulties during densification by HIP arise from the evaporation of adsorbed volatiles as well as from the strong swelling of the powder compact due to the sintering reaction. Several HIP cycle designs were tested in order to prevent the bloating of the capsule and to control internal stresses due to the misfit of the thermal expansion of the entire phases. In comparison to single phase B 4 C ceramics, bending strength was improved to 1030 MPa, K Ic to 5.2 MPa/m, while hardness was comparable with HV1=38 GPa. Wear test were performed and related to the toughening mechanisms. (orig.) With 56 refs., 9 tabs., 64 figs

  12. 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)

  13. Evaluation of control rod motion simulator research reactors

    International Nuclear Information System (INIS)

    Sanda

    2010-01-01

    Motion simulator has been carried out testing of the reactor control rod using a servomotor. Reactor control rod motion at any point should be in the right position, one of the motors that can move in a precise and correct the servo motor. To ensure that the servo motor to move in accordance with the desired program, then the servomotor function test for motor work to ensure the performance of the appliance. 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 Vout nets at 24 V, 6.5 A with 12 Ω 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 Vout nets at 12V, 4.2 A with a 6 Ω 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 Vout 24 V, 4.5 A with 12 Ω 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)

  14. BWR ATWS mitigation by Fine Motion Control Rod

    International Nuclear Information System (INIS)

    Rohatgi, U.S.; Cheng, H.S.; Khan, H.; Mallen, A.; Diamond, D.

    1994-01-01

    Two main methods of ATWS mitigation in a SBWR are: fine Motion control Rods (FMCRD) and Boron injection via the Standby Liquid control System (SLCS). This study has demonstrated that the use of FMCRD along with feedwater runback mitigated the conditions due to reactivity insertion and possible ATWS in a BWR which is similar to SBWR

  15. Estimation and control in HTGR fuel rod fabrication

    International Nuclear Information System (INIS)

    Downing, D.J.; Bailey, J.M.

    1980-01-01

    A control algorithm has been derived for an HTGR Fuel Rod Fabrication Process utilizing the method of G.E.P. Box and G.M. Jenkins. The estimator is a Kalman filter and is compared with a Least Square estimator and a standard control chart. The effects of system delays are presented. 1 ref

  16. Characterisation of reactor control rod drives. Specification 1-6

    International Nuclear Information System (INIS)

    1975-03-01

    The committee 'Kernreaktorregelung' of VDI/VDE-Gesellschaft Mess- und Regelungstechnik has developed 6 specifications (Typenblaetter) of reactor control rod drives. The specifications are aimed at giving engineers in reactor control systems an outline concerning the function as well as some construction characteristics. (orig./LN) [de

  17. Estimation and control in HTGR fuel rod fabrication

    International Nuclear Information System (INIS)

    Downing, D.J.; Bailey, M.J.

    1980-01-01

    A control algorithm has been derived for a HTGR Fuel Rod Fabrication Process utilizing the method of Box and Jenkins. The estimator is a Kalman filter and is compared with a Least Square estimator and a standard control chart. The effects of system delays are presented

  18. A digital position-indication system for control rods

    International Nuclear Information System (INIS)

    Nishizawa, Yukio; Hayakawa, Toshifumi

    1979-01-01

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

  19. A mathematical method for boiling water reactor control rod programming

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  20. Energy analysis of control rod drive mechanism in HTR-10

    International Nuclear Information System (INIS)

    Bo Hanliang; Wu Yuanqiang

    2000-01-01

    This paper presents a theoretical model for the control rod drive mechanism for the 10 MW High Temperature Gas Cooled Reactor (HTR-10) and analyzes accidents which may occur in the drive mechanism, for example, chain break, coupling damage and other damage scenarios. The results show that the matching problem between buffer capability and coupling strength is the main reason for coupling damage; increased temperatures would reduce eddy damping and cause a mismatch between buffer capability and coupling strength; and the displacement of the buffer spring will affect the coupling force. The results provide a theoretical basis for the design of the control rod drive mechanism for HTR-10

  1. 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

  2. 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.

  3. Calculation of control rods in rectangular reactor, and applications (1960)

    International Nuclear Information System (INIS)

    Goshen, S.; Pazy, A.

    1960-01-01

    The aim of this report is to find a method for estimating the anti-reactivity of control rods perpendicular to the axis in a cylindrical pile. The paper is divided into two parts. In the first is given a method of calculating control rods in a rectangular pile, similar to the Nordheim-Scalettar method for cylindrical piles. As an example the formulas are given for the theories of one and two neutron groups, the generalisation for several groups being evident. In the second part we find by a variation method a formula for estimating the Laplacian of a pile, which may be divided into parallelepipeds for which the Laplacian are given. Finally, this formula is used to calculate the anti-reactivity of rods perpendicular to the axis in a cylindrical pile. (author) [fr

  4. Control device for handling device of control rod drive

    International Nuclear Information System (INIS)

    Sasaki, Toshiya

    1998-01-01

    A predetermined aimed portion of control rod drives disposed in a pedestal is photographed, and image data and camera data including the position of the camera are outputted. Edge cut out processing image data are formed based on the outputted image data, and aimed image data and aimed camera data obtained when previously positioning the handling device precisely to a predetermined aimed position are stored. The aimed image data are taken out from the aimed image data file to prepare computer graphic image data, and there is disposed an image superposing processing portion for comparing images based on the computer graphic image data and images based on the image data for edge cut out processing, as well as comparing the aimed camera data and the camera data, and displaying each of them to an image display portion. (I.S.)

  5. Study of the processes of changing the crystal structure of boron carbide after the destruction of a nuclear reactor as a result of earthquake

    International Nuclear Information System (INIS)

    Mammadov, Kh.; Mirzayev, M.; Garibov, R.; Allahverdiyev, G.

    2017-01-01

    The territories of the Trans Caucasian Republics are characterized by high seismic activity. Therefore, the occurrence of cases of anthropogenic catastrophe including in the territories of nuclear reactors is not ruled out in case of natural disasters. Studies to create detectors based on B 4 C for recording ''cold'', ''hot'' and ''fast'' neutrons in order to increase the safety of nuclear reactors have been carried out in recent years. The B 4 C crystal structure is highly stable at relatively large intervals of temperature and pressure. The study of the thermo physical properties of samples of boron carbide irradiated with ionizing beams is interesting from the point of view to study of the stability of the structure and the stability of this compound. The thermal properties of B 4 C irradiated with ionizing γ radiation from a 60''Co source were investigated using the differential scanning calorimetric (DSC) methods. Upon irradiation with ionizing γ rays, a discrete change in the energy of the atoms occurs at the sites of the crystal lattice, the formation of active centers (radicals, ions, electrons), defects in the crystal lattice, the evaporation of crystalline hydrates, which are present in small amounts in bulk and crystalline compounds. The melting point of B 2 O 3 is 723 K, for boron 2348 K, for B 4 C 2623 K. The melting enthalpy for B 2 O 3 is 24.6 kJ/mol. With increasing temperature, the heat capacity and entropy of the non irradiated and irradiated B 4 C samples are increased. The nature of the change in the enthalpy and the Gibbs potential with increasing temperature depends on the presence of oxygen upon irradiation and during thermogravimetric analysis in the temperature range 298-1300 K. Changes in the values of thermodynamic functions occur due to the formation (under the influence of ionizing radiation) of excited atoms, active centers, defects in the crystal structure of

  6. The influence of hydrogen on the chemical, mechanical, optical/electronic, and electrical transport properties of amorphous hydrogenated boron carbide

    International Nuclear Information System (INIS)

    Nordell, Bradley J.; Karki, Sudarshan; Nguyen, Thuong D.; Rulis, Paul; Caruso, A. N.; Paquette, Michelle M.; Purohit, Sudhaunshu S.; Li, Han; King, Sean W.; Dutta, Dhanadeep; Gidley, David; Lanford, William A.

    2015-01-01

    Because of its high electrical resistivity, low dielectric constant (κ), high thermal neutron capture cross section, and robust chemical, thermal, and mechanical properties, amorphous hydrogenated boron carbide (a-B x C:H y ) has garnered interest as a material for low-κ dielectric and solid-state neutron detection applications. Herein, we investigate the relationships between chemical structure (atomic concentration B, C, H, and O), physical/mechanical properties (density, porosity, hardness, and Young's modulus), electronic structure [band gap, Urbach energy (E U ), and Tauc parameter (B 1/2 )], optical/dielectric properties (frequency-dependent dielectric constant), and electrical transport properties (resistivity and leakage current) through the analysis of a large series of a-B x C:H y thin films grown by plasma-enhanced chemical vapor deposition from ortho-carborane. The resulting films exhibit a wide range of properties including H concentration from 10% to 45%, density from 0.9 to 2.3 g/cm 3 , Young's modulus from 10 to 340 GPa, band gap from 1.7 to 3.8 eV, Urbach energy from 0.1 to 0.7 eV, dielectric constant from 3.1 to 7.6, and electrical resistivity from 10 10 to 10 15 Ω cm. Hydrogen concentration is found to correlate directly with thin-film density, and both are used to map and explain the other material properties. Hardness and Young's modulus exhibit a direct power law relationship with density above ∼1.3 g/cm 3 (or below ∼35% H), below which they plateau, providing evidence for a rigidity percolation threshold. An increase in band gap and decrease in dielectric constant with increasing H concentration are explained by a decrease in network connectivity as well as mass/electron density. An increase in disorder, as measured by the parameters E U and B 1/2 , with increasing H concentration is explained by the release of strain in the network and associated decrease in structural disorder. All of these correlations in a

  7. Control rod for the operation of nuclear reactor

    International Nuclear Information System (INIS)

    Ishida, Hiromi

    1987-01-01

    Purpose: To conduct spectrum shift operation without complicating the reactor core structures, reducing the probability of failures. Constitution: An operation control rod which is driven while passed vertically in the reactor core comprises a strong absorption portion, moderation portion and weak moderation portion defined orderly from above to below and the length for each of the portions is greater than the effective reactor core height. If the operation control rod is lifted to the maximum limit in the upward direction of the reactor core, the weak moderation portion is corresponded over the effective length of the reactor core. Since the weak moderation portion is filled with zirconium and moderators are not present in the operation control rod, water draining gap is formed, neutron spectral shift is formed, excess reactivity is suppressed, absorption of neutrons to fuel fertile material is increased and the formation of nuclear fission material is increased. From the middle to the final stage of the cycle, the control rod is lowered, by which the moderator/fuel effective volume ratio is increased to increase the reactivity. (Kamimura, M.)

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

    International Nuclear Information System (INIS)

    West, G.

    1982-11-01

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

  9. Fuel and control rod failure behavior during degraded core accidents

    International Nuclear Information System (INIS)

    Chung, K.S.

    1984-01-01

    As a part of the pretest and posttest analyses of Light Water Reactor Source Term Experiments (STEP) which are conducted in the Transient Reactor Test (TREAT) facility, this paper investigates the thermodynamic and material behaviors of nuclear fuel pins and control rods during severe core degradation accidents. A series of four STEP tests are being performed to simulate the characteristics of the power reactor accidents and investigate the behavior of fission product release during these accidents. To determine the release rate of the fission products from the fuel pins and the control rod materials, information concerning the timing of the clad failure and the thermodynamic conditions of the fuel pins and control rods are needed to be evaluated. Because the phase change involves a large latent heat and volume expansion, and the phase change is a direct cause of the clad failure, the understanding of the phase change phenomena, particularly information regarding how much of the fuel pin and control rod materials are melted are very important. A simple energy balance model is developed to calculate the temperature profile and melt front in various heat transfer media considering the effects of natural convection phenomena on the melting and freezing front behavior

  10. Self-actuation type electromagnet for control rod retaining mechanism

    International Nuclear Information System (INIS)

    Saito, Makoto; Gunji, Minoru.

    1989-01-01

    The present invention concerns a self-actuation type electromagnet for automatically inserting a control rod into a reactor core for the reactor scram upon occurrence of abnormality in FBR type reactors, etc. That is, a mechanism for preventing scorching is disposed to an attracting portion of a split type core thereby enabling reliable detachment of a control rod. For this purpose, less scorching material is embedded to the attracting portion between each of the core portions, with the surface being slightly protruded. In such an attracting portion, a fine gap is formed between each of the core portions by the contact of less scorching materials. Further, the scorching material is embedded into a metal ring, which is screw-coupled to one of the core portions such that the position is adjustable in the direction of the control rod. Further, the less scorching material is made of alumina. As a result, the attracting portion is neither scorched or fused even when it is used for a long period of time in liquid sodium at high temperature. Therefore, when the electromagnet loses the attracting force, the control rod drops surely. (I.S.)

  11. 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

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

    International Nuclear Information System (INIS)

    Leme, Francisco Louzano; Mattar Neto, Miguel

    2002-01-01

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

  13. 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

  14. Device for rearranging control rods of experimental reactors

    International Nuclear Information System (INIS)

    Louda, J.

    1975-01-01

    The invention claims a means for the adjustment of control rods in experimental reactors with a continuously variable pitch of the fuel element spacer. The proposed device permits obtaining maximum variability in the physical modelling of nuclear power reactor cores in experimental reactors. (F.M.)

  15. 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

  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. Nanomechanical and in situ TEM characterization of boron carbide thin films on helium implanted substrates: Delamination, real-time cracking and substrate buckling

    Energy Technology Data Exchange (ETDEWEB)

    Framil Carpeño, David, E-mail: david.framil-carpeno@auckland.ac.nz [Department of Chemical and Materials Engineering, The University of Auckland, 20 Symonds Street, Auckland 1010 (New Zealand); Ohmura, Takahito; Zhang, Ling [Strength Design Group, Structural Materials Unit, National Institute for Materials Science, 1-2-1 Sengen, Tsukuba, Ibaraki 305-0047 (Japan); Leveneur, Jérôme [National Isotope Centre, GNS Science, 30 Gracefield Road, Gracefield, Lower Hutt 5010 (New Zealand); Dickinson, Michelle [Department of Chemical and Materials Engineering, The University of Auckland, 20 Symonds Street, Auckland 1010 (New Zealand); Seal, Christopher [International Centre for Advanced Materials, The University of Manchester, Oxford Road, Manchester M13 9PL (United Kingdom); Kennedy, John [National Isotope Centre, GNS Science, 30 Gracefield Road, Gracefield, Lower Hutt 5010 (New Zealand); Hyland, Margaret [Department of Chemical and Materials Engineering, The University of Auckland, 20 Symonds Street, Auckland 1010 (New Zealand)

    2015-07-15

    Boron carbide coatings deposited on helium-implanted and unimplanted Inconel 600 were characterized using a combination of nanoindentation and transmission electron microscopy. Real-time coating, cracking and formation of slip bands were recorded using in situ TEM-nanoindentation, allowing site specific events to be correlated with specific features in their load–displacement curves. Cross-sections through the residual indent impression showed a correlation between pop-outs in the load–displacement curves and coating delamination, which was confirmed with cyclic indentation experiments. Inconel exhibits (-11-1) and (1-1-1) twin variants in its deformed region beneath the indenter, organized in bands with a ladder-like arrangement. The nanomechanical properties of the metal–ceramic coating combinations exhibit a marked substrate effect as a consequence of helium implantation.

  18. Nanomechanical and in situ TEM characterization of boron carbide thin films on helium implanted substrates: Delamination, real-time cracking and substrate buckling

    International Nuclear Information System (INIS)

    Framil Carpeño, David; Ohmura, Takahito; Zhang, Ling; Leveneur, Jérôme; Dickinson, Michelle; Seal, Christopher; Kennedy, John; Hyland, Margaret

    2015-01-01

    Boron carbide coatings deposited on helium-implanted and unimplanted Inconel 600 were characterized using a combination of nanoindentation and transmission electron microscopy. Real-time coating, cracking and formation of slip bands were recorded using in situ TEM-nanoindentation, allowing site specific events to be correlated with specific features in their load–displacement curves. Cross-sections through the residual indent impression showed a correlation between pop-outs in the load–displacement curves and coating delamination, which was confirmed with cyclic indentation experiments. Inconel exhibits (-11-1) and (1-1-1) twin variants in its deformed region beneath the indenter, organized in bands with a ladder-like arrangement. The nanomechanical properties of the metal–ceramic coating combinations exhibit a marked substrate effect as a consequence of helium implantation

  19. Creep behavior and wear resistance of Al 5083 based hybrid composites reinforced with carbon nanotubes (CNTs) and boron carbide (B{sub 4}C)

    Energy Technology Data Exchange (ETDEWEB)

    Alizadeh, Ali [Faculty of Materials & Manufacturing Processes, Malek-e-Ashtar University of Technology, Tehran (Iran, Islamic Republic of); Abdollahi, Alireza, E-mail: alirezaabdollahi1366@gmail.com [Faculty of Materials & Manufacturing Processes, Malek-e-Ashtar University of Technology, Tehran (Iran, Islamic Republic of); Biukani, Hootan [Faculty of Engineering, South Tehran Branch, Islamic Azad University, Tehran (Iran, Islamic Republic of)

    2015-11-25

    In the current research, aluminum based hybrid composite reinforced with boron carbide (B{sub 4}C) and carbon nanotubes (CNTs) was produced by powder metallurgy method. creep behavior, wear resistance, surface roughness, and hardness of the samples were investigated. To prepare the samples, Al 5083 powder was milled with boron carbide particles and carbon nanotubes using planetary ball mill under argon atmosphere with ball-to-powder weight ratio of 10:1 for 5 h. Afterwards, the milled powders were formed by hot press process at 380{sup °}C and then were sintered at 585{sup °}C under argon atmosphere for 2 h. There was shown to be an increase in hardness values of composite with an increase in B{sub 4}C content. The micrograph of worn surfaces indicate a delamination mechanism due to the presence of CNTs and abrasion mechanism in composite containing 10 vol.%B{sub 4}C. Moreover, it was shown that increasing B{sub 4}C content increases the wear resistance by 3 times under a load of 20 N and 10 times under a load of 10 N compared to CNTs-reinforced composite. surface roughness of the composite containing 5 vol.%CNT has shown to be more than other samples. The results of creep test showed that adding carbon nanotubes increases creep rate of Al 5083 alloy; however, adding B{sub 4}C decreases its creep rate. - Highlights: • Al 5083/(CNTs + B{sub 4}C) hybrid composite was produced by powder metallurgy method. • Creep behavior, wear resistance, surface roughness, and Hardness of samples were investigated. • Addition of CNTs to Al 5083 matrix reduces alloy hardness, wear resistance and creep strength. • By addition of B{sub 4}C and composite hybridization, creep strength and wear resistance increased. • Surface roughness of Al-5 vol.%CNT has shown to be more than other samples.

  20. Control rod blow out protection system

    International Nuclear Information System (INIS)

    Dietrich, J.R.; Flinn, W.S.; Groves, M.D.

    1976-01-01

    A control system is described which is comprised of a plurality of low worth absorber elements with individual hydraulic actuator assemblies, positioned within the reactor vessel. Axial distortions and safety hazards are minimized by this arrangement. (E.C.B.)

  1. Urea controlled hydrothermal synthesis of ammonium aluminum carbonate hydroxide rods

    Science.gov (United States)

    Wang, Fang; Zhu, Jianfeng; Liu, Hui

    2018-03-01

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

  2. Renovating process for Pressurized Water Reactor control rod assemblies and corresponding control

    International Nuclear Information System (INIS)

    Jahnke, S.; Ple, P.

    1989-01-01

    In the first PWRs the control rods are moving by the intermediary of electromagnetic mechanisms where the power fed to the electromagnets is selected by a hard wired logic circuit connected to the controldesh by another logic control. For renovating the control rod assemblies each power assembly is replaced by an electronic assembly containing an ordinator and power supply interfaces [fr

  3. 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

  4. Diagnostic device for failures in control rod drives

    International Nuclear Information System (INIS)

    Okutani, Tetsuro.

    1982-01-01

    Purpose: To enable to concretely point out a failure position when a failure might occur by diagnosing the failure without affecting the variation to the state of a reactor core. Constitution: A frequency switching circuit is provided in an inverter for controlling the rotating speed of a motor for discharging and charging a control rod. Then, a voltage detector is provided at asemiconductor switch provided between the inverter and the motor. When a high frequency control signal is input to the inverter in diagnosing a failure, the switching speed of the switch is accelerated, a current hardly flows through the motor, and even if the inverter is operated, the motor will not rotate. Thus, the failure of a control rod drive can be diagnosed without affecting any influence to the state of a reactor core. (Kamimura, M.)

  5. 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

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

    International Nuclear Information System (INIS)

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

    2008-01-01

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

  7. Production control in HTGR fuel rod fabrication

    International Nuclear Information System (INIS)

    Downing, D.J.; Bailey, M.

    1979-06-01

    Purpose of this report was (1) to determine which techniques (Kalman Filter, weighted least squares, Shewhart control chart) are capable of detecting drift or step changes earliest in a manufacturing process, and (2) what method would work well in maintaining the manufacturing process at an acceptable level of quality. To solve part (1) simulation studies were performed for various test cases of interest. No single technique was superior in all of these cases, but the Kalman Filter appeared to be more robust to various process changes. The weighted least squares did a good job when the weight was near unity (0.9977) and failed when the weight was small (0.63). The Shewhart control chart is better for detecting step changes than for trends. Several methods wre compared to try to answer part (2). In this report the model building and forecasting was done using the methods of Box and Jenkins

  8. 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.

  9. Drive mechanism nuclear reactor control rod

    International Nuclear Information System (INIS)

    Brooks, J.G. Jr.; Maure, D.R.; Meijer, C.H.

    1978-01-01

    An improved method and apparatus for operating magnetic stepping-type mechanisms. The current flowing in the coils of magnetic stepping-type mechanisms of the kind, for instance, that are used in control-element drive mechanisms is sensed and used to monitor operation of the mechanism. Current waveforms that characterize the motion of the mechanism are used to trigger changes in drive voltage and to verify that the drive mechanism is operating properly. In addition, incipient failures are detected through the observation of differences between the observed waveform and waveforms that characterize proper operation

  10. Releasing method of connection of control rod and its drive mechanism in a reactor

    International Nuclear Information System (INIS)

    Ishida, Kazuo; Futatsugi, Masao.

    1976-01-01

    Object: To disengage a control rod from a control rod drive device in a boiling water reactor with a minimal failure of the device, when connection there between cannot be released in a normal manner. Structure: First, a part of a piston tube in the control rod drive device is withdrawn externally of a control rod housing and cut. Next, a discharge tool, which is designed to be connected with the cut piston tube, is connected to the remainder of the piston tube within the housing and the aforesaid piston tube is pushed into the index tube. The index tube is then cut by the discharge tool. Thus, the control rod drive device and the control rod may be separated. Thereafter, the control rod may be removed from the top of the reactor container whereas the control rod drive device removed from the bottom thereof. (Ikeda, J.)

  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. Device for discharging drain in a control rod driving apparatus

    International Nuclear Information System (INIS)

    Ikeda, Tadasu; Ikuta, Takuzo; Yoshida, Tomiji; Tsukahara, Katsumi.

    1975-01-01

    Object: To efficiently and safely collect and discharge drain by a simple construction in which a drain cover and a drain tank in a control rod driving apparatus are integrally formed, and an overhauling wrench of said apparatus and a drain hose are mounted on the drain tank. Structure: When a mounting bolt is untightened by a torque wrench so as to be removed from a flange surface of the control rod driving apparatus in a nuclear reactor, axial movement of said apparatus is absorbed by a spring so that drain containing a radioactive material is discharged into a drain tank through the flange surface of said apparatus and is then guided into a collecting tank through a drain hose. (Kamimura, M.)

  13. Toward an early detection of PWR control rod anomalous dropping

    International Nuclear Information System (INIS)

    Blazquez, J.; Vallejo, I.

    1998-01-01

    Some anomalous PWR control rods dropping occurred in the past. It is assumed to be caused by a geometrical deformation of its guide tube, which might be related with neutron fluence and its sharp changes. Now at days, this problem is an open field of research, oriented to the understanding and prevention of the event. Work here is focused toward early detection. A differential equation modelling control rod free fall movement is found. There result three acceleration terms: gravity; friction with fluid; and friction with its guide tube. From recorded Plant measurements, both friction coefficients are estimated. The one from guide tube experiences a large variation in case of anomalous dropping; so relationship with neutron fluence is proposed for the prevention purpose. (Author)

  14. 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.

  15. Hot rolled composite billet for nuclear control rods

    International Nuclear Information System (INIS)

    Miller, G.E.

    1976-01-01

    This invention relates to a composite plate shaped billet, useful in the fabrication of nuclear control rods, which comprises a core of stainless steel containing about 2 percent boron 10, a thin coating of zirconia on the surfaces of said core, and said zirconia coating being completely encased in a jacket of mild steel, said composite having been hot rolled between about 1075 0 and about 1165 0 C. 1 claim, 8 figures

  16. Conceptual design of control rod regulating system for plate type fuels of Triga-2000 reactor

    International Nuclear Information System (INIS)

    Eko Priyono; Saminto

    2016-01-01

    Conceptual design of the control rod regulating system for plate type fuel of TRIGA-2000 reactor has been made. Conceptual design of the control rod regulating system for plate type fuel of TRIGA-2000 reactor was made with refer to study result of instrument and control system which is used in BATAN'S reactor. Conceptual design of the control rod regulating system for plate type fuel of TRIGA-2000 reactor consist of 4 segments that is control panel, translator, driver and display. Control panel is used for regulating, safety and display control rod, translator is used for signal processing from control panel, driver is used for driving control rod and display is used for display control rod level position. The translator was designed in 2 modes operation i.e operation by using PLC modules and IC TTL modules. These conceptual design can be used as one of reference of control rod regulating system detail design. (author)

  17. 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.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-06-01

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

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

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  20. 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.

  1. 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)

  2. 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)

  3. 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)

  4. 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)

  5. A device for the hydraulic control of nuclear reactor control rods

    International Nuclear Information System (INIS)

    Frisch, Erling; Frisch, D.R.; Andrews, H.N.

    1974-01-01

    A device for driving and locking the control rods of a nuclear reactor. This device comprises a hydraulic driving piston mounted in a cylinder provided with a construction for absorbing shocks. The piston is provided, at is extremity, with a locking device adapted to engage a stationary lock, it being possible to control the latter for freeing said piston locking device; with such an arrangement, the control rod is normally maintained in position, and it can be freed only by a positive signal. Moreover, the control rod movements are slowed down, so as to prevent the gripping device from being damaged. This device can be used in the nuclear industry [fr

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

    International Nuclear Information System (INIS)

    Luetzow, H.J.

    1981-01-01

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

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

    International Nuclear Information System (INIS)

    Scherer, W.; Neef, H.J.

    1976-07-01

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

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

    International Nuclear Information System (INIS)

    Flessner, H.; Paeserack, U.

    1974-01-01

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

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

    International Nuclear Information System (INIS)

    Damiano, B.

    1989-01-01

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

  10. 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

  11. Theoretical calibration of grey and black control rods of gas-graphite power reactors

    International Nuclear Information System (INIS)

    Joksimovic, V.

    1964-01-01

    Full text: Calculation of calibration curve for particular control rod batches is of significant importance for safety and operation reasons. The procedure presented in this paper is based on the two following criteria: Constants of the lattice region with control rods are determined by supercell method. Effective multiplication constant of the core dependent on the insertion of control rods was determined by dividing the core onto two axial and radial zones. Calculation of the black control rods takes into account epithermal absorption. Thermal extrapolated length of the control rods was calculated by using Kushneriuk-McKey relation. The extrapolated length of the grey rods and the epithermal extrapolated length of the black rods were calculated by diffusion theory. Correlation procedure was used for calculation of epithermal extrapolated length. The complete mathematical procedure was programmed for calculations on the digital ZUSE-Z-23 and ELLIOTT-803-B computers

  12. Top closure for control rod drive for nuclear reactor

    International Nuclear Information System (INIS)

    Raas, J.H.; Schwartz, J.I.

    1978-01-01

    A removable top closure and venting assembly for the tubular housing of a control rod drive includes a mounting ring threadably inserted in the upper end of the housing, a fluid-sealing closure member beneath the mounting ring and which is mounted in and coupled to the mounting ring by means of a ball and socket joint, a gas vent defined by interconnecting passages extending through the closure and through the ball and socket joint, and a vent valve accessible from the top of the closure assembly. 3 claims, 2 figures

  13. Fabrication of control rod system of RSG-GAS

    International Nuclear Information System (INIS)

    Hari-Sudirdjo; Setyono; Hendra-Prasetya

    2003-01-01

    Two unit absorbers, they are part of RSG-GAS control rod system, have been fabricated. One set absorber consist of two absorber plates and absorber casing. Absorber plate is made of Ag In Cd ( 80%, 15%, 5% ) alloy, which is cladded by stainless steel plate SS-316. Ag In Cd absorber plate has size of 625 mm x 60 mm x 3.3 mm, while cladding plat has thickness of 0.8 mm. Fabrication of two set absorbers has been conducted according to the plan

  14. Clinch River Breeder Reactor secondary control rod system

    International Nuclear Information System (INIS)

    McKeehan, E.R.; Sim, R.G.

    1977-01-01

    The shutdown system for the Clinch River Breeder Reactor (CRBR) includes two independent systems--a primary and a secondary system. The Secondary Control Rod System (SCRS) is a new design which is being developed by General Electric to be independent from the primary system in order to improve overall shutdown reliability by eliminating potential common-mode failures. The paper describes the status of the SCRS design and fabrication and testing activities. Design verification testing on the component level is largely complete. These component tests are covered with emphasis on design impact results. A prototype unit has been manufactured and system level tests in sodium have been initiated

  15. 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.

  16. Method for depleting BWRs using optimal control rod patterns

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  17. Stepping movement analysis of control rod drive mechanism

    International Nuclear Information System (INIS)

    Xu Yantao; Zu Hongbiao

    2013-01-01

    Background: Control rod drive mechanism (CRDM) is one of the important safety-related equipment for nuclear power plants. Purpose: The operating parameters of stepping movement, including lifting loads, step distance and step velocity, are all critical design targets. Methods: FEA and numerical simulation are used to analyze stepping movement separately. Results: The motion equations of the movable magnet in stepping movement are established by load analysis. Gravitation, magnetic force, fluid resistance and spring force are all in consideration in the load analysis. The operating parameters of stepping movement are given. Conclusions: The results, including time history curves of force, speed and etc, can positively used in the design of CRDM. (authors)

  18. Control rod drive mechanism stator loss of coolant test

    International Nuclear Information System (INIS)

    Besel, L.; Ibatuan, R.

    1977-04-01

    This report documents the stator loss of coolant test conducted at HEDL on the lead unit Control Rod Drive Mechanism (CRDM) in February, 1977. The purpose of the test was to demonstrate scram capability of the CRDM with an uncooled stator and to obtain a time versus temperature curve of an uncooled stator under power. Brief descriptions of the test, hardware used, and results obtained are presented in the report. The test demonstrated that the CRDM could be successfully scrammed with no anomalies in both the two-phase and three-phase stator winding hold conditions after the respective equilibrium stator temperatures had been obtained with no stator coolant

  19. Method for automatic control rod operation using rule-based control

    International Nuclear Information System (INIS)

    Kinoshita, Mitsuo; Yamada, Naoyuki; Kiguchi, Takashi

    1988-01-01

    An automatic control rod operation method using rule-based control is proposed. Its features are as follows: (1) a production system to recognize plant events, determine control actions and realize fast inference (fast selection of a suitable production rule), (2) use of the fuzzy control technique to determine quantitative control variables. The method's performance was evaluated by simulation tests on automatic control rod operation at a BWR plant start-up. The results were as follows; (1) The performance which is related to stabilization of controlled variables and time required for reactor start-up, was superior to that of other methods such as PID control and program control methods, (2) the process time to select and interpret the suitable production rule, which was the same as required for event recognition or determination of control action, was short (below 1 s) enough for real time control. The results showed that the method is effective for automatic control rod operation. (author)

  20. Managing the aging of BWR control rod drive systems

    International Nuclear Information System (INIS)

    Greene, R.H.; Farmer, W.S.

    1992-01-01

    This Phase I Nuclear Plant Aging Research (NPAR) study examines the aging phenomena associated with BWR control and rod drive mechanisms (CRDMs) and assesses the merits of various methods of ''imaging'' this aging. Information for this study was acquired from (1) the results of a special CRDM aging questionnaire distributed to each US BWR utility, (2) a first-of-its-kind workshop held to discuss CRDM aging and maintenance concerns, (3) an analysis of the Nuclear Plant Reliability Data System (NPRDS) failure cases attributed to the control rod drive (CRD) system, and (4) personal information exchange with nuclear industry CRDM maintenance experts. The report documenting the findings of this research, NUREG-5699, will be published this year. Nearly 23% of the NPRDS CRD system component failure reports were attributed to the CRDM. The CRDM components most often requiring replacement due to aging are the Graphitar seals. The predominant causes of aging for these seals are mechanical wear and thermal embrittlement. More than 59% of the NPRDS CRD system failure reports were attributed to components that comprise the hydraulic control unit (HCU). The predominant HCU components experiencing the effects of service wear and aging are value seals, discs, seats, stems, packing, and diaphragms

  1. Conceptual Design of Bottom-mounted Control Rod Drive Mechanism

    International Nuclear Information System (INIS)

    Lee, Jin Haeng; Kim, Sanghaun; Yoo, Yeonsik; Cho, Yeonggarp; Kim, Dongmin; Kim, Jong In

    2013-01-01

    The arrangement of the BMCRDMs and irradiation holes in the core is therefore easier than that of the top-mounted CRDM. Hence, many foreign research reactors, such as JRR-3M, JMTR, OPAL, and CARR, have adopted the BMCRDM concept. The purpose of this paper is to introduce the basic design concept on the BMCRDM. The major differences of the CRDMs between HANARO and KJRR are compared, and the design features and individual system of the BMCRDM for the KJRR are described. The Control Rod Drive Mechanism (CRDM) is a device to regulate the reactor power by changing the position of a Control Absorber Rod (CAR) and to shut down the reactor by fully inserting the CAR into the core within a specified time. The Bottom-Mounted CRDM (BMCRDM) for the KiJang Research Reactor (KJRR) is a quite different design concept compared to the top-mounted CRDM such as HANARO and JRTR. The main drive mechanism of the BMCRDM is located in a Reactivity Control Mechanism (RCM) room under the reactor pool bottom, which makes the interference with equipment in the reactor pool reduced

  2. Conceptual Design of Bottom-mounted Control Rod Drive Mechanism

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jin Haeng; Kim, Sanghaun; Yoo, Yeonsik; Cho, Yeonggarp; Kim, Dongmin; Kim, Jong In [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-05-15

    The arrangement of the BMCRDMs and irradiation holes in the core is therefore easier than that of the top-mounted CRDM. Hence, many foreign research reactors, such as JRR-3M, JMTR, OPAL, and CARR, have adopted the BMCRDM concept. The purpose of this paper is to introduce the basic design concept on the BMCRDM. The major differences of the CRDMs between HANARO and KJRR are compared, and the design features and individual system of the BMCRDM for the KJRR are described. The Control Rod Drive Mechanism (CRDM) is a device to regulate the reactor power by changing the position of a Control Absorber Rod (CAR) and to shut down the reactor by fully inserting the CAR into the core within a specified time. The Bottom-Mounted CRDM (BMCRDM) for the KiJang Research Reactor (KJRR) is a quite different design concept compared to the top-mounted CRDM such as HANARO and JRTR. The main drive mechanism of the BMCRDM is located in a Reactivity Control Mechanism (RCM) room under the reactor pool bottom, which makes the interference with equipment in the reactor pool reduced.

  3. Standard test methods for chemical, mass spectrometric, and spectrochemical analysis of nuclear-grade aluminum oxide and aluminum oxide-boron carbide composite pellets

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    1994-01-01

    1.1 These test methods cover procedures for the chemical, mass spectrometric, and spectrochemical analysis of nuclear-grade aluminum oxide and aluminum oxide-boron carbide composite pellets to determine compliance with specifications. 1.2 The analytical procedures appear in the following order: Sections Boron by Titrimetry 7 to 13 Separation of Boron for Mass Spectrometry 14 to 19 Isotopic Composition by Mass Spectrometry 20 to 23 Separation of Halides by Pyrohydrolysis 24 to 27 Fluoride by Ion-Selective Electrode 28 to 30 Chloride, Bromide, and Iodide by Amperometric Microtitrimetry 31 to 33 Trace Elements by Emission Spectroscopy 34 to 46 1.3 The values stated in SI units are to be regarded as the standard. 1.4 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use. (F...

  4. A combination method for simulation of secondary knock-on atoms of boron carbide induced by neutron irradiation in SPRR-300

    International Nuclear Information System (INIS)

    Wu, Jian-Chun; Feng, Qi-Jie; Liu, Xian-Kun; Zhan, Chang-Yong; Zou, Yu; Liu, Yao-Guang

    2016-01-01

    A multiscale sequence of simulation should be used to predict properties of materials under irradiation. Binary collision theory and molecular dynamics (MDs) method are commonly used to characterize the displacement cascades induced by neutrons in a material. In order to reduce the clock time spent for the MD simulation of damages induced by high-energy primary knock-on atoms (PKAs), the damage zones were split into sub-cascade according to the sub-cascade formation criteria. Two well-known codes, Geant4 and TRIM, were used to simulate high-energy PKA-induced cascades in B_4C and then produce the secondary knock-on atom (SKA) energy spectrum. It has been found that both high-energy primary knock-on B and C atoms move a long range in the boron carbide. These atoms produce sub-cascades at the tip of trajectory. The energy received by most of the SKAs is <10 keV, which can be used as input to reduce the clock time spent for MD simulation.

  5. 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

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

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  7. Simulating control rod and fuel assembly motion using moving meshes

    Energy Technology Data Exchange (ETDEWEB)

    Gilbert, D. [Department of Electrical and Computer Engineering, McMaster University, 1280 Main Street West, Hamilton Ontario, L8S 4K1 (Canada)], E-mail: gilbertdw1@gmail.com; Roman, J.E. [Departamento de Sistemas Informaticos y Computacion, Universidad Politecnica de Valencia, Camino de Vera s/n, 46022 Valencia (Spain); Garland, Wm. J. [Department of Engineering Physics, McMaster University, 1280 Main Street West, Hamilton Ontario, L8S 4K1 (Canada); Poehlman, W.F.S. [Department of Computing and Software, McMaster University, 1280 Main Street West, Hamilton Ontario, L8S 4K1 (Canada)

    2008-02-15

    A prerequisite for designing a transient simulation experiment which includes the motion of control and fuel assemblies is the careful verification of a steady state model which computes k{sub eff} versus assembly insertion distance. Previous studies in nuclear engineering have usually approached the problem of the motion of control rods with the use of nonlinear nodal models. Nodal methods employ special approximations for the leading and trailing cells of the moving assemblies to avoid the rod cusping problem which results from the naive volume weighted cell cross-section approximation. A prototype framework called the MOOSE has been developed for modeling moving components in the presence of diffusion phenomena. A linear finite difference model is constructed, solutions for which are computed by SLEPc, a high performance parallel eigenvalue solver. Design techniques for the implementation of a patched non-conformal mesh which links groups of sub-meshes that can move relative to one another are presented. The generation of matrices which represent moving meshes which conserve neutron current at their boundaries, and the performance of the framework when applied to model reactivity insertion experiments is also discussed.

  8. Control rod driving mechanism of reactor, control device and operation method therefor

    International Nuclear Information System (INIS)

    Ariyoshi, Masahiko; Matsumoto, Fujio; Matsumoto, Koji; Kinugasa, Kunihiko; Nara, Yoshihiko; Otama, Kiyomaro; Mikami, Takao

    1998-01-01

    The present invention provides a device for and a method of directly driving control rods of an FBR type reactor linearly by a cylinder type linear motor while having a driving shaft as an electric conductor. Namely, a linear induction motor drives a driving shaft connected with a control rod and vertically moving the control rod by electromagnetic force as an electric conductor. The position of the control rod is detected by a position detector. The driving shaft is hung by a wire by way of an electromagnet which is attachably/detachably held. With such a constitution, the driving shaft connected with the control rod can be vertically moved linearly, stopped or kept. Since they can be driven smoothly at a wide range speed, the responsibility and reliability of the reactor operation can be improved. In addition, since responsibility of the control rod operation is high, scram can be conducted by the linear motor. Since the driving mechanism can be simplified, maintenance and inspection operation can be mitigated. (I.S.)

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

    International Nuclear Information System (INIS)

    Mizoo, Nobutatsu; Nakano, Masafumi

    1977-07-01

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

  10. Automatic exchange unit for control rod drive device

    International Nuclear Information System (INIS)

    Nasu, Seiji; Sasaki, Masayoshi.

    1982-01-01

    Purpose: To enable automatic reoperation and continuation without external power interruption remedy device at the time of recovering the interrupted power soruce during automatic positioning operation. Constitution: In case of an automatic exchange unit for a control rod drive device of the control type for setting the deviation between the positioning target position and the present position of the device to zero, the position data of the drive device of the positioning target value of the device is automatically read, and an interlock of operation inhibit is applied to a control system until the data reading is completed and automatic operation start or restart conditions are sequentially confirmed. After the confirmation, the interlock is released to start the automatic operation or reoperation. Accordingly, the automatic operation can be safely restarted and continued. (Yoshihara, H.)

  11. Underwater characterization of control rods for waste disposal using SMOPY

    Energy Technology Data Exchange (ETDEWEB)

    Gallozzi-Ulmann, A.; Couturier, P.; Amgarou, K.; Rothan, D.; Menaa, N. [CANBERRA France,1 rue des Herons, 78182 ST Quentin Yvelines Cedex (France); Chard, P. [CANBERRA UK, Lower Dunbeath House, Forss Business Park, Thurso, Caithness KW14 7UZ (United Kingdom)

    2015-07-01

    Storage of spent fuel assemblies in cooling ponds requires careful control of the geometry and proximity of adjacent assemblies. Measurement of the fuel burnup makes it possible to optimise the storage arrangement of assemblies taking into account the effect of the burnup on the criticality safety margins ('burnup credit'). Canberra has developed a measurement system for underwater measurement of spent fuel assemblies. This system, known as 'SMOPY', performs burnup measurements based on gamma spectroscopy (collimated CZT detector) and neutron counting (fission chamber). The SMOPY system offers a robust and waterproof detection system as well as the needed capability of performing radiometric measurements in the harsh high dose - rate environments of the cooling ponds. The gamma spectroscopy functionality allows powerful characterization measurements to be performed, in addition to burnup measurement. Canberra has recently performed waste characterisation measurements at a Nuclear Power Plant. Waste activity assessment is important to control costs and risks of shipment and storage, to ensure that the activity level remains in the range allowed by the facility, and to declare activity data to authorities. This paper describes the methodology used for the SMOPY measurements and some preliminary results of a radiological characterisation of AIC control rods. After describing the features and normal operation of the SMOPY system, we describe the approach used for establishing an optimum control rod geometric scanning approach (optimum count time and speed) and the method of the gamma spectrometry measurements as well as neutron check measurements used to verify the absence of neutron sources in the waste. We discuss the results obtained including {sup 60}Co, {sup 110m}Ag and {sup 108m}Ag activity profiles (along the length of the control rods) and neutron results including Total Measurement Uncertainty evaluations. Full self-consistency checks were

  12. Scram characteristics of the control rods of a pressurized water reactor under seismic conditions

    International Nuclear Information System (INIS)

    Fujita, Katsuhisa; Shinohara, Yoshikazu; Nakatogawa, Tetsuto; Nanbu, Kiyoshi; Nomura, Tomonori.

    1987-01-01

    Control rod drop verification experiments of a pressurized water reactor under seismic conditions are performed to confirm the insertion function of control rods into a core. To evaluate these tests, computer simulations are performed. A fuel assembly, control rods, guide tube and other associated structures are immersed in a water tank, and shaken by four hydraulic shakers. The scram time of control rods under seismic conditions was measured, and confirmed to meet the scram function. Moreover, vibrational response characteristics of core structures and dropping behavior of control rods in consideration of collisions are calculated by using a finite difference method. The behavior of the dropping control rods and the scram time obtained by the computer simulation show a very good agreement with the verification experimental results. (author)

  13. Analytical estimation of control rod shadowing effect for excess reactivity measurement of HTTR

    International Nuclear Information System (INIS)

    Nakano, Masaaki; Fujimoto, Nozomu; Yamashita, Kiyonobu

    1999-01-01

    The fuel addition method is generally used for the excess reactivity measurement of the initial core. The control rod shadowing effect for the excess reactivity measurement has been estimated analytically for High Temperature Engineering Test Reactor (HTTR). 3-dimensional whole core analyses were carried out. The movements of control rods in measurements were simulated in the calculation. It was made clear that the value of excess reactivity strongly depend on combinations of measuring control rods and compensating control rods. The differences in excess reactivity between combinations come from the control rod shadowing effect. The shadowing effect is reduced by the use of plural number of measuring and compensating control rods to prevent deep insertion of them into the core. The measured excess reactivity in the experiments is, however, smaller than the estimated value with shadowing effect. (author)

  14. Control rod interaction models for use in 2D and 3D reactor geometries

    International Nuclear Information System (INIS)

    Bannerman, R.C.

    1985-11-01

    Control rod interaction models are developed for use in two-dimensional and three-dimensional reactor geometries. These models allow the total worth of any combination of control rods inserted to be determined from the individual worths in conjunction with an algorithm representing interaction effects between them. The validity of the assumptions is demonstrated for fast and thermal systems showing modelling errors of 1#percent# or less in inserted control rod worths. The models are ideally suited for most reactor systems. (UK)

  15. HCPWR type fuel elements design with respect to its control rods

    International Nuclear Information System (INIS)

    Abbate, M.J.; Sbaffoni, M.M.; Patino, N.E.; Torasso, O.

    1992-01-01

    The high conversion reactors (HCPWR) can improve the nuclear fuel utilization. One of its present problems is the optimization of the control rods' worth and its relationship with the void coefficient. This investigation means, starting from one reference's design, to analyze several possibilities on the number and distribution of the control rods. As main result, one design including 24 control rods in an optimized distribution, is recommended. (author)

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

    International Nuclear Information System (INIS)

    Kalcheva, Silva; Koonen, Edgar

    2009-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Norinder, O; Nyman, K

    1962-06-15

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

  18. Computerized supervision and control system for movement at the RP-10 reactor control rods bank

    International Nuclear Information System (INIS)

    Padilla M, C.E.

    1998-01-01

    The project involves the use of a compatible microcomputer, Labwindows/CVI software, as well as National Instruments data acquisition cards AT-MIO16-E10 and PC-DIO96 to modify the sequence of movement of the reactor's rods and control them from a graphic interface in a computer's monitor. This graphic presentation is set as console of virtual instruments from where rod movement can be conducted. Normal rod movement, bank rod movement, and rod calibration have been considered. These experiences involve different logic of rod movements, which will determine movement sequence. Control of the automatic range of a current amplifier module was also considered. This module is know as 'automatic pilot amplifier' and given the strategic location of its detector (compensated ionizing camera) at the reactor's core, it delivers neutron flux current considered as reference to superficial neutron flux distribution at the reactor's core. Lecture and monitoring of this signal allows taking the reactor to a certain power, current of this signal is proportional to the power we want the reactor to reach. Advantages obtained with this system include the update of the control console, more uniform distribution of neutron flux, with lower and uniform burnup of nuclear fuel. (author)

  19. Study on anti-seismic test of control rod driving system suspended by magnetic force

    International Nuclear Information System (INIS)

    Zhang Zhihua; Qian Dazhi; Xu Xianqi; Huang Hongwen; Zhang Zhengming; Wu Xinxin; Hu Xiao

    2012-01-01

    To verify the stability, reliability and security function in extreme conditions, the anti-seismic test of control rod drive line was conducted. Drop-time of control rod drive line in different earthquake intensities was got. The response and strain values of control rod drive line acceleration on SL-1, SL-2 level were measured. Safety functions of control rod drive line were validated in different work conditions. Anti-seismic test data shows that the driving system can keep the structure's integrality and realize operation function under OBE and SSE. (authors)

  20. Dynamic insertion analysis of control rods of BWR under seismic excitation

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  1. The safety feature of hydraulic driving system of control rod for 200 MW nuclear heating reactor

    International Nuclear Information System (INIS)

    Chi Zongbo; Wu Yuanqiang

    1997-01-01

    The hydraulic driving system of control rod is used as control rod drive mechanism in 200 MW nuclear heating reactor. Design of this system is based on passive system, integrating drive and guide of control rod. The author analyzes the inherent safety and the design safety of this system, with mechanism of control rod not ejecting when the pressure of pressure vessel is lost, and calculating result of core not exposing when the amount of coolant is drained by broken pipe. The results indicate that this system has good safety feature, and assures reactor safety under any accident conditions, providing important technology support for 200 MW nuclear heating reactor with inherent safety feature

  2. Preliminary performance test of control rod position indicator for ballscrew type CEDM

    International Nuclear Information System (INIS)

    Yoo, J. Y.; Kim, J. H.; Hu, H.; Lee, J. S.; Kim, J. I.

    2003-01-01

    The reliability and accuracy of the information on control rod position are very important to the reactor safety and the design of the core protection system. A survey on the RSPT(Reed Switch Position Transmitter) type control rod position indication system and its actual implementation in the exiting nuclear power plants in Korea was performed first. The prototype of control rod position indicator having the high performance for the ballscrew type CEDM was developed on the basis of RSPT technology identified through the survey. The characteristics of control rod position indicator was defined and documented through design procedure and preliminary performance test

  3. Establishment of analysis procedure for control rod reactivity worth

    Energy Technology Data Exchange (ETDEWEB)

    Song, Hoon; Kim, Young Il; Kim, Sang Ji; Kim, Young In

    2001-03-01

    As to the calculation method of control rod reactivity relating to hexagonal assembly, which are used generally in fast reactor, we have investigated the calculation method, the problems to rise during calculation, the degrees of calculation and the enhancement of calculation modeling so on, and estimated the application of calculation method through comparison and analysis of calculation result using the effective cross section generation system, TRANSX/TWODANT, and neutron flux calculation system, diffusion theory code DIF-3D, which are belonged to K-CORE System, and determined the basic calculation method, and extracted the present calculation problem in case of application in K-CORE System and the future improvement items so on.

  4. Establishment of analysis procedure for control rod reactivity worth

    International Nuclear Information System (INIS)

    Song, Hoon; Kim, Young Il; Kim, Sang Ji; Kim, Young In

    2001-03-01

    As to the calculation method of control rod reactivity relating to hexagonal assembly, which are used generally in fast reactor, we have investigated the calculation method, the problems to rise during calculation, the degrees of calculation and the enhancement of calculation modeling so on, and estimated the application of calculation method through comparison and analysis of calculation result using the effective cross section generation system, TRANSX/TWODANT, and neutron flux calculation system, diffusion theory code DIF-3D, which are belonged to K-CORE System, and determined the basic calculation method, and extracted the present calculation problem in case of application in K-CORE System and the future improvement items so on

  5. The Hitrex Programme: unperturbed HTR lattice and control rod measurements

    Energy Technology Data Exchange (ETDEWEB)

    Beynon, A J; Nunn, D L

    1972-06-15

    Reactivity, power distributions, plutonium production and fast neutron graphite damage are being studied at Berkeley Nuclear Laboratories (BNL) on the HTR 'Hitrex' reactor under cold clean conditions. Rod interactions, important in assessing local criticality hazards, are receiving special attention in the measurements. The proposals for the first two series of measurements on Hitrex are discussed in this note, Hitrex 1a being the unperturbed reactor, and Hitrex 1b the same fuel array but with a number of different control absorber loadings in it. Common to both series will be cross pin, cross block and cross core measurements of power rating, thermal spectrum and damage dose distributions, so that these will be known as functions of the fuel, reflector and absorber environment.

  6. Study on dynamic lifting characteristics of control rod drive mechanism

    International Nuclear Information System (INIS)

    Shen Xiaoyao

    2012-01-01

    Based on the equations of the electric circuit and the magnetic circuit and analysis of the dynamic lifting process for the control rod drive mechanism (CRDM), coupled magnetic-electric-mechanical equations both for the static status and the dynamic status are derived. The analytical method is utilized to obtain the current and the time when the lift starts. The numerical simulation method of dynamic analysis recommended by ASME Code is utilized to simulate the dynamic lifting process of CRDM, and the dynamic features of the system with different design gaps are studied. Conclusions are drawn as: (1) the lifting-start time increases with the design gap, and the time for the lifting process is longer with larger gaps; (2) the lifting velocity increases with time; (3) the lifting acceleration increases with time, and with smaller gaps, the impact acceleration is larger. (author)

  7. Method of cleaning pipeline in control rod drive

    International Nuclear Information System (INIS)

    Baba, Mikiya.

    1993-01-01

    A step of filtering cleaning water by a provisional filter unit and a step of returning filtered cleaning water to a provisional tank are disposed. That is, purified water is stored in the provisional tank and it is sucked by a driving pump under pressure by way of a suction filter into the pipelines in a control rod drive system to clean them. Purified water after the cleaning is filtered by the provisional filter unit and returned to the provisional tank by way of provisional pipelines to form a closed loop. A great amount of purified water to be used is no more necessary by thus changing the water passing cleaning method to the recycling cleaning method, which moderate influences on other steps using purified water and ensure a cleaning step for pipelines in a CRD system, in addition, save the steps for plant construction greatly. (N.H.)

  8. Radial brake assembly for a control rod drive

    International Nuclear Information System (INIS)

    Hekmati, A.; Gibo, E.Y.

    1992-01-01

    This patent describes a brake assembly for a control rod drive for selectively preventing travel of a control rod in a nuclear reactor vessel. It comprises a shaft having a longitudinal centerline axis; means for selectively rotating the shaft in a first direction and in a second direction, opposite to the first direction; a stationary housing having a central aperture receiving the shaft; a frame fixedly joined to the housing and having a guide hole; a rotor disc fixedly connected to the shaft for rotation therewith and having at least one rotor tooth extending radially outwardly from a perimeter thereof, the rotor tooth having a locking surface and an inclined surface extending therefrom in a circumferential direction; a brake member disposed adjacent to the rotor disc perimeter and including a base, at least one braking tooth having a locking surface extending therefrom in a circumferential direction, and a plunger extending radially outwardly from the base and slidably joined to the frame through the guide hole; the rotor tooth and the braking tooth being complementary to each other; and means for selectively positioning the brake member in a deployed position abutting the rotor disc perimeter for allowing the braking tooth locking surface to contact the rotor tooth locking surface for preventing rotation of the shaft in the first direction, and in a retracted position spaced radially away from the rotor disc for allowing the rotor disc and the shaft to rotate without restraint from the brake member, the positioning means including a tubular solenoid fixedly joined to the frame and having a central bore disposed around the brake member plunger and effective for sliding the brake member plunger relative to the frame for positioning the brake member in the deployed and retracted positions

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  10. Detection and mitigating rod drive control system degradation in Westinghouse PWRs

    International Nuclear Information System (INIS)

    Gunther, W.; Sullivan, K.

    1990-01-01

    A study of the effects of aging on the Westinghouse Control Rod Drive (CRD) System was performed as part of the US NRC's Nuclear Plant aging Research (NPAR) Program. For the study, the CRD system boundary includes the power and logic cabinets associated with the manual control rod movement, and the control rod mechanism itself. The aging-related degradation of the interconnecting cables and connectors and the rod position indicating system also were considered. This paper presents the results of that study pertaining to the electrical and instrumentation portions of the CRD system including ways to detect and mitigate system degradation

  11. The Interaction Between Control Rods as Estimated by Second-Order One-Group Perturbation Theory

    International Nuclear Information System (INIS)

    Persson, Rolf

    1966-10-01

    The interaction effect between control rods is an important problem for the reactivity control of a reactor. The approach of second order one-group perturbation theory is shown to be attractive due to its simplicity. Formulas are derived for the fully inserted control rods in a bare reactor. For a single rod we introduce a correction parameter b, which with good approximation is proportional to the strength of the absorber. For two and more rods we introduce an interaction function g(r ij ), which is assumed to depend only on the distance r ij between the rods. The theoretical expressions are correlated with the results of several experiments in R0, ZEBRA and the Aagesta reactor, as well as with more sophisticated calculations. The approximate formulas are found to give quite good agreement with exact values, but in the case of about 8 or more rods higher-order effects are likely to be important

  12. The Interaction Between Control Rods as Estimated by Second-Order One-Group Perturbation Theory

    Energy Technology Data Exchange (ETDEWEB)

    Persson, Rolf

    1966-10-15

    The interaction effect between control rods is an important problem for the reactivity control of a reactor. The approach of second order one-group perturbation theory is shown to be attractive due to its simplicity. Formulas are derived for the fully inserted control rods in a bare reactor. For a single rod we introduce a correction parameter b, which with good approximation is proportional to the strength of the absorber. For two and more rods we introduce an interaction function g(r{sub ij}), which is assumed to depend only on the distance r{sub ij} between the rods. The theoretical expressions are correlated with the results of several experiments in R0, ZEBRA and the Aagesta reactor, as well as with more sophisticated calculations. The approximate formulas are found to give quite good agreement with exact values, but in the case of about 8 or more rods higher-order effects are likely to be important.

  13. On-line fuel and control rod integrity management in BWRs

    International Nuclear Information System (INIS)

    Larsson, Irina; Sihver, Lembit

    2011-01-01

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

  14. Optical-optical double resonance, laser induced fluorescence, and revision of the signs of the spin-spin constants of the boron carbide (BC) free radical

    Science.gov (United States)

    Sunahori, Fumie X.; Nagarajan, Ramya; Clouthier, Dennis J.

    2015-12-01

    The cold boron carbide free radical (BC X 4Σ-) has been produced in a pulsed discharge free jet expansion using a precursor mixture of trimethylborane in high pressure argon. High resolution laser induced fluorescence spectra have been obtained for the B 4Σ--X 4Σ- and E 4Π-X 4Σ- band systems of both 11BC and 10BC. An optical-optical double resonance (OODR) scheme was implemented to study the finer details of both band systems. This involved pumping a single rotational level of the B state with one laser and then recording the various allowed transitions from the intermediate B state to the final E state with a second laser by monitoring the subsequent E-X ultraviolet fluorescence. In this fashion, we were able to prove unambiguously that, contrary to previous studies, the spin-spin constant λ is negative in the ground state and positive in the B 4Σ- excited state. It has been shown that λ″ expected based on a semiempirical second order perturbation theory calculation of the magnitude of the spin-spin constant. The OODR spectra have also been used to validate our assignments of the complex and badly overlapped E 4Π-X 4Σ- 0-0 and 1-0 bands of 11BC. The E-X 0-0 band of 10BC was found to be severely perturbed. The ground state main electron configuration is …3σ24σ25σ11π22π0 and the derived bond lengths show that there is a 0.03 Å contraction in the B state, due to the promotion of an electron from the 4σ antibonding orbital to the 5σ bonding orbital. In contrast, the bond length elongates by 0.15 Å in the E state, a result of promoting an electron from the 5σ bonding orbital to the 2π antibonding orbitals.

  15. Calculation of control rod worth with mutual interaction

    International Nuclear Information System (INIS)

    Balthar, M.C.V.; Oliveira Vellozo, S. de; Carvalho Vital, H. de

    1989-01-01

    This work presents a two-dimensional model for determining the total worth of a set of N absorbing rods. The model simplifies the evaluation of the interaction coefficient among rods by analysing them in pairs and attributes to it a purely geometrical character. Comparisons with conventional calculational methods indicate that the results are in error by less than 6%. (author) [pt

  16. Control rod shadowing and anti-shadowing effects in a large gas-cooled fast reactor

    International Nuclear Information System (INIS)

    Girardin, G.; Chawla, R.; Rimpault, G.; Coddington, P.

    2007-01-01

    An investigation of control rod shadowing and anti-shadowing (interaction) effects has been carried out in the context of a design study of the control rod pattern for the large 2400 MWth Generation IV Gas-cooled Fast Reactor (GFR). For the calculations, the deterministic code system ERANOS-2.0 has been used, in association with a full core model including a European Fast Reactor (EFR)-type pattern for the control rods. More specifically, the core contains a total of 33 control (CSD) and safety (DSD) rods implemented in three banks: -1) a first bank of 6 CSD rods, placed at 64 cm from core centre in the inner fuel zone (Pu content 16.3 % vol.), -2) a safety bank consisting of 9 DSD rods, at an average distance of 118 cm, and -3) a third bank with 18 CSD rods, placed at 171 cm, i.e. at the interface between the inner and outer (Pu content 19.2 % vol.) core regions. Each control rod has been modelled as a homogeneous material containing 90%-enriched B 4 C, steel and helium. Considerable shadowing effects have been observed between the first bank and the safety bank, as also between individual rods within the first bank. Large anti-shadowing effects take place in an even greater number of the studied rod configurations. The largest interaction is between the two CSD banks, the anti-shadowing value being 46% in this case, implying that the total rod worth is increased by a factor of almost 2 when compared to the sum of the individual bank values. Additional investigations have been performed, in particular the computation of the first order eigenvalue and the eigenvalue separation. The main finding is that the interactions are lower when one of the control rod banks is located at a radial position corresponding to half the core radius. (authors)

  17. Development of embedded Control System for Control and Safety Rod Drive Mechanisms (CSRDMs) of PFBR

    International Nuclear Information System (INIS)

    Kameswari, K.; Palanisami, K.; Thirugnana Murthy, D.; Murali, N.; Satyamurty, S.A.V.

    2013-01-01

    Prototype Fast Breeder Reactor (PFBR), a 500 MWe, Sodium cooled, fast breeder reactor is nearing completion at Kalpakkam, Tamil Nadu. PFBR has two independent, fast acting and diverse shutdown systems, one with nine Control and Safety Rods (CSRs) and another with three Diverse Safety Rods (DSRs), with independent driving mechanisms called CSRDMs and DSRDMs respectively. This paper deals with the development of Real Time Computer based Control system for controlling nine CSRDMs with model based software development environment - SCADE (Safety Critical Application Development Environment). (author)

  18. Control-rod interference effects observed during reactor physics experiments with nuclear ship 'MUTSU'

    International Nuclear Information System (INIS)

    Itagaki, Masafumi; Miyoshi, Yoshinori; Gakuhari, Kazuhiko; Okada, Noboru; Sakai, Tomohiro.

    1993-01-01

    The control rods in the reactor of the nuclear ship MUTSU are classified into four groups: groups G1 and G2 are located in the central part of the core, while groups G3 and G4 are in the peripheral zone of the core. Several types of mutual interference effects among these control-rod groups were observed during reactor physics experiments with this reactor. During normal hot operations, positive shadowing was dominant between the G1 and G2 groups; the degree of the shadowing effect of one rod group depended on the position of the other rod group. Both positive and negative shadowing effects occurred between an inner rod group (G1 or G2) and an outer group (G3 or G4) depending on the three-dimensional arrangement of the control rods. The rod worths of G1 and G2 increased as a result of slight core burnup, about 1,400 MWd/t, mainly due to the decrease in shadowing effects resulting from a change in control-rod pattern. A three-dimensional diffusion calculation with internal control-rod boundary conditions has proved to be useful for analyzing these various interaction effects. (author)

  19. Control rod pattern exchange in a BWR/6 utilizing gang mode withdrawal

    International Nuclear Information System (INIS)

    Auvil, A.B. Jr.; Aldemir, T.; Hajek, B.K.

    1986-01-01

    The use of checkerboard pattern of alternating inserted and fully withdrawn control rods and the uneven void distribution in boiling water reactor (BWR) cores can cause large burnup gradients even after a short time of operation. To compensate for these effects, power has to be reshaped periodically (typically every two full-power months) by individually manipulating the control rods. During this manipulation process (called the control rod pattern exchange), the core power is reduced to 60% of nominal power by means of flow reduction to limit power swings to tolerable levels and to ensure that fuel thermal limits are not exceeded. A control rod pattern exchange by individual rod manipulation typically takes 4 to 8 h and represents a large cost burden to the utility in terms of reduced system output. The latest generation of BWRs, the BWR/6, possesses the capability to simultaneously move up to four symmetrically located control rods. The rods corresponding to a given gang may have rotational symmetry, mirror symmetry, or a combination of the two. This paper presents a pattern exchange procedure that exploits the capability of gang mode rod withdrawal to reduce the pattern exchange execution time and radial power distribution asymmetry associated with individual rod manipulation. The working model used in the study is the Perry Nuclear Power Plant Unit 1, located in Perry, Ohio, and owned by the Cleveland Electric Illuminating Company

  20. 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

  1. Method of changing the control rod pattern in BWR type reactors

    International Nuclear Information System (INIS)

    Yoshida, Kenji.

    1984-01-01

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

  2. A connection-disconnection device for the nuclear reactor control rods

    International Nuclear Information System (INIS)

    Drean, H.; Breant, C.

    1992-01-01

    An automatic and relatively light system is developed to connect and disconnect the control rods in a nuclear reactor; it is designed to apply, in a controlled manner, constraints on the rods and to detect a possible misalignment of the connection-disconnection means in the corresponding butt

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  4. Coupling device of the control rod and of the drive mechanism

    International Nuclear Information System (INIS)

    Savary, F.

    1986-01-01

    The invention proposes a coupling device removable in which the connection between the upper head of the control rod and the drive mechanism is a real rigid fixing, in the mechanical sense of the term, suppressing longitudinal play and allowing to restrict the momenta occurring when locating the control rods [fr

  5. Control rod position fault diagnosis and its software realization of pressurized water reactor

    International Nuclear Information System (INIS)

    Chang Zhengke; Shao Dinghong

    2004-11-01

    PLC software is adopted in the Rod Position Monitoring System of QS2NPS. By this software, the position of control rods can be monitored in real time, the abnormal phenomena can be identified immediately, the correctness and timeliness of fault diagnosis are improved remarkably. the identification and recordance of rod position fault, the performance validation of measure channel are realized also. The function and effect of this software are introduced. (authors)

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

    International Nuclear Information System (INIS)

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

    1991-01-01

    This patent describes, for use in facilitating replacement of a neutron absorber control rod on a control assembly spider structure, an end plug. It comprises a pair of separate upper and lower plug portions; the upper section of the upper plug portion being configured for rigid attachment; the middle section of the upper plug portion having angularly displaced flat surfaces formed on the exterior

  7. Optimization of boiling water reactor control rod patterns using linear search

    International Nuclear Information System (INIS)

    Kiguchi, T.; Doi, K.; Fikuzaki, T.; Frogner, B.; Lin, C.; Long, A.B.

    1984-01-01

    A computer program for searching the optimal control rod pattern has been developed. The program is able to find a control rod pattern where the resulting power distribution is optimal in the sense that it is the closest to the desired power distribution, and it satisfies all operational constraints. The search procedure consists of iterative uses of two steps: sensitivity analyses of local power and thermal margins using a three-dimensional reactor simulator for a simplified prediction model; linear search for the optimal control rod pattern with the simplified model. The optimal control rod pattern is found along the direction where the performance index gradient is the steepest. This program has been verified to find the optimal control rod pattern through simulations using operational data from the Oyster Creek Reactor

  8. A method of taking control rod history into account in core simulation calculations for BWR'S

    International Nuclear Information System (INIS)

    Hojerup, C.F.; Nonbol, E.

    1990-01-01

    The problem of taking control rod history into account in core simulator codes using precalculated cross sections has been examined, and two methods have been devised and tested. The very demanding first method, using the accumulated control rod in burn-up as a parameter, turned out to be even more inaccurate than the much less demanding second method, which only requires two full burn-up histories, one with the control rod in all the time, and another with the control rod out all the time. From the analysis it can be seen that the proper treatment of the control rod history is quite important, both for the cross sections, as several per cent on the reactivity are at stake, as for the pin powers, which for some pins are very much affected

  9. The feasibility of using neural networks for determination of control rod elevation in a PWR

    International Nuclear Information System (INIS)

    Garis, N.S.; Temesvari, E.; Pazsit, I.

    1996-08-01

    This paper presents the results of a preliminary study on using neural networks for determination of the axial position of control rods in PWRs. The method is based on the dependence of the axial flux profile on control rod elevation in a reactor. This flux profile can be measured by e.g. a moveable detector in an operating plant. However, in this preliminary study the flux profile is only calculated using an advanced core code for several axial positions of a partially inserted control rod. The calculated fluxes with corresponding positions of the control rod are used for training a neural network. Using the trained network it is then possible to determine the unknown axial position of a control rod elevation from the corresponding axial flux profile. 10 refs

  10. Experimental study of the pressure discharge process for the hydraulic control rod drive system stepped cylinder

    International Nuclear Information System (INIS)

    Wang, Jinhua; Bo, Hanliang; Zheng, Wenxiang

    2002-01-01

    The pressure discharge process from the stepped cylinder of the Hydraulic Control Rod Drive System (HCRDS) was studied experimentally in the HCRDS experimental loop for the 200 MW Nuclear Heating Reactor (NHR-200). The results showed that the differential pressure between the outside and the inside of the stepped cylinder increased rapidly to the desired value so that the force induced by the differential pressure which pushes the out tube of stepped cylinder was large enough. Therefore, if the hydraulic control rod were jammed, the pressure could push the hydraulic control rod to overcome the frictional resistance to insert the control rod into the reactor core. The experimental results verified that this design would solve the problem of hydraulic control rod jamming during an accident. (author)

  11. Prevention and preservation aid system for control rod drives

    International Nuclear Information System (INIS)

    Ishisato, Shin-ichi; Yamamoto, Yoko.

    1992-01-01

    The system of the present invention can select control rod drives (CRD) as an object of inspection, and can manage maintenance hysteresis even by unskilled persons upon maintenance operation for the CRD. That is, the system of the present invention comprises a data base concerning prevention and preservation for the CRD and hydraulic pressure control unit (HCU), a data base management device for retrieving and managing the intelligence of the data base and a maintenance data base for storing data measured based on the data base on every periodical inspections. Further, it also comprises a function for displaying, on a map, the CRD to be disassembled and inspected upon periodical inspection on every inspection recommendation priority groups, based on these data base. Further, it also comprises a function for evaluating exchange hysteresis maintenance data for incore structures which require periodical exchange. As a result, high reliability of the CRD can be maintained and reliability of a nuclear power plant can further be improved. (I.S.)

  12. Bottom-mounted control rod drive mechanism for KJRR

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jin Haeng; Kim, Sanghaun; Yoo, Yeon-Sik, E-mail: yooys@kaeri.re.kr; Cho, Yeong-Garp; Huh, Hyung; Lee, Hyokwang; Sun, Jong-Oh; Ryu, Jeong-Soo

    2016-04-15

    Highlights: • The basic design features and characteristics of the KJRR BMCRDM are described. • The similarities and differences of some research reactor CRDMs are compared. • The current status of the design and development of the CRDM is described. • The future plan of the qualification tests of the CRDM is summarized. - Abstract: The KIJANG research reactor (KJRR), which is currently being designed by Korea Atomic Energy Research Institute, is a pool type research reactor with 15 MW of thermal power. Contrary to the top-mounted control rod drive mechanism (CRDM), the main drive mechanism of the KJRR CRDM is located in a reactivity control mechanism room under the reactor pool bottom. Recently, we accomplished the design and development of a prototype CRDM. In this paper, we introduce the basic design concept of the bottom-mounted CRDM for KJRR, and compare the similarities and differences of some research reactor CRDMs. The current status of the prototype CRDM development based on a finite element analysis and experimental verification, and the future plan of the CRDM qualification tests, are both described.

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

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  14. Control-rod parametrical studies in the framework of the PRE-RACINE and RACINE programs

    International Nuclear Information System (INIS)

    Humbert, G.; Ruelle, B.; Daguzan, G.; Stanculescu, A.; Kappler, F.; Scholtyssek, W.; Bouscavet, D.; Martini, M.; Broccoli, U.

    1982-01-01

    A control-rod experimental program is presented. This program, established in the frame of PRE-RACINE and RACINE common DEBENE, Italian and French experiments at MASURCA facility, is still under progress at the moment. The results, limited to single central rod worth are already available. For these experiments, a parametrical approach has been used. The effects of rod worth, varied separatly by rod side, boron enrichment and core size, on experiment to calculation relative discrepancy (E-C)/C can be drawn out

  15. On-line method to identify control rod drops in Pressurized Water Reactors

    International Nuclear Information System (INIS)

    Souza, T.J.; Martinez, A.S.; Medeiros, J.A.C.C.; Palma, D.A.P.; Gonçalves, A.C.

    2014-01-01

    Highlights: • On-line method to identify control rod drops in PWR reactors. • Identification method based on the readings of the ex-core detector. • Recognition of the patterns in the ex-core detector responses. - Abstract: A control rod drop event in PWR reactors leads to an unsafe operating condition. It is important to quickly identify the rod to minimise undesirable effects in such a scenario. The goal of this work is to develop an online method to identify control rod drops in PWR reactors. The method entails the construction of a tool based on ex-core detector responses. It proposes to recognize patterns in the neutron ex-core detectors responses and thus to make an online identification of a control rod drop in the core during the reactor operation. The results of the study, as well as the behaviour of the detector responses demonstrated the feasibility of this method

  16. Study for identification of control rod drops in PWR reactors at any burnup step

    International Nuclear Information System (INIS)

    Souza, Thiago J.; Martinez, Aquilino S.; Medeiros, Jose A.C.C.; Goncalves, Alessandro C.

    2013-01-01

    The control rod drop event in PWR reactors induces an unsafe operating condition. Therefore, in a scenario of a control rod drop is important to quickly identify the rod to minimize undesirable effects. The objective of this work is to develop an on-line method for identification of control rod drop in PWR reactors. The method consists on the construction of a tool that is based on the ex-core detector responses. Therefore, it is proposed to recognize patterns in the neutron ex-core detectors responses and thus to identify on-line a control rod drop in the core during the reactor operation. The results of the study, as well as the behavior of the detector responses, demonstrated the feasibility of this method. (author)

  17. Study for identification of control rod drops in PWR reactors at any burnup step

    Energy Technology Data Exchange (ETDEWEB)

    Souza, Thiago J.; Martinez, Aquilino S.; Medeiros, Jose A.C.C.; Goncalves, Alessandro C., E-mail: tsouza@nuclear.ufrj.br, E-mail: aquilino@lmp.ufrj.br, E-mail: canedo@lmp.ufrj.br, E-mail: alessandro@nuclear.ufrj.br [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE/UFRJ), RJ (Brazil). Programa de Engenharia Nuclear; Palma, Daniel A.P., E-mail: dapalma@cnen.gov.br [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil)

    2013-07-01

    The control rod drop event in PWR reactors induces an unsafe operating condition. Therefore, in a scenario of a control rod drop is important to quickly identify the rod to minimize undesirable effects. The objective of this work is to develop an on-line method for identification of control rod drop in PWR reactors. The method consists on the construction of a tool that is based on the ex-core detector responses. Therefore, it is proposed to recognize patterns in the neutron ex-core detectors responses and thus to identify on-line a control rod drop in the core during the reactor operation. The results of the study, as well as the behavior of the detector responses, demonstrated the feasibility of this method. (author)

  18. Application of a spatial modal kinetic model for determination of control rod worths

    International Nuclear Information System (INIS)

    Gomez, A.; Waldman, R.M.

    1993-01-01

    A high-precision rod drop method based on a modal kinetic model, with low dependence on detector location, is proposed to measure the reactivity worth of control rods. This value is obtained from data adjustment for the delayed evolution. It is necessary to maintain the experimental data fluctuation in a small value so that the error of the control rod worth should not be large. A model was developed in order to relate the fluctuation with some parameters which may be modified in the measuring process. The method was applied in the RA-6 reactor to measure control rod worth. For practical purpose it was found that the method can be applied to 15 dollars and it does not depend on relative detector and control rod locations, as the method based on the Point Reactor Model does. (author). 2 refs

  19. Analysis of the burnup of the control rods with the COREMASTER-Presto code

    International Nuclear Information System (INIS)

    Hernandez, J.L.; Alonso, G.; Perusquia, R.; Montes, J.L.; Hernandez, H.

    2003-01-01

    An evaluation of the capacity of the COREMASTER-Presto code, to evaluate generically the burnt of the control bars in the Laguna Verde reactors plant (CLV) is made. It was found that the code only reports burnt values of the control rods in MWD/TM, in spite of having with a second order polynomial model, for the conversion to remainder of the Boron-10 (B-10). It was observed that said model is adequate only for burnt smaller to 45,000 MWD/TM. To evaluate the burnt of the control rods it was reproduced the balance cycle of 18 months for the CLV, executing Cm-Presto during 13 consecutive cycles. First without rod burnt, taking this as the base case. Later on, cases with 1, 2 and up to 13 cycles with rod burnt were generated. When comparing results it was observed that the control rods pattern it loses reactivity lineally with the burnt one. By each 10 G Wd/T of burnt of the nucleus it is decreased the reactivity of the pattern rods ∼ 1 pcm in hot condition and of ∼ 20 pcm in cold condition. When burning three cycles those rods more burnt reached the 13,900 MWD/TM, equivalent to 36% of B-10 reduction, near value to 34% proposed by aging in the one lost study of B-10. It was observed that Cm-Presto it doesn't burn the superior node of the control rods when these are completely extracted. A one big lost of B-10, of the order of 50%, it represents only a decrease of 11% of the reactivity value of the rod. One can affirm that even when it is strongly decreased the content of B-10, the rod is continue considering as a black absorber, that is to say, thermal neutron that enters in the neutron rod that is absorbed. (Author)

  20. Calculation of control rod oscillations in a hexagonal flow channel by means of the non-stationary pressure distribution around the rods

    International Nuclear Information System (INIS)

    Grunwald, G.; Mueller, E.

    1983-08-01

    For the computation of control rod oscillations in a flow channel we set up the differential equations for the non-stationary pressure distribution around the control elements which are coupled with the motion equations of the rods. The equation system is solved by means of a finite difference method. An example shows the efficiency of the numerical calculation procedure. (author)

  1. Experience feedback of operational events of the control rod assembly and its drive mechanism in nuclear power plants

    International Nuclear Information System (INIS)

    Zhou Hong; Xiao Zhi; Tao Shusheng; Zheng Lixin; Chen Zhaolin

    2013-01-01

    Seventeen operational events of the control rod assembly and its drive mechanism are collected from 1992 to 2012 important nuclear operational events and feedback in referred nuclear power plants. After investigated and classified, several important issues, such as the impact of control rod swell and fuel assembly distortion, control rod drive mechanism leakage, and the control system reliability of control rod, are emphatically analyzed. Some suggestions of experience feedback are proposed. (authors)

  2. Aging assessment of BWR control rod drive systems

    International Nuclear Information System (INIS)

    Greene, R.H.

    1991-01-01

    This study examines the aging phenomena associated with boiling water reactor (BWR) control rod drive mechanisms (CRDMs) and assess the merits of various methods of managing this aging. Information for this study was acquired from (1) the results of a special CRDM aging questionnaire distributed to each US BWR utility, (2) a first-of-its-kind workshop held to discuss CRDM aging and maintenance concerns, (3) an analysis of Nuclear Plant Reliability Data System (NPRDS) failure cases attributed to the CRD system, and (4) personal information exchange with industry experts. As part of this study, nearly 3500 NPRDS failure reports have been analyzed to examine the prevailing failure trends for CRD system components. An investigation was conducted to summarize the occurrence frequency of these component failures, discovery methods, reported failure causes, their respective symptoms, and actions taken by utilities to restore component and system service. The results of this research have identified the predominant CRDM failure modes and causes. In addition, recommendations are presented that identify specific actions utilities can implement to mitigate CRDM aging. An evaluation has also been made of certain maintenance practices and tooling which have enabled some utilities to reduce ALARA exposures received from routine CRDM replacement and rebuilding activities. 5 refs., 8 figs., 2 tabs

  3. Aging assessment of BWR control rod drive systems

    International Nuclear Information System (INIS)

    Greene, R.H.

    1992-01-01

    This Phase 1 Nuclear Plant Aging Research (NPAR) study examines the aging phenomena associated with boiling water reactor (BWR) control rod drive mechanisms (CRDMs) and assesses the merits of various methods of managing this aging. Information for this study was acquired from (1) the results of a special CRDM aging questionnaire distributed to each US BWR utility, (2) a first-of-its-kind workshop held to discuss CRDM aging and maintenance concerns, (3) an analysis of NPRDS failure cases attributed to the CRD system, and (4) personal information exchange. As part of this study, nearly 3,500 NPRDS failure reports have been analyzed to examine the prevailing failure trends for CRD system components. An investigation has been conducted that summarizes the occurrence frequency of these component failures, discovery methods, reported failure causes, their respective symptoms, and actions taken by utilities to restore component and system service. The results of this research have identified the predominant CRDM failure modes and causes. In addition, recommendations are presented regarding specific actions that utilities can implement to mitigate CRDM aging. An evaluation has also been made of certain practices and tooling which have enabled some utilities to reduce ALARA exposures received from routine CRDM replacement and rebuilding activities

  4. Uncontrolled withdrawal of a control rod without scram

    International Nuclear Information System (INIS)

    Abou-EL-Maaty, Talal

    2008-01-01

    In the present work the thermal-hydraulics of reactivity-induced transients in low enriched uranium (LEU) core of a typical material test research reactor (MTR) are analyzed using the previous program developed by Khater et al. The analysis was done for uncontrolled withdrawal of a control rod with scram-disabled conditions. Initiating reactivity events with and without the influence of reactivity efficiency curve ('S' curve) were considered. The results of the proposed transients are analyzed and compared with each other. In transient without the 'S' curve influence, a high primary peak power of 406.18 MW is attained and a clad melt down takes place after 1.85 s. In the transient with the 'S' curve influence, a high super prompt-critical situation is produced (1.762$ at 0.895 s) with a very high primary peak power of 801.05 MW at 0.912 s. Also, a fast clad melt down is resulted in the hot channel at 1.088 s and a stable film boiling is established. This study indicates that, compared to the application of linear reactivity curve, the application of the reactivity efficiency curve results in the prediction of higher peaks in power and temperatures (fuel, clad and coolant) with a fast clad melt down

  5. Uncontrolled withdrawal of a control rod without SCRAM

    International Nuclear Information System (INIS)

    Abou-El-Maaty, T.

    2007-01-01

    In the present work, the thermal-hydraulic analysis of reactivity-induced transients in a Low Enriched Uranium (LEU) core of a typical material test research reactor is conducted using the previous program developed Khater et al. The analysis was done for the uncontrolled withdrawal of a control rod under scram-disabled conditions. The initiating event reactivity was considered with and without influence of the reactivity efficiency curve (''S'' curve). The results of the transient calculations are analyzed and compared with each other. In the transient without the ''S'' curve influence, a high primary peak power of 406.18 MW is attained and a clad melt down is occurring after 1.85 s. In the transient with the ''S'' curve influence, a super prompt highly critical situation is produced (1.762 $ at 0.895 s) with a very high primary peak power of 801.05 MW at 0.912 s. A fast clad melt down is resulting in the hot channel at 1.088 s and a stable film boiling is occurring. This study shows that the influence of the reactivity efficiency curve results in higher peaks in power and temperatures (fuel, clad and coolant) with a fast clad melt down than that of a linear assumption. (orig.)

  6. Effect of component aging on PWR control rod drive systems

    International Nuclear Information System (INIS)

    Grove, E.; Gunther, W.; Sullivan, K.

    1992-01-01

    An aging assessment of PWR control rod drive (CRD) systems has been completed as part of the US NRC Nuclear Plant Aging Research (NPAR) Program. The design, construction, maintenance, and operation of the Babcock ampersand Wilcox (B ampersand W), Combustion Engineering (CE), and Westinghouse (W) systems were evaluated to determine the potential for degradation as each system ages. Operating experience data were evaluated to identify the predominant failure modes, causes, and effects. This, coupled with an assessment of the materials of construction and operating environment, demonstrate that each design is subject to degradation, which if left unchecked, could affect its safety function as the plant ages. An industry survey, conducted with the assistance of EPRI and NUMARC, identified current CRD system maintenance and inspection practices. The results of this survey indicate that some plants have performed system modifications, replaced components, or augmented existing preventive maintenance practices in response to system aging. The survey results also supported the operating experience data, which concluded that the timely replacement of degraded components, prior to failure, was not always possible using existing condition monitoring techniques. The recommendations presented in this study also include a discussion of more advanced monitoring techniques, which provide trendable results capable of detecting aging

  7. Aging assessment of BWR control rod drive systems

    International Nuclear Information System (INIS)

    Greene, R.H.

    1991-01-01

    This Phase 1 Nuclear Plant Aging Research (NPAR) study examines the aging phenomena associated with boiling water reactor (BWR) control rod drive mechanisms (CRDMs) and assesses the merits of various methods of managing this aging. Information for this study was acquired from (1) the results of a special CRDM aging questionnaire distributed to each US BWR utility, (2) a first-of-its-kind workshop held to discuss CRDM aging and maintenance concerns, (3) an analysis of NPRDS failure cases attributed to the CRD system, and (4) personal information exchange. As part of this study, nearly 3,500 NPRDS failure reports have been analyzed to examine the prevailing failure trends for CRD system components. An investigation has been conducted that summarizes the occurrence frequency of these component failures, discovery methods, reported failure causes, their respective symptoms, and actions taken by utilities to restore component and system service. The results of this research have identified the predominant CRDM failure modes and causes. In addition, recommendations are presented regarding specific actions that utilities can implement to mitigate CRDM aging. An evaluation has also been made of certain practices and tooling which have enabled some utilities to reduce ALARA exposures received from routine CRDM replacement and rebuilding activities

  8. Experimental Breeder Reactor-II automatic control-rod-drive system

    International Nuclear Information System (INIS)

    Christensen, L.J.

    1983-01-01

    A computer-controlled automatic control rod drive system (ACRDS) was designed and operated in EBR-II during reactor runs 121 and 122. The ACRDS was operated in a checkout mode during run 121 using a low worth control rod. During run 122 a high worth control rod was used to perform overpower transient tests as part of the LMFBR oxide fuels transient testing program. The testing program required an increase in power of 4 MW/s, a hold time of 12 minutes and a power decrease of 4 MW/s. During run 122, 13 power transients were performed

  9. New approach for control rod position indication system for light water power reactor

    International Nuclear Information System (INIS)

    Bahuguna, Sushil; Dhage, Sangeeta; Nawaj, S.; Salek, C.; Lahiri, S.K.; Marathe, P.P.; Mukhopadhyay, S.; Taly, Y.K.

    2015-01-01

    Control rod position indication system is an important system in a nuclear power plant to monitor and display control rod position in all regimes of reactor operation. A new approach to design a control rod position indication system for sensing absolute position of control rod in Light Water Power Reactor has been undertaken. The proposed system employs an inductive type, hybrid measurement strategy providing both analog position as well as digital zone indication with built-in temperature compensation. The new design approach meets single failure criterion through redundancy in design without sacrificing measurement resolution. It also provides diversity in measurement technique by indirect position sensing based on analysis of drive coil current signature. Prototype development and qualification at room temperature of the control rod position indication system (CRPIS) has been demonstrated. The article presents the design philosophy of control rod position indication system, the new measurement strategy for sensing absolute position of control rod, position estimation algorithm for both direct and indirect sensing and a brief account associated processing electronics. (author)

  10. Optical-optical double resonance, laser induced fluorescence, and revision of the signs of the spin-spin constants of the boron carbide (BC) free radical

    Energy Technology Data Exchange (ETDEWEB)

    Sunahori, Fumie X. [Department of Chemistry and Physics, Franklin College, Franklin, Indiana 46131 (United States); Nagarajan, Ramya; Clouthier, Dennis J., E-mail: dclaser@uky.edu [Department of Chemistry, University of Kentucky, Lexington, Kentucky 40506-0055 (United States)

    2015-12-14

    The cold boron carbide free radical (BC X {sup 4}Σ{sup −}) has been produced in a pulsed discharge free jet expansion using a precursor mixture of trimethylborane in high pressure argon. High resolution laser induced fluorescence spectra have been obtained for the B {sup 4}Σ{sup −}–X {sup 4}Σ{sup −} and E {sup 4}Π–X {sup 4}Σ{sup −} band systems of both {sup 11}BC and {sup 10}BC. An optical-optical double resonance (OODR) scheme was implemented to study the finer details of both band systems. This involved pumping a single rotational level of the B state with one laser and then recording the various allowed transitions from the intermediate B state to the final E state with a second laser by monitoring the subsequent E–X ultraviolet fluorescence. In this fashion, we were able to prove unambiguously that, contrary to previous studies, the spin-spin constant λ is negative in the ground state and positive in the B {sup 4}Σ{sup −} excited state. It has been shown that λ″ < 0 is in fact expected based on a semiempirical second order perturbation theory calculation of the magnitude of the spin-spin constant. The OODR spectra have also been used to validate our assignments of the complex and badly overlapped E {sup 4}Π–X {sup 4}Σ{sup −} 0-0 and 1-0 bands of {sup 11}BC. The E–X 0-0 band of {sup 10}BC was found to be severely perturbed. The ground state main electron configuration is …3σ{sup 2}4σ{sup 2}5σ{sup 1}1π{sup 2}2π{sup 0} and the derived bond lengths show that there is a 0.03 Å contraction in the B state, due to the promotion of an electron from the 4σ antibonding orbital to the 5σ bonding orbital. In contrast, the bond length elongates by 0.15 Å in the E state, a result of promoting an electron from the 5σ bonding orbital to the 2π antibonding orbitals.

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

    Science.gov (United States)

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

    2016-01-01

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

  12. Calculation of neutron activation of control rods of a nuclear reactor, using MCNP5

    International Nuclear Information System (INIS)

    Pena V, J.D.

    2016-01-01

    The control rods of a nuclear reactor are activated by neutron irradiation. The generated activity produces a dose around the rod which is irrelevant inside the reactor, but significant when the rod is withdrawn and placed in a storage pool, because this dose is a potential risk to the surrounding personnel. On the other hand, most of the activation occurs in the stainless steel components of the rod. The Monte Carlo model can reliably determine the activation produced in a stainless steel part exposed to a neutron flux in a reactor and the dose measurement around this part. This thesis presents the Monte Carlo models developed for the activation of the control rods of the TRIGA Mark III reactor of Instituto Nacional de Investigaciones Nucleares (ININ) when only standard fuel was available. Therefore, the validations of the Monte Carlo models are reliable. (Author)

  13. The effect on cross sections for Quad Cities by introducing control rod history in the assembly program LEWARD

    International Nuclear Information System (INIS)

    Nonboel, E.

    1989-08-01

    This paper shows the effect on the 2-group neutron cross sections for the BWR reactor Quad Cities by introducing control rod ''history'' in the assembly program LEWARD. Control rod ''history'' is a concept which accounts for the movement of control rods during operation. (author)

  14. Polynomial curve fitting for control rod worth using least square numerical analysis

    International Nuclear Information System (INIS)

    Muhammad Husamuddin Abdul Khalil; Mark Dennis Usang; Julia Abdul Karim; Mohd Amin Sharifuldin Salleh

    2012-01-01

    RTP must have sufficient excess reactivity to compensate the negative reactivity feedback effects such as those caused by the fuel temperature and power defects of reactivity, fuel burn-up and to allow full power operation for predetermined period of time. To compensate this excess reactivity, it is necessary to introduce an amount of negative reactivity by adjusting or controlling the control rods at will. Control rod worth depends largely upon the value of the neutron flux at the location of the rod and reflected by a polynomial curve. Purpose of this paper is to rule out the polynomial curve fitting using least square numerical techniques via MATLAB compatible language. (author)

  15. Temperature and Stresses Estimation in Reactivity Control Rods for CAREM-25 Reactor

    International Nuclear Information System (INIS)

    Markiewicz, Mario; Estevez, Esteban

    2000-01-01

    The reactivity control rods are a critical component regarding safety.Its correct operation when required must be ensured.For this purpose, this component must maintain its operating capacity during all its residence time and under any foreseen operation condition.To evaluate the behaviour of reactivity control rods, it is necessary to analyse the demands they are exposed to, determining from the mechanical point of view, the residence time in the reactor core.In this report, using analytical calculations, the parameters affecting the performance of the reactivity control rods are analysed, with the objective of determine from the mechanical point of view, its behaviour and residence time

  16. Calculation of optimum control rod operation programme for boiling water reactor

    International Nuclear Information System (INIS)

    Fehr, L.

    1978-01-01

    Control rod operation programmes are calculated based on a three dimensional Boiling Water Reactor situation model. The position of the control rods at variosu burn-ups is chosen by an optimisation so that the sum of the square deviations of the load density distribution from an optimum distribution ('Haling' distribution) are minimised. Other conditions are remaining critical and observing the thermal limits for central fuel element melting and critical heat surface loading. As an example, an optimum control rod operation programme for the first cycle in Lengen nuclear power station is calculated and is compared with the programme actually used. (orig.) 891 HP [de

  17. Comparison of cell homogenization methods considering interaction effect between fuel cells and control rod cells

    International Nuclear Information System (INIS)

    Takeda, T.; Uto, N.

    1988-01-01

    Several methods to determine cell-averaged group cross sections and anisotropic diffusion coefficients which consider the interaction effect between core fuel cells and control rods or control rod followers have been compared to discuss the physical meaning included in cell homogenization. As the cell homogenization methods considered are the commonly used flux-weighting method, the reaction rate preservation method and the reactivity preservation method. These homogenization methods have been applied to control rod worth calculations in 1-D slab cores to investigate their applicability. (author). 6 refs, 2 figs, 9 tabs

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

    Energy Technology Data Exchange (ETDEWEB)

    Rowsell, David Leon [Idaho National Laboratory (INL), Idaho Falls, ID (United States)

    2015-06-01

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

  19. Seismic appraisal test of control rod drive mechanism of China experiment fast reactor

    International Nuclear Information System (INIS)

    Song Qing; Yang Hongyi; Jing Yueqing; Wen Jing; Liu Guijuan; Sun Lei

    2008-01-01

    The structure of the control rod drive mechanism in pool type sodium-cooled fast reactor is the characterized by long, thin, and geometric nonlinearity, and the seismic load is multiple activation. The anti-seismic evaluation is always paid great attention by the countries developing the technology worldwide. This article introduces the seismic appraisal test of the control rod drive mechanism of China Experimental Fast Reactor (CEFR) performed on a seismic platform which is vertical shaft style and multiple activation. The result of the test shows the structural integrity and the function of the control rod drive mechanism could meet the design requirements of the earthquake intensity. (authors)

  20. Failure Mode and Effects Analysis (FMEA) of the solid state full length rod control system

    International Nuclear Information System (INIS)

    Shopsky, W.E.

    1977-01-01

    The Full Length Rod Control System (FLRCS) controls the power to the rod drive mechanisms for rod movement in response to signals received from the Reactor Control System or from signals generated through Reactor Operator action. Rod movement is used to control reactivity of the reactor during plant operation. The Full Length Rod Control System is designed to perform its reactivity control function in conjunction with the Reactor Control and Protection System, to maintain the reactor core within design safety limits. By the use of a Failure Mode and Effects Analysis, it is shown that the FLRCS will perform its reactivity control functions considering the loss of single active components. That is, sufficient fault limiting control circuits are provided which blocks control rod movement and/or indicates presence of a fault condition at the Control Board. Reactor operator action or automatic reactor trip will thus mitigate the consequences of potential failure of the FLRCS. The analysis also qualitatively demonstrates the reliability of the FLRCS to perform its intended function