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Sample records for rod une methode

  1. Measurement of the anti reactivity of a control rod of G1, by a slow oscillation method; Mesure de l'antireactivite d'une barre de reglage de G1 pour une methode d'oscillation lente

    Energy Technology Data Exchange (ETDEWEB)

    Breton, D; Leroy, J; Vidal, R [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1957-07-01

    It is possible to determine the effect of the end of a control rod on the reactivity of the pile by measuring the modulation induced in the neutron flux by the slow oscillation of this control rod. The total effect of the control rod can be deduced, given certain hypothesis and corrections, from the experimental curve giving the effect of the end of the rod as a function of its position. This method has the advantage of permitting the measurement of very large anti reactivities, such as p= 10{sup -2} for example, which would not be possible by other kinetic methods. Thus the control rod B{sub 3}, in the low position, brings about a reduction in reactivity equal to 1130 p.c.m. {+-} 30 in the pile charged with 518 fuel elements, on one side only of the slit. We have compared the oscillation method with the classical divergence method, in the fields where the two measurements were possible: a satisfactory agreement was found. We have established that the phase displacement between the oscillation of the rod and the modulation of the flux varied greatly with the position of the rod. This variation cannot be explained on the basis of the dynamic model independent of space; we have attributed it to the influence of spatial harmonics of the flux distribution, and have determined a correction which frees the measurements of this influence. (author) [French] II est possible de determiner l'effet de l'extremite d'une barre de reglage sur la reactivite de la pile, a partir de la mesure de la modulation induite dans le flux neutronique par l'oscillation lente de cette barre de reglage. L'effet total de la barre de reglage peut etre deduit, moyennant certaines hypotheses et certaines corrections, de la courbe experimentale donnant l'effet de l'extremite de la barre en fonction de sa position. Cette methode a l'avantage de rendre possible la mesure d'antireactivites tres grandes, telles que p = 10{sup -2} par exemple, ce qui ne serait pas possible par d'autres methodes

  2. Method of inserting fuel rod

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  3. A method for the quantitative determination of uranium-233 in an irradiated thorium rod; Une methode de dosage de l'uranium 233 contenu dans un barreau de thorium irradie

    Energy Technology Data Exchange (ETDEWEB)

    Bathellier, A; Sontag, R; Chesne, A [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

    1961-07-01

    A rapid method for the quantitative determination of uranium-233 in irradiated thorium is described. A 30 per cent solution of trilaurylamine in xylene is used to extract the uranium from an aqueous hydrochloric acid solution and separate it from the thorium. This may be followed by {alpha} counting or fluorimetry. The practical operating conditions of the separation are discussed in detail. (author) [French] Une methode rapide de dosage de l'uranium-233 contenu dans le thorium irradie est decrite. Elle utilise la trilauryfamine a 30 pour cent dans le xylene pour extraire l'uranium d'une dissolution aqueuse chlorhydrique et le separer du thorium. Le comptage {alpha} ou la fluorimetrie sont alors possibles. Les conditions operatoires de la separation sont discutees et precisees. (auteur)

  4. Inspecting method for fuel rods

    International Nuclear Information System (INIS)

    Watanabe, Masaaki; Kogure, Sumio.

    1976-01-01

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

  5. 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.)

  6. 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

  7. Determination of stresses in a sheath connected to a rod; Determination des contraintes dans une gaine liee a un barreau

    Energy Technology Data Exchange (ETDEWEB)

    Bernard, J L; Gauthron, M [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1957-07-01

    Uranium rods introduced into a pile must be protected by impermeable metal canning. Depending on the type of canning adopted, it can be assumed that in certain cases there is a longitudinally rigid connection between this tubular sheath and the rod, either along the whole length (a threaded rod for example) or only at the ends. Even unintentional points of contact, of mechanical or physico-chemical origin, can sometimes be produced accidentally. During pile operation and the resulting variable thermal cycles, the rod and the canning will tend to expand each according to its own expansion law. Given the respective surfaces of rod and canning involved in a cross-section, and the mechanical properties of the two materials considered, it can be legitimately supposed that the canned rod will follow the expansion law of uranium. It follows that the canning, always compelled to follow the expansions of the rod, will be subject to stresses and this study is aimed at their determination. (author) [French] Les barreaux d'uranium introduits dans une pile doivent etre proteges par une gaine metallique etanche. Suivant le mode de gainage adopte, on peut admettre qu'il existe dans certains cas une liaison rigide dans le sens longitudinal entre cette gaine tubulaire et le barreau, sur la totalite de leur longueur (barreau filete par exemple) ou simplement a leurs extremites. Meme s'ils n'ont pas ete provoques, des points d'accrochage, d' origine mecanique ou physico-chimique, peuvent accidentellement nous ramener parfois a ce cas. Lors du fonctionnement de la pile et des cycles thermiques variables qui en resultent, le barreau et la gaine vont tendre a se dilater chacun suivant sa loi de dilatation propre. Etant donne les sections respectives de barreau et de gaine mises en jeu dans une section droite, et les caracteristiques mecaniques des deux materiaux consideres, on peut legitimement admettre que le barreau gaine va suivre la loi de dilatation de l'uranium. Il s'ensuit que la

  8. Method for compacting spent nuclear reactor fuel rods

    International Nuclear Information System (INIS)

    Wachter, W.J.

    1988-01-01

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

  9. Fuel rod failure detection method and system

    International Nuclear Information System (INIS)

    Assmann, H.; Janson, W.; Stehle, H.; Wahode, P.

    1975-01-01

    The inventor claims a method for the detection of a defective fuel rod cladding tube or of inleaked water in the cladding tube of a fuel rod in the fuel assembly of a pressurized-water reactor. The fuel assembly is not disassembled but examined as a whole. In the examination, the cladding tube is heated near one of its two end plugs, e.g. with an attached high-frequency inductor. The water contained in the cladding tube evaporates, and steam bubbles or a condensate are detected by the ultrasonic impulse-echo method. It is also possible to measure the delay of the temperature rise at the end plug or to determine the cooling energy required to keep the end plug temperature stable and thus to detect water ingression. (DG/AK) [de

  10. Method and apparatus for inspection of nuclear fuel rods

    International Nuclear Information System (INIS)

    Wachter, W.J.

    1977-01-01

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

  11. Method and apparatus for compacting spent nuclear reactor fuel rods

    International Nuclear Information System (INIS)

    Wachter, W.J.

    1988-01-01

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

  12. 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

  13. 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

  14. Method for wrapping a wire round a nuclear fuel rod

    International Nuclear Information System (INIS)

    Nakayasu, Fumio.

    1974-01-01

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

  15. 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.)

  16. 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.

  17. 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.)

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

    Science.gov (United States)

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

    2017-11-01

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

  19. Motion simulation of hydraulic driven safety rod using FSI method

    International Nuclear Information System (INIS)

    Jung, Jaeho; Kim, Sanghaun; Yoo, Yeonsik; Cho, Yeonggarp; Kim, Jong In

    2013-01-01

    Hydraulic driven safety rod which is one of them is being developed by Division for Reactor Mechanical Engineering, KAERI. In this paper the motion of this rod is simulated by fluid structure interaction (FSI) method before manufacturing for design verification and pump sizing. A newly designed hydraulic driven safety rod which is one of reactivity control mechanism is simulated using FSI method for design verification and pump sizing. The simulation is done in CFD domain with UDF. The pressure drop is changed slightly by flow rates. It means that the pressure drop is mainly determined by weight of moving part. The simulated velocity of piston is linearly proportional to flow rates so the pump can be sized easily according to the rising and drop time requirement of the safety rod using the simulation results

  20. 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.)

  1. Method of manufacturing nuclear fuel rods

    International Nuclear Information System (INIS)

    Sato, Masao; Oyama, Masatoshi; Yamamoto, Takanobu.

    1976-01-01

    Object: To discriminate the properties of light white deposits on a clad tube during the process of manufacturing nuclear fuel rods and then remove this to reproduce a good clad tube, thereby enhancing a yield of the clad tube. Structure: When a light white deposits is found to be appeared on outer or inner surface of coating during the process of appearance inspection, this is then permitted to subject to treatment of hot water immersion and discrimination. Requirements for removal of adhered matter in the process of treatment of hot water immersion are that deioned water of specific resistance 5 x 10 5 ohms or more is used with water temperature maintained at 60 to 100 0 C for immersion treatment for 10 to 30 minutes. In this case, however, if the water temperature is more than 80 0 C, the immersion time can be set less than 10 minutes. With the addition of such process described above, about 2.5% of total receiving number can be reproduced. (Yoshihara, H.)

  2. 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

  3. Fuel rod design by statistical methods for MOX fuel

    International Nuclear Information System (INIS)

    Heins, L.; Landskron, H.

    2000-01-01

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

  4. 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.)

  5. 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

  6. A probabilistic design method for LMFBR fuel rods

    International Nuclear Information System (INIS)

    Peck, S.O.; Lovejoy, W.S.

    1977-01-01

    Fuel rod performance analyses for design purposes are dependent upon material properties, dimensions, and loads that are statistical in nature. Conventional design practice accounts for the uncertainties in relevant parameters by designing to a 'safety factor', set so as to assure safe operation. Arbitrary assignment of these safety factors, based upon a number of 'worst case' assumptions, may result in costly over-design. Probabilistic design methods provide a systematic way to reflect the uncertainties in design parameters. PECS-III is a computer code which employs Monte Carlo techniques to generate the probability density and distribution functions for time-to-failure and cumulative damage for sealed plenum LMFBR fuel rods on a single rod or whole core basis. In Monte Carlo analyses, a deterministic model (that maps single-valued inputs into single-valued outputs) is coupled to a statistical 'driver'. Uncertainties in the input are reflected by assigning probability densities to the input parameters. Dependent input variables are considered multivariate normal. Independent input variables may be arbitrarily distributed. Sample values are drawn from these input densities, and a complete analysis is done by the deterministic model to generate a sample point in the output distribution. This process is repeated many times, and the number of times each output value occurs is accumulated. The probability that some measure of rod performance will fall within given limits is estimated by the relative frequency with which the Monte Carlo samples fall within tho

  7. Production method of {alpha} particles; Une methode de production des particules {alpha}

    Energy Technology Data Exchange (ETDEWEB)

    Prevot, F [Commissariat a l' Energie Atomique, Saclay(France). Centre d' Etudes Nucleaires

    1953-07-01

    It is proposed a method to get an intense beam of {alpha} particles. With a source of ordinary ions, we form a helium beam, once ionized, it is accelerated with an energy of a few hundreds of keV. While crossing a matter any that can be a thin leaf or a gaseous blade, the second electron of helium is pulled with a yield that only depends on the energy of the beam of helium and that is equal to 1/2 for 650 keV. (author) [French] Il est propose une methode pour obtenir un faisceau intense de particules {alpha}. Avec une source d'ions ordinaire, on forme un faisceau d'helium une fois ionise qu'on accelere avec une energie de quelques centaines de keV. En traversant une matiere quelconque qui peut etre sous forme de feuille mince ou de lame gazeuse, le deuxieme electron de l'helium est arrache avec un rendement qui ne depend que de l'energie du faisceau d'helium et qui vaut 1/2 pour 650 keV. (auteur)

  8. 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.

  9. 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

  10. Method for determining detailed rod worth profiles at low power in the fast test reactor

    International Nuclear Information System (INIS)

    Sevenich, R.A.

    1975-08-01

    A method for obtaining a detailed rod worth profile at low power for a slow control rod insertion is presented. The accuracy of the method depends on a preparatory experiment in which the test rod is dropped quickly to yield, upon analysis, the magnitude of the rod worth and an effective source value. These numbers are employed to initialize the inverse kinetics analysis for the slow insertion. Corrections for changes in detection efficiency are not included for the simulated experiments. (U.S.)

  11. 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.

  12. 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.)

  13. Method for verifying the pressure in a nuclear reactor fuel rod

    International Nuclear Information System (INIS)

    Jones, W.J.

    1979-01-01

    Disclosed is a method of accurately verifying the pressure contained in a sealed pressurized fuel rod by utilizing a pressure balance measurement technique wherein an end of the fuel rod extends through and is sealed in a wall of a small chamber. The chamber is pressurized to the nominal (desired) fuel rod pressure and the fuel rod is then pierced to interconnect the chamber and fuel rod. The deviation of chamber pressure is noted. The final combined pressure of the fuel rod and drill chamber is substantially equal to the nominal rod pressure; departure of the combined pressure from nominal is in direct proportion to departure of rod pressure from nominal. The maximum error in computing the rod pressure from the deviation of the combined pressure from nominal is estimated at plus or minus 3.0 psig for rod pressures within the specified production limits. If the rod pressure is corrected for rod void volume using a digital printer data record, the accuracy improves to about plus or minus 2.0 psig

  14. 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

  15. Failed fuel rod detection method by ultrasonic wave

    International Nuclear Information System (INIS)

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

    1990-01-01

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

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

    Science.gov (United States)

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

    2011-11-01

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

  17. Critical heat flux detection in rods simulating fuel elements by using dilation method

    International Nuclear Information System (INIS)

    Mesquita, A.Z.

    1993-01-01

    In out-reactor heat transfer experiments, fuel elements are often simulated by electrically heated rods. In order to prevent the heating rod from being damaged by burnout, when the critical heat flux occurs a safety system is provided which checks the axial thermal expansion of the rod. In case of sudden temperature increase, the corresponding elongation causes a fast interruption of the electrical power supply. The experiments presented here show that this method is more effective than one that uses thermocouples. (author)

  18. 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.)

  19. A method for designing fiberglass sucker-rod strings with API RP 11L

    International Nuclear Information System (INIS)

    Jennings, J.W.; Laine, R.E.

    1991-01-01

    This paper presents a method for using the API recommended practice for the design of sucker-rod pumping systems with fiberglass composite rod strings. The API method is useful for obtaining quick, approximate, preliminary design calculations. Equations for calculating all the composite material factors needed in the API calculations are given

  20. Contribution au developpement d'une methode de controle des procedes dans une usine de bouletage

    Science.gov (United States)

    Gosselin, Claude

    overall (economic) satisfaction in the production process, but rather in proposing it as an "observer" of the system's state. The first implementation steps have already demonstrated the method's feasibility as well as other numerous industrial impacts on production processes within the company. Namely, the emergence of the economical aspect as a strategic variable that assures better governance of production processes where quality variables present strategic issues.

  1. 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

  2. Power ramp testing method for PWR fuel rod at research reactor

    International Nuclear Information System (INIS)

    Zhou Yidong; Zhang Peisheng; Zhang Aimin; Gao Yongguang; Wang Huarong

    2003-01-01

    A tentative power ramp test for short PWR fuel rod has been conducted at the Heavy Water Research Reactor (HWRR) in China Institute of Atomic Energy (CIAE). The test fuel rod was cooled by the circulating water in the test loop. The power ramp was realized by moving solid neutron-absorbing screen around the fuel rod. The linear power of the fuel rod increased from 220 W/cm to 340 W/cm with a power ramp rate of 20 W/cm/min. The power of the fuel rod was monitored by both in-core thermal and nuclear measurement sensors in the test rig. This test provides experiences for further developing the power ramp test methods for PWR fuel rods at research reactor. (author)

  3. 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

  4. Nuclear fuel rod helium leak inspection apparatus and method

    International Nuclear Information System (INIS)

    Ahmed, H.J.

    1991-01-01

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

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

    International Nuclear Information System (INIS)

    Walton, L.A.

    1981-01-01

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

  6. 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

  7. Study on dynamic rod worth measurement method and its test verification

    International Nuclear Information System (INIS)

    Wu Lei; Liu Tongxian; Zhao Wenbo; Li Songling; Yu Yingrui

    2015-01-01

    An advanced rod worth measurement technique, the dynamic rod worth measurement method (DRWM) has been developed. Static Spatial Factors (SSF) and Dynamic Spatial Factor (DSF) were introduced to improve the inverse kinetics method. The three dimensional steady and transient simulations for the measurement process was carried out to calculate the modification factors. The rod worth measurement, test was performed on a research reactor to verify DRWM. The results showed that the DRWM method provided the improved accuracy and could be a replacement of the traditional methods. (authors)

  8. System and method for consolidating spent fuel rods

    International Nuclear Information System (INIS)

    Baudro, T.O.

    1987-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-08-01

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

  10. Tabular method of critical heat flux description in square packing rod bundles

    International Nuclear Information System (INIS)

    Bobkov, V.P.; Smogalev, I.P.

    2003-01-01

    Elaborations of harnessing tabular method for the description and calculation of critical heat fluxes in square packing rod bundles are presented. The tabular method for fuel rod triangular assemblies derived from using basic table for critical heat fluxes in triangular fuel assemblies demonstrates good results. For the harnessing tabular method in square packing rod bundles correction functions reflecting specific geometry were found. Comparative evaluations of calculated values for the critical heat fluxes with experimental ones are presented. Good agreement of calculations with experiments is noted in all range of parameters [ru

  11. Fuel Rod Vibration Measurement Method using a Flap and its Verification

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-10-15

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

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

    International Nuclear Information System (INIS)

    Kalcheva, Silva; Koonen, Edgar

    2008-01-01

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

  13. Practical use of control rod calibration system with the inverse kinetics method

    International Nuclear Information System (INIS)

    Yamanaka, Haruhiko; Hayashi, Kazuhiko; Motohashi, Jun; Kawashima, Kazuhito; Ichimura, Toshiyuki; Tamai, Kazuo; Takeuti, Mitsuo

    2002-01-01

    The control rod calibration results in the JRR-3 are used as a reactivity standard to measure and manage the reactivity change in the core. The total travel of all six control rods has been calibrated by an inverse kinetics method (IK method) during an annual maintenance period. The IK method has the great merit in saving measuring time compared with the conventional positive period method (PP method). The JRR-3 control rod calibration system was renovated and put into practical use in order to improve reliability and function by accumulating 10-year experience with the IK method in the JRR-3. The report shows the function, the performance and results of verification of the JRR-3 control rod calibration system. (author)

  14. 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.)

  15. Apparatus and method for preventing the rotation of rods used in nuclear fuel assemblies

    International Nuclear Information System (INIS)

    Pilgrim, L.G. Jr.; Jackson, L.F.

    1985-01-01

    Apparatus and method for preventing the rotation of one or more elongated rods used in nuclear fuel assemblies include an end plug secured to one longitudinal end of such an elongated rod and having an out-of-cavity, non-round structure affixed thereto and configured to mate with a complementary shaped structure in a lower tie plate of a nuclear fuel assembly in such a manner as to prevent the rotation of the rod about its longitudinal axis. In one embodiment, the end plug includes a pair of flats formed on a portion of the end plug and configured to abut against a pair of flats formed on the outer surface of a cylindrical boss or sleeve of the lower tie plate, thereby to prevent the rotation of the rod. In another embodiment, four grooves, disposed 90 0 apart about the periphery of an end plug of a rod form a spline. The grooves are configured to receive four, radially inwardly protruding, key members disposed 90 0 apart about the periphery of a sleeve secured to the lower tie plate, thereby to prevent the rotation of the rod. In a further embodiment, a sleeve is secured to an end plug of a rod and includes four elongated slots disposed 90 0 apart about the periphery of the sleeve and configured in width, depth and spacing to receive and mate with four web portions of the lower tie plate of the nuclear fuel assembly, thereby to secure the rod against rotation about its longitudinal axis

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

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  17. Determination of the most reactivity control rod by pseudo-harmonics perturbation method

    International Nuclear Information System (INIS)

    Freire, Fernando S.; Silva, Fernando C.; Martinez, Aquilino S.

    2005-01-01

    Frequently it is necessary to compute the change in core multiplication caused by a change in the core temperature or composition. Even when this perturbation is localized, such as a control rod inserted into the core, one does not have to repeat the original criticality calculation, but instead we can use the well-known pseudo-harmonics perturbation method to express the corresponding change in the multiplication factor in terms of the neutron flux expanded in the basis vectors characterizing the unperturbed core. Therefore we may compute the control rod worth to find the most reactivity control rod to calculate the fast shutdown margin. In this thesis we propose a simple and precise method to identify the most reactivity control rod. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-01-15

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

  19. 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.)

  20. Method and apparatus for sizing nuclear fuel rod cladding tubes

    International Nuclear Information System (INIS)

    Koehler, L.

    1976-01-01

    Nuclear fuel rod cladding tubes are sized internally to diameters precisely fitting nuclear fuel pellets with which the tubes are charged by externally applying hydraulic pressure to short lengths of each tube. The pressure is applied while the tube is stationary. The tube is then moved to bring a new length within the hydraulic pressure zone. The volume of the hydraulic liquid used and the pressure applied to this liquid is such that the liquid is compressed slightly so that the length being sized yields, the expansion of the liquid then completing the sizing. The lengths being sized step-by-step are internally supported by either the fuel pellets or a mandrel having the same diameter as the pellets

  1. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

    Bjoereus, K

    1969-07-01

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

  3. 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.)

  4. Gas sealing welding method and device for nuclear fuel rod

    International Nuclear Information System (INIS)

    Seki, Masayuki; Nishiyama, Motokuni; Kamimura, Katsuichiro; Yagi, Eiji; Nakase, Tsuyoshi; Kobogata, Sadao; Taniguchi, Jun-ichi; Uesugi, Yoshisaku.

    1995-01-01

    An end plug and a cladding tube are held by clamping, respectively, by opposing movable electrode and static electrode. The movable electrode is forwarded toward the static electrode. The end plug and the cladding tube are abutted and held at a slight gap between their end faces. A region to be welded is surrounded by a pressurizing chamber and the side of the chamber is evacuated and He gas is filled in the cladding tube. Then, one of the electrodes is forwarded, to seal the abutted end faces of the end plug and the cladding tube. Then, pressure and welding current required for welding are applied to the abutted ends, and He gas is sealed in the vessel. The displacement of pressurization caused by slipping when the required pressure is applied to the abutted ends is detected by a sensor, and the operation of the welding control device for starting current supply is terminated by the detection signals. Abutment accuracy between the abutment of the cladding tube and the end plug as a nuclear fuel rod can be ensured, to further improve and stabilize the welding quality. (N.H.)

  5. End plug for fuel rod and welding method therefor

    International Nuclear Information System (INIS)

    Yoneda, Hiroshi; Murakami, Kazuo; Oyama, Jun-ichi.

    1996-01-01

    An end plug of a fuel rod comprises a pressure-insertion portion having a diameter somewhat greater than the inner diameter of a fuel cladding tube and a welding portion having a diameter substantially the same as the outer diameter of the cladding tube. A V-shaped recess having an outer diameter smaller than the greatest outer diameter of the pressure-insertion portion is formed over the entire circumferential surface of the outer circumference of the connection portion of the pressure-insertion portion and the welding portion. The pressure-insertion portion of the end plug is inserted to the end of the cladding tube till the end of the cladding tube abuts against the inclined surface of the V-shaped recess. The abutting surfaces of the end plug and the cladding tube are subjected to resistance welding in this state. The inner portion bulged from the inclined surface of the V-shaped recess is filled in the recess in a molten state. Lowering of temperature of the cladding tube in the vicinity of the welded portion is decreased by γ heat during reactor operation. Accordingly, lowering of ductility of the cladding tube and degradation of material of the welded region due to segregation of hydrogen in the cladding tube can be suppressed. (I.N.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1986-05-01

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

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

    International Nuclear Information System (INIS)

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

    1986-05-01

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

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

    International Nuclear Information System (INIS)

    Arimescu, V.E.; Heins, L.

    2001-01-01

    Advances in modeling fuel rod behavior and accumulations of adequate experimental data have made possible the introduction of quantitative methods to estimate the uncertainty of predictions made with best-estimate fuel rod codes. The uncertainty range of the input variables is characterized by a truncated distribution which is typically a normal, lognormal, or uniform distribution. While the distribution for fabrication parameters is defined to cover the design or fabrication tolerances, the distribution of modeling parameters is inferred from the experimental database consisting of separate effects tests and global tests. The final step of the methodology uses a Monte Carlo type of random sampling of all relevant input variables and performs best-estimate code calculations to propagate these uncertainties in order to evaluate the uncertainty range of outputs of interest for design analysis, such as internal rod pressure and fuel centerline temperature. The statistical method underlying this Monte Carlo sampling is non-parametric order statistics, which is perfectly suited to evaluate quantiles of populations with unknown distribution. The application of this method is straightforward in the case of one single fuel rod, when a 95/95 statement is applicable: 'with a probability of 95% and confidence level of 95% the values of output of interest are below a certain value'. Therefore, the 0.95-quantile is estimated for the distribution of all possible values of one fuel rod with a statistical confidence of 95%. On the other hand, a more elaborate procedure is required if all the fuel rods in the core are being analyzed. In this case, the aim is to evaluate the following global statement: with 95% confidence level, the expected number of fuel rods which are not exceeding a certain value is all the fuel rods in the core except only a few fuel rods. In both cases, the thresholds determined by the analysis should be below the safety acceptable design limit. An indirect

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

    International Nuclear Information System (INIS)

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

    1988-12-01

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

  10. Recovery of Americium-241 from lightning rod by the method of chemical treatment

    International Nuclear Information System (INIS)

    Cruz, W.H.

    2013-01-01

    About 95% of the lightning rods installed in the Peruvian territory have set in their structures, pose small amounts of radioactive sources such as Americium-241 ( 241 Am), fewer and Radium 226 ( 226 Ra) these are alpha emitters and have a half life of 432 years and 1600 years respectively. In this paper describes the recovery of radioactive sources of 241 Am radioactive lightning rods using the conventional chemical treatment method using agents and acids to break down the slides. The 241 Am recovered was as excitation source and alpha particle generator for analysing samples by X Ray Fluorescence, for fixing the stainless steel 241 Am technique was used electrodeposition. (author)

  11. A method of sub-critical experimentation, 'the neutrostat'; Une methode d'experimentation sous critique 'le neutrostat'

    Energy Technology Data Exchange (ETDEWEB)

    Martelly, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1958-07-01

    The method proposed is designed for the study of neutronic properties of a sample (its material buckling, for example) and consists in submitting the sample to uniform surface density irradiation, its surface being a sphere or a cylinder which is supposed, temporarily, to be infinite. Neutron density in the sample will thus be uniform if its laplacian is nil: any curve in the distribution clearly indicating its absorbent or multiplying properties. In the case of a sample with multiplying power, density is identical to that in an active core, thus measurement of buckling will be free from considerable systematic error causes. The thermic equivalent of this type of irradiation would be a thermostat with an external heat source distributed uniformly over its surface: its temperature would be uniform. It is this analogy that has led us to baptize it the 'Neutrostat'. (author)Fren. [French] En vue d'etudier les proprietes neutroniques d'un milieu (son laplacien 'matiere' par exemple), la methode proposee consiste a le soumettre a des conditions d'irradiation uniforme sur sa surface, celle-ci ayant la forme d'une sphere ou d'un cylindre que nous supposons provisoirement infini. Les neutrons s'y trouvent alors repartis avec une densite uniforme si le milieu est un diffuseur pur. Toute courbure de cette repartition sera un indice sensible de ses proprietes absorbantes ou multiplicatrices. Dans le cas d'un milieu multiplicateur, la repartition est identique a celle qui regne au milieu d'une pile critique et nous verrons que la mesure du laplacien est alors exempte de causes d'erreurs systematiques importantes. L'equivalent thermique d'un tel mode d'irradiation serait un dispositif thermostatique dont la source de chaleur externe serait repartie uniformement sur la surface: il y regnerait une temperature uniforme. C'est cette analogie qui nous a guides dans le choix du vocable propose dans le titre: 'Neutrostat'. (auteur)

  12. The probabilistic method of WWER fuel rod strength estimation using the START-3 code

    International Nuclear Information System (INIS)

    Bibilashvili, Yu.K.; Medvedev, A.V.; Bogatyr, S.M.; Sokolov, F.F.; Khramtsov, M.V.

    2001-01-01

    During the last years probability methods of studying were widely used to determine the influence exerted by the geometry, technology and performance parameters of a fuel rod on the characteristics of its condition. Despite the diversity of probability methods their basis is formed by the simplest schema of the Monte-Carlo method (MC). This schema assumes a great number of the realizations of a random value and the statistical assessment of its characteristics. To generate random values, use is usually made of a pseudo-random number generator. The application of the quasi-random sequence elements in place of the latter substantially reduces the machine time since it promotes a quicker convergence of the method. Probability methods used to study the characteristics of a fuel rod condition can be considered to be an auxiliary means of deterministic calculations that allows the assessment of the conservatism degree of design calculations. (author)

  13. Evaluation of the rod ejection accident in Westinghouse Pressurized Water Reactors using spatial kinetics methods

    International Nuclear Information System (INIS)

    Risher, D.H. Jr.

    1975-01-01

    The consequences of a rod ejection accident are investigated in relation to the latest, high power density Westinghouse reactors. Limiting criteria are presented, based on experimental evidence, and if not exceeded these criteria will ensure that there will be no interference with core cooling capability, and radiation releases, if any, will be within the guidelines of 10CFR100. A basis is presented for the conservative selection of plant parameters to be used in the analysis, such that the analysis is applicable to a wide range of past, present, and future reactors. The calculational method employs a one-dimensional spatial kinetics computer code and a transient fuel heat transfer computer code to determine the hot spot fuel temperature versus time following a rod ejection. Using these computer codes, the most limiting hot channel factor (which does not cause the fuel damage limit criteria to be exceeded) has been determined as a function of the ejected rod worth. By this means, the limit criteria have been translated into ejected rod worths and hot channel factors which can be used effectively by the nuclear designer and safety analyst. The calculational method is shown to be conservative, compared to the results of a three-dimensional spatial kinetics analysis

  14. Recursive-operator method in vibration problems for rod systems

    Science.gov (United States)

    Rozhkova, E. V.

    2009-12-01

    Using linear differential equations with constant coefficients describing one-dimensional dynamical processes as an example, we show that the solutions of these equations and systems are related to the solution of the corresponding numerical recursion relations and one does not have to compute the roots of the corresponding characteristic equations. The arbitrary functions occurring in the general solution of the homogeneous equations are determined by the initial and boundary conditions or are chosen from various classes of analytic functions. The solutions of the inhomogeneous equations are constructed in the form of integro-differential series acting on the right-hand side of the equation, and the coefficients of the series are determined from the same recursion relations. The convergence of formal solutions as series of a more general recursive-operator construction was proved in [1]. In the special case where the solutions of the equation can be represented in separated variables, the power series can be effectively summed, i.e., expressed in terms of elementary functions, and coincide with the known solutions. In this case, to determine the natural vibration frequencies, one obtains algebraic rather than transcendental equations, which permits exactly determining the imaginary and complex roots of these equations without using the graphic method [2, pp. 448-449]. The correctness of the obtained formulas (differentiation formulas, explicit expressions for the series coefficients, etc.) can be verified directly by appropriate substitutions; therefore, we do not prove them here.

  15. An approach to automated chromosome analysis; Etudes pour une methode d'automatisation des analyses chromosomiques

    Energy Technology Data Exchange (ETDEWEB)

    Le Go, Roland

    1972-05-03

    The methods of approach developed with a view to automatic processing of the different stages of chromosome analysis are described in this study divided into three parts. Part 1 relates the study of automated selection of metaphase spreads, which operates a decision process in order to reject ail the non-pertinent images and keep the good ones. This approach has been achieved by Computing a simulation program that has allowed to establish the proper selection algorithms in order to design a kit of electronic logical units. Part 2 deals with the automatic processing of the morphological study of the chromosome complements in a metaphase: the metaphase photographs are processed by an optical-to-digital converter which extracts the image information and writes it out as a digital data set on a magnetic tape. For one metaphase image this data set includes some 200 000 grey values, encoded according to a 16, 32 or 64 grey-level scale, and is processed by a pattern recognition program isolating the chromosomes and investigating their characteristic features (arm tips, centromere areas), in order to get measurements equivalent to the lengths of the four arms. Part 3 studies a program of automated karyotyping by optimized pairing of human chromosomes. The data are derived from direct digitizing of the arm lengths by means of a BENSON digital reader. The program supplies' 1/ a list of the pairs, 2/ a graphic representation of the pairs so constituted according to their respective lengths and centromeric indexes, and 3/ another BENSON graphic drawing according to the author's own representation of the chromosomes, i.e. crosses with orthogonal arms, each branch being the accurate measurement of the corresponding chromosome arm. This conventionalized karyotype indicates on the last line the really abnormal or non-standard images unpaired by the program, which are of special interest for the biologist. (author) [French] Ce travail expose les methodes d'approche etudiees en vue

  16. Blackness coefficients, effective diffusion parameters, and control rod worths for thermal reactors - Methods

    Energy Technology Data Exchange (ETDEWEB)

    Bretscher, M M [Argonne National Laboratory, Argonne, IL 60439 (United States)

    1985-07-01

    Simple diffusion theory cannot be used to evaluate control rod worths in thermal neutron reactors because of the strongly absorbing character of the control material. However, reliable control rod worths can be obtained within the framework of diffusion theory if the control material is characterized by a set of mesh-dependent effective diffusion parameters. For thin slab absorbers the effective diffusion parameters can be expressed as functions of a suitably-defined pair of 'blackness coefficients'. Methods for calculating these blackness coefficients in the P1, P3, and P5 approximations, with and without scattering, are presented. For control elements whose geometry does not permit a thin slab treatment, other methods are needed for determining the effective diffusion parameters. One such method, based on reaction rate ratios, is discussed. (author)

  17. An algorithm for seeking the optimum value of a function: 'random' method; Un algorithme de recherche de l'optimum d'une fonction: la methode random

    Energy Technology Data Exchange (ETDEWEB)

    Guais, J C [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1967-07-01

    After a brief survey of classical techniques for static optimization, we present a Random seeking method for any function, of an arbitrary number of variables, with constraints. The resulting program is shown and illustrated by some examples. The comparison with classical methods points out the advantages of Random in some cases where analytic procedures fail or require too much calculation time. (author) [French] Apres une rapide revue des differents procedes actuels d'optimisation statique, on expose une methode de recherche aleatoire du minimum (ou du maximum) d'une fonction quelconque, definie sur un nombre theoriquement illimite de parametres independants, avec contraintes. Le programme resultant est presente. Il est illustre par quelques exemples simples et compare a des methodes d'optimisation classiques; Ceci montre en particulier que le programme RANDOM permet une recherche aisee d'extrema dans certains cas ou d'autres programmes ne conduisent pas a des solutions satisfaisantes ou bien demandent un temps calcul prohibitif. (auteur)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-01-01

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

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

    International Nuclear Information System (INIS)

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

    1992-01-01

    This paper reports on a two-step method that is established for the generation of a long-term control rod program for boiling water reactors (BWRs). The new method assumes a time-variant target power distribution in core depletion. In the new method, the BWR control rod programming is divided into two steps. In step 1, a sequence of optimal, exposure-dependent Haling power distribution profiles is generated, utilizing the spectral shift concept. In step 2, a set of exposure-dependent control rod patterns is developed by using the Haling profiles generated at step 1 as a target. The new method is implemented in a computer program named OCTOPUS. The optimization procedure of OCTOPUS is based on the method of approximation programming, in which the SIMULATE-E code is used to determine the nucleonics characteristics of the reactor core state. In a test in cycle length over a time-invariant, target Haling power distribution case because of a moderate application of spectral shift. No thermal limits of the core were violated. The gain in cycle length could be increased further by broadening the extent of the spetral shift

  20. FEM simulation of friction testing method based on combined forward rod-backward can extrusion

    DEFF Research Database (Denmark)

    Nakamura, T; Bay, Niels; Zhang, Z. L

    1997-01-01

    A new friction testing method by combined forward rod-backward can extrusion is proposed in order to evaluate frictional characteristics of lubricants in forging processes. By this method the friction coefficient mu and the friction factor m can be estimated along the container wall and the conical...... curves are obtained by rigid-plastic FEM simulations in a combined forward rod-backward can extrusion process for a reduction in area R-b = 25, 50 and 70 percent in the backward can extrusion. It is confirmed that the friction factor m(p) on the punch nose in the backward cart extrusion has almost...... in a mechanical press with aluminium alloy A6061 as the workpiece material and different kinds of lubricants. They confirm the analysis resulting in reasonable values for the friction coefficient and the friction factor....

  1. Nodal methods for calculating nuclear reactor transients, control rod patterns, and fuel pin powers

    International Nuclear Information System (INIS)

    Cho, Byungoh.

    1990-01-01

    Nodal methods which are used to calculate reactor transients, control rod patterns, and fuel pin powers are investigated. The 3-D nodal code, STORM, has been modified to perform these calculations. Several numerical examples lead to the following conclusions: (1) By employing a thermal leakage-to-absorption ratio (TLAR) approximation for the spatial shape of the thermal fluxes for the 3-D Langenbuch-Maurer-Werner (LMW) and the superprompt critical transient problems, the convergence of the conventional two-group scheme is accelerated. (2) By employing the steepest-ascent hill climbing search with heuristic strategies, Optimum Control Rod Pattern Searcher (OCRPS) is developed for solving control rod positioning problem in BWRs. Using the method of approximation programming the objective function and the nuclear and thermal-hydraulic constraints are modified as heuristic functions that guide the search. The test calculations have demonstrated that, for the first cycle of the Edwin Hatch Unit number-sign 2 reactor, OCRPS shows excellent performance for finding a series of optimum control rod patterns for six burnup steps during the operating cycle. (3) For the modified two-dimensional EPRI-9R problem, the least square second-order polynomial flux expansion method was demonstrated to be computationally about 30 times faster than a fine-mesh finite difference calculation in order to achieve comparable accuracy for pin powers. The basic assumption of this method is that the reconstructed flux can be expressed as a product of an assembly form function and a second-order polynomial function

  2. Optimization method of rod-type burnable poisons for nuclear designs of HTGRs

    International Nuclear Information System (INIS)

    Yamashita, Kiyonobu

    1994-01-01

    In block-type HTGRs, control rod insertion depths into cores had to be maintained as small as possible at full power operations, to avoid a fuel temperature rise. Thus, specifications (poison atom density (N BP ) and radius (r)) of rod-type burnable poisons (BPs) had to be optimized so that the effective multiplication factor (k eff ) would be constant at a minimum value throughout a planned burnup period. However, the optimization had been a time-consuming work until now since survey calculations had to be done for most possible combinations of N BP and r. To solve this problem, I have found a optimization method consisting of two steps. In the first step, approximation formulas describing a time-dependent relation among effective absorption cross sections (Σ aBP ), N BP and r are used to select promising combinations of N BP and r beforehand. In the second step, the best combination of N BP and r is determined by a comparison between Σ aBP of each promising combination and expected one. The number of survey calculations was reduced to about 1/10 by the optimization method. The change in k eff for 600 burnup days was reduced to 2%Δk by the method. Hence, it was made possible to operate reactors practically without inserting the control rods into cores. (author)

  3. Effect of repeated sterilization by different methods on strength of carbon fiber rods used in external fixator systems.

    Science.gov (United States)

    Unal, Omer Kays; Poyanli, Oguz Sukru; Unal, Ulku Sur; Mutlu, Hasan Huseyin; Ozkut, Afsar Timucin; Esenkaya, Irfan

    2018-05-16

    We set out to reveal the effects of repeated sterilization, using different methods, on the carbon fiber rods of external fixator systems. We used a randomized set of forty-four unused, unsterilized, and identical carbon fiber rods (11 × 200 mm), randomly assigned to two groups: unsterilized (US) (4 rods) and sterilized (40 rods). The sterilized rods were divided into two groups, those sterilized in an autoclave (AC) and by hydrogen peroxide (HP). These were further divided into five subgroups based on the number of sterilization repetition to which the fibers were subjected (25-50-75-100-200). A bending test was conducted to measure the maximum bending force (MBF), maximum deflection (MD), flexural strength (FS), maximum bending moment (MBM) and bending rigidity (BR). We also measured the surface roughness of the rods. An increase in the number of sterilization repetition led to a decrease in MBF, MBM, FS, BR, but increased MD and surface roughness (p < 0.01). The effect of the number of sterilization repetition was more prominent in the HP group. This study revealed that the sterilization method and number of sterilization repetition influence the strength of the carbon fiber rods. Increasing the number of sterilization repetition degrades the strength and roughness of the rods.

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

    International Nuclear Information System (INIS)

    Nissen, K.L.

    1988-06-01

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

  5. 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

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

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  7. Development of three methods for control rod position monitoring based on fixed in-core neutron detectors

    International Nuclear Information System (INIS)

    Peng, Xingjie; Li, Qing; Wang, Kan

    2015-01-01

    Highlights: • Three methods are utilized separately to unfold the control rod position from the fixed in-core neutron detector measurements. • Fixed in-core neutron detector measurements are simulated by neutronics code SMART. • Numerical results show that all these methods can unfold the control rod position accurately. • Two correction strategies are proposed to correct the simulated fixed in-core detector signals. - Abstract: Nuclear reactor core power distribution on-line monitoring system is very important in core surveillance, and this system should have the ability to indicate some abnormal conditions, such as the unacceptable control rod misalignment. In this study, the methodologies of radial basis function neural network (RBFNN), group method of data handling (GMDH) and Levenberg–Marquardt (LM) algorithm are utilized separately to unfold the control rod position from the fixed in-core neutron detector measurements. For using these methods, a large number of in-core detector signals corresponding to various known rod positions are needed. These data can be generated by an advanced core calculation code. In this study, the neutronics code SMART was used. The simulation results show that all these methods can unfold the control rod position accurately, and the performance comparison shows that the regularized RBFNN performs best. Two correction strategies are proposed to correct the simulated fixed in-core detector signals and improve the rod position monitoring accuracy when there are mismatches between actual physical factors and modeled physical factors

  8. 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.)

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

    International Nuclear Information System (INIS)

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

    1957-01-01

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

  10. Method of monitoring fuel-rod vibrations in a nuclear fuel reactor

    International Nuclear Information System (INIS)

    Kawamura, Makoto; Takai, Katsuaki.

    1985-01-01

    Purpose: To monitor the vibration modes of fuel rods continuously and on real time during operation of a PWR type nuclear reactor. Method: Vibrations of fuel rods during reactor operation are mainly caused by the lateral flow of coolants flowing through the gaps at the joints of reactor core buffle plates into a reactor core and fretting damages may possibly be caused to the fuel rod support portions due to the vibrations. In view of the above, self-powered detectors are disposed at a plurality of axial positions for the respective peripheral fuel assemblies in adjacent with the buffle plates and the detection signals from neutron detectors, that is, the fluctuations in neutrons are subjected to a frequency analysis during the operation period. The neutron detectors are disposed at the periphery of the reactor core, because the fuel assemblies disposed at the peripheral portion directly undergo the lateral flow from the joints of the buffle plates and vibrates most violently. Thus, the vibration situations can be monitored continuously, in a three demensional manner and on real time. (Moriyama, K.)

  11. Method and apparatus for the production of a nuclear fuel rod

    International Nuclear Information System (INIS)

    Ballard, A.S.; Cooper, R.G.; Davis, D.E.

    1975-01-01

    The method designs the manufacture of e.g. rod-shaped fuel element fillings in which fuel particles are suspended within a liquid and solidifiable binder such as graphite powder in pitch. The fuel particles are filled into cavities whose cross-sections correspond to those of the fuel rods. After closing with a covering plate, a piston exerts a force from below on it until its solidification. To follow, the liquid binder is injected through lower openings in the cavities. Due to the lubricity of the binder, the cavities are heated to 150 to 175 0 C, the packing of particles are homogenized. This procedure is further supported by the constant pressure of the pistons. Excess binder and air can flow out through openings in the covering plate. After cooling and solidification of the binder as well as after removal of the covering plate, the piston thrusts out the formed bodies or fuel rods from the cavities by an upwards movement. (DG/LH) [de

  12. Leak detection device for control rod drive and detection method therefor

    International Nuclear Information System (INIS)

    Imasaki, Yoshio.

    1997-01-01

    The present invention provides a detection device for leak of cooling water from a sealed axial portion of control rod drives (CRD) disposed in a BWR type reactor and a monitoring method therefor. Namely, the CRD transfers rotation at the sealed axial portion and elevates/lowers a piston to insert/withdraw control rod into/from the reactor core. High pressure water is injected upon occurrence of scram to urge the piston upwardly thereby rapidly inserting the control rods. Leak detection pipelines are laid from the sealed axial portion. A flow glass is connected to the leak detection pipelines. Then, cooling water leaked from the sealed axial portion flows in the leak detection pipelines and flows into the flow glass. The flow rate of cooling water leaked from the sealed axial portion of the CRD can thus be detected by monitoring the flow glass. In addition, a flowmeter is connected to the leak detection pipelines, or the flowmeter and the flow glass are connected, and a flowmeter is connected downstream. Then, the flow rate of the leaked cooling water can be detected automatically. (I.S.)

  13. Huitzoctli: A system to design Control Rod Pattern for BWR's using a hybrid method

    International Nuclear Information System (INIS)

    Castillo, Alejandro; Ortiz-Servin, Juan Jose; Perusquia, Raul; Morales, Luis B.

    2011-01-01

    Highlights: → The system was developed to design Control Rod Patterns for Boiling Water Reactors. → The critical reactor core and the thermal limits were fulfilled in all tested cases. → The Fuel Loading Pattern remains without changes during the iterative process. → The system uses the heuristics techniques: Scatter Search and Tabu Search. → The effective multiplication factor k eff at the EOC was improved in all tested cases. - Abstract: Huitzoctli system was developed to design Control Rod Patterns for Boiling Water Reactors (BWR). The main idea is to obtain a Control Rod Pattern under the following considerations: (a) the critical reactor core state is satisfied, (b) the axial power distribution must be adjusted to a target axial power distribution proposal, and (c) the maximum Fraction of Critical Power Ratio (MFLCPR), the maximum Fraction of Linear Power Density (FLPD) and the maximum Fraction of Average Planar Power Density (MPGR) must be fulfilled. Those parameters were obtained using the 3D CM-PRESTO code. In order to decrease the problem complexity, Control Cell Core load strategy was implemented; in the same way, intermediate axial positions and core eighth symmetry were took into account. In this work, the cycle length was divided in 12 burnup steps. The Fuel Loading Pattern is an input data and it remains without changes during the iterative process. The Huitzoctli system was developed to use the combinatorial heuristics techniques Scatter Search and Tabu Search. The first one was used as a global search method and the second one as a local search method. The Control Rod Patterns obtained with the Huitzoctli system were compared to other Control Rod Patterns designs obtained with other optimization techniques, under the same operating conditions. The results show a good performance of the system. In all cases the thermal limits were satisfied, and the axial power distribution was adjusted to the target axial power distribution almost

  14. Development of mechanical analysis module for simulation of SFR fuel rod behavior using finite element method

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Andong; Jeong, Hyedong; Suh, Namduk [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of); Kim, Hyochan; Yang, Yongsik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    Korean SFR developer decided to adapt metal fuel, current study focused on the metal fuel instead of oxide fuel. The SFR metal fuel has been developed by Korea Atomic Energy Research Institute (KAERI) and many efforts focused on designing and manufacturing the metal fuel. Since a nuclear fuel is the first barrier to protect radioactive isotope release, the fuel's integrity must be secured during steady-state operation and accident condition within an acceptable range. Whereas the design and evaluation methodologies, code systems and test procedures of a light water reactor fuel are sufficiently established, those of the SFR fuel needs more technical advances. In the view of regulatory point, there are still many challenging issues which are required to secure the safety of fuel and reactors. For this reason, the Korea Institute of Nuclear Safety (KINS) has launched the new project to develop the regulatory technology for SFR system including a fuel area. The ALFUS code was developed by CRIEPI and employs mechanistic model for fission gas release and swelling of fuel slug. In the code system, a finite element method was introduced to analyze the fuel and cladding's mechanical behaviors. The FEAST code is more advanced code system for SFR which adopted mechanistic FGR and swelling model but still use analytical model to simulate fuel and cladding mechanical behavior. Based on the survey of the previous studies, fuel and cladding mechanical model should be improved. Analysis of mechanical behavior for fuel rod is crucial to evaluate overall rod's integrity. In addition, it is because contact between fuel slug and cladding or an over-pressure of rod internal pressure can cause rod failure during steady-state and other operation condition. The most of reference codes have simplified mechanical analysis model, so called 'analytical mode', because the detailed mechanical analysis requires large amount of calculation time and computing power. Even

  15. Development of mechanical analysis module for simulation of SFR fuel rod behavior using finite element method

    International Nuclear Information System (INIS)

    Shin, Andong; Jeong, Hyedong; Suh, Namduk; Kim, Hyochan; Yang, Yongsik

    2014-01-01

    Korean SFR developer decided to adapt metal fuel, current study focused on the metal fuel instead of oxide fuel. The SFR metal fuel has been developed by Korea Atomic Energy Research Institute (KAERI) and many efforts focused on designing and manufacturing the metal fuel. Since a nuclear fuel is the first barrier to protect radioactive isotope release, the fuel's integrity must be secured during steady-state operation and accident condition within an acceptable range. Whereas the design and evaluation methodologies, code systems and test procedures of a light water reactor fuel are sufficiently established, those of the SFR fuel needs more technical advances. In the view of regulatory point, there are still many challenging issues which are required to secure the safety of fuel and reactors. For this reason, the Korea Institute of Nuclear Safety (KINS) has launched the new project to develop the regulatory technology for SFR system including a fuel area. The ALFUS code was developed by CRIEPI and employs mechanistic model for fission gas release and swelling of fuel slug. In the code system, a finite element method was introduced to analyze the fuel and cladding's mechanical behaviors. The FEAST code is more advanced code system for SFR which adopted mechanistic FGR and swelling model but still use analytical model to simulate fuel and cladding mechanical behavior. Based on the survey of the previous studies, fuel and cladding mechanical model should be improved. Analysis of mechanical behavior for fuel rod is crucial to evaluate overall rod's integrity. In addition, it is because contact between fuel slug and cladding or an over-pressure of rod internal pressure can cause rod failure during steady-state and other operation condition. The most of reference codes have simplified mechanical analysis model, so called 'analytical mode', because the detailed mechanical analysis requires large amount of calculation time and computing power. Even

  16. QC methods and means during pellets and fuel rods manufacturing at JSC 'MSZ'

    International Nuclear Information System (INIS)

    Kouznetsov, A.I.

    2000-01-01

    The report contains the description of the main methods and devices used in fabrication of pellets and fuel rods to prove their conformity to the requirements of technical specifications. The basic principals, range and accuracy of methods and devices are considered in detail, as well as system of metrological support of measurements. The latter includes the metrological certification and periodical verification of the devices, metrological qualification of measurement procedures, standard samples provision and checking the correctness of the analyses performance. If one makes an overall review of testing methods used in different fuel production plants he will find that most part of methods and devices are very similar. There are still some variations in methods which could be a subject for interesting discussions among specialists. This report contains a brief review of testing methods and devices used at our plant. More detailed description is given to methods which differ from those commonly used. (author)

  17. A simple method for in vivo measurement of implant rod three-dimensional geometry during scoliosis surgery.

    Science.gov (United States)

    Salmingo, Remel A; Tadano, Shigeru; Fujisaki, Kazuhiro; Abe, Yuichiro; Ito, Manabu

    2012-05-01

    Scoliosis is defined as a spinal pathology characterized as a three-dimensional deformity of the spine combined with vertebral rotation. Treatment for severe scoliosis is achieved when the scoliotic spine is surgically corrected and fixed using implanted rods and screws. Several studies performed biomechanical modeling and corrective forces measurements of scoliosis correction. These studies were able to predict the clinical outcome and measured the corrective forces acting on screws, however, they were not able to measure the intraoperative three-dimensional geometry of the spinal rod. In effect, the results of biomechanical modeling might not be so realistic and the corrective forces during the surgical correction procedure were intra-operatively difficult to measure. Projective geometry has been shown to be successful in the reconstruction of a three-dimensional structure using a series of images obtained from different views. In this study, we propose a new method to measure the three-dimensional geometry of an implant rod using two cameras. The reconstruction method requires only a few parameters, the included angle θ between the two cameras, the actual length of the rod in mm, and the location of points for curve fitting. The implant rod utilized in spine surgery was used to evaluate the accuracy of the current method. The three-dimensional geometry of the rod was measured from the image obtained by a scanner and compared to the proposed method using two cameras. The mean error in the reconstruction measurements ranged from 0.32 to 0.45 mm. The method presented here demonstrated the possibility of intra-operatively measuring the three-dimensional geometry of spinal rod. The proposed method could be used in surgical procedures to better understand the biomechanics of scoliosis correction through real-time measurement of three-dimensional implant rod geometry in vivo.

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

    Science.gov (United States)

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

    2016-08-17

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

  19. 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.)

  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. New method of determining the thermal utilization factor of a cell; Nouvelle methode de determination du facteur d'utilisation thermique d'une cellule

    Energy Technology Data Exchange (ETDEWEB)

    Amouyal, A; Benoist, P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1956-07-01

    A new formula for the thermal utilization factor is derived, which, while comparable in simplicity to the formula given by elementary diffusion theory, furnishes much more precise results. This is clearly brought out by comparison with the results given by the S{sub n} and spherical harmonics methods. (author) [French] Une nouvelle expression du facteur d'utilisation thermique, d'une simplicite comparable a celle de Ia theorie elementaire, est etablie. La comparaison avec les resultats fournis par la methode S{sub n} et les methodes d'harmoniques spheriques montre que la precision obtenue par cette formule est tres superieure a celle que donne la theorie elementaire. (auteur)

  2. Utilization of the MAT method to analyze the nucleate boiling boundary in rod bundles subchannels

    International Nuclear Information System (INIS)

    Pedron, M.Q.

    1983-01-01

    The digital program PANTERA-1P, a new version of the COBRA-IIIC code, developed at CDTN, is directed to the thermal-hydraulic analysis of water cooled rod bundles and reactor cores, insteady state and transient conditions. Both the new and the old code versions have identical capacities in what concerns evaluation of fluid variables, nevertheless PANTERA-1P has better and faster performance. Improvements introduced in the scheme for solution of the conservation equations have contributed significantly to reduce the computer time, without affecting the accuracy of results. While the momentum equations are solved in COBRA-IIIC for the crossflow distribution, the PANTERA-1P code solves these equations for the pressure distribution by using the MAT method (Modified and Advanced Theta). The calculation of the pressure coefficient matrix has been optimized and simultaneous linear equations are solved optionally by means of the transpose elimination with storage requirements or the successive over-relaxation methods. The program presents others features specially in what concerns the thermal conduction model for fuel rods and the critical heat flux calculations options. A new input/output scheme is provided for optional use of the British or Internacional System of Units. The results of the program are compared to the critical heat flux experimental data and to the results of COBRA-IIIC. Excellent agreement is observed in both cases. (Author) [pt

  3. Advances in methods of commercial FBR core characteristics analyses. Investigations of a treatment of the double-heterogeneity and a method to calculate homogenized control rod cross sections

    Energy Technology Data Exchange (ETDEWEB)

    Sugino, Kazuteru [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center; Iwai, Takehiko

    1998-07-01

    A standard data base for FBR core nuclear design is under development in order to improve the accuracy of FBR design calculation. As a part of the development, we investigated an improved treatment of double-heterogeneity and a method to calculate homogenized control rod cross sections in a commercial reactor geometry, for the betterment of the analytical accuracy of commercial FBR core characteristics. As an improvement in the treatment of double-heterogeneity, we derived a new method (the direct method) and compared both this and conventional methods with continuous energy Monte-Carlo calculations. In addition, we investigated the applicability of the reaction rate ratio preservation method as a advanced method to calculate homogenized control rod cross sections. The present studies gave the following information: (1) An improved treatment of double-heterogeneity: for criticality the conventional method showed good agreement with Monte-Carlo result within one sigma standard deviation; the direct method was consistent with conventional one. Preliminary evaluation of effects in core characteristics other than criticality showed that the effect of sodium void reactivity (coolant reactivity) due to the double-heterogeneity was large. (2) An advanced method to calculate homogenize control rod cross sections: for control rod worths the reaction rate ratio preservation method agreed with those produced by the calculations with the control rod heterogeneity included in the core geometry; in Monju control rod worth analysis, the present method overestimated control rod worths by 1 to 2% compared with the conventional method, but these differences were caused by more accurate model in the present method and it is considered that this method is more reliable than the conventional one. These two methods investigated in this study can be directly applied to core characteristics other than criticality or control rod worth. Thus it is concluded that these methods will

  4. Specific features of the determination of the pellet-cladding gap of the fuel rods by non-destructive method

    International Nuclear Information System (INIS)

    Amosov, S.V.; Pavlov, S.V.

    2002-01-01

    This report describes the specific features of determining the pellet-cladding gap of the irradiated WWER-1000 fuel rods by nondestructive method. The method is based on the elastic radial deformation of the cladding up to its contact with the fuel. The value of deformation of cladding till its contacting fuel when radial force changes from F max to 0 is proposed as a measuring parameter for determination of the diametrical gap. Because of the features of compression method, the obtained gap value is not analog of the gap measured on micrograph of the fuel rod cross-section. Results of metallography can provide only qualitative evaluation of its method efficiency. Comparison of the values determined by non-destructive method and metallography for WWER-1000 fuel rods with burnup from 25 to 55 MWd/kg U testified that the results of compression method can be used as a low estimate of the pellet-cladding gap value. (author)

  5. Numerical study on the rotation of an elastic rod in a viscous fluid using an immersed boundary method

    International Nuclear Information System (INIS)

    Maniyeri, Ranjith; Kang, Sang Mo

    2012-01-01

    We present a three dimensional computational model based on an immersed boundary (IB) method to study the hydrodynamic features of a solid flexible cylindrical rod in a viscous fluid driven at one side by a tiny motor. The elastic rod is modelled by a number of circular cross-sections with twelve IB points on each cross-section. Three types of elastic links are created from each IB point to obtain an elastic network model of the rod and the first cross-section is modelled as the motor part. The elastic forces are computed based on an elastic energy approach and the motor forces are obtained from the applied angular frequency of rotation of the motor. The Stokes equations governing the fluid are solved on a staggered Cartesian grid system using the fractional-step based finite-volume method. Numerical simulations are performed to demonstrate the three dynamical stages of rod motion- twirling, whirling and overwhirling for different rotational frequency of the motor. It is revealed that for low rotational frequencies, the rod undergoes stable rigid body motion known as twirling. For high rotational frequencies of the motor, it is observed that the rod initially undergoes whirling motion and attains an unstable helical shape. Further, it is noticed that a discontinuous shape transition occurs for the rod and it folds back on itself. This unstable motion is referred to as overwhirling. It is also found that there exists a critical value of angular frequency of rotation of the motor below which the rod is subjected to twirling motion and above which it undergoes overwhirling motion

  6. The research for a method for controlling atmospheric pollution in a plutonium treatment cell; Recherche d'une methode de controle de la pollution atmospherique dans une cellule traitant le polonium

    Energy Technology Data Exchange (ETDEWEB)

    Bouteiller, E [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

    1960-07-01

    The aerosols likely to contain {sup 210}Po are trapped by bubbling through an acid solution. The problem of determining the polonium concentration of the solution has been solved by combining it with telluric acid. The reduction is carried out at pH = 0 using a solution of stannous chloride. We have studied the factors influencing coprecipitation; nature of the medium, pH, amounts of telluric acid and of stannous chloride. The average yield for the precipitation is 93 per cent. After precipitation, the polonium is filtered on paper and its activity is measured by means of a scintillation counter. The method makes it possible to measure 1/20 of the maximum permissible dose with an accuracy of 20 per cent. (author) [French] Les aerosols susceptibles de contenir du {sup 210}Po sont captes par barbotage dans une solution acide. Le probleme de la concentration de la solution en polonium a ete resolu en entrainant celui-ci avec de l'acide tellurique. La reduction est effectuee a pH = 0 par une solution de chlorure stanneux. Nous avons etudie les facteurs influencant la coprecipitation: nature du milieu, pH, quantite d'entraineur, quantite de chlorure stanneux. Le rendement moyen de la precipitation est de 93 pour cent. Apres precipitation, le polonium est filtre sur papier et son activite mesuree au compteur a scintillations. La methode permet de mesurer le 1/20 de la dose maximum permise avec une precision de 20 pour cent. (auteur)

  7. Developpement d'une methode calorimetrique de mesure des pertes ac pour des rubans supraconducteurs a haute temperature critique

    Science.gov (United States)

    Dolez, Patricia

    Le travail de recherche effectue dans le cadre de ce projet de doctorat a permis la mise au point d'une methode de mesure des pertes ac destinee a l'etude des supraconducteurs a haute temperature critique. Pour le choix des principes de cette methode, nous nous sommes inspires de travaux anterieurs realises sur les supraconducteurs conventionnels, afin de proposer une alternative a la technique electrique, presentant lors du debut de cette these des problemes lies a la variation du resultat des mesures selon la position des contacts de tension sur la surface de l'echantillon, et de pouvoir mesurer les pertes ac dans des conditions simulant la realite des futures applications industrielles des rubans supraconducteurs: en particulier, cette methode utilise la technique calorimetrique, associee a une calibration simultanee et in situ. La validite de la methode a ete verifiee de maniere theorique et experimentale: d'une part, des mesures ont ete realisees sur des echantillons de Bi-2223 recouverts d'argent ou d'alliage d'argent-or et comparees avec les predictions theoriques donnees par Norris, nous indiquant la nature majoritairement hysteretique des pertes ac dans nos echantillons; d'autre part, une mesure electrique a ete realisee in situ dont les resultats correspondent parfaitement a ceux donnes par notre methode calorimetrique. Par ailleurs, nous avons compare la dependance en courant et en frequence des pertes ac d'un echantillon avant et apres qu'il ait ete endommage. Ces mesures semblent indiquer une relation entre la valeur du coefficient de la loi de puissance modelisant la dependance des pertes avec le courant, et les inhomogeneites longitudinales du courant critique induites par l'endommagement. De plus, la variation en frequence montre qu'au niveau des grosses fractures transverses creees par l'endommagement dans le coeur supraconducteur, le courant se partage localement de maniere a peu pres equivalente entre les quelques grains de matiere

  8. Zirconium-based alloys, nuclear fuel rods and nuclear reactors including such alloys, and related methods

    Science.gov (United States)

    Mariani, Robert Dominick

    2014-09-09

    Zirconium-based metal alloy compositions comprise zirconium, a first additive in which the permeability of hydrogen decreases with increasing temperatures at least over a temperature range extending from 350.degree. C. to 750.degree. C., and a second additive having a solubility in zirconium over the temperature range extending from 350.degree. C. to 750.degree. C. At least one of a solubility of the first additive in the second additive over the temperature range extending from 350.degree. C. to 750.degree. C. and a solubility of the second additive in the first additive over the temperature range extending from 350.degree. C. to 750.degree. C. is higher than the solubility of the second additive in zirconium over the temperature range extending from 350.degree. C. to 750.degree. C. Nuclear fuel rods include a cladding material comprising such metal alloy compositions, and nuclear reactors include such fuel rods. Methods are used to fabricate such zirconium-based metal alloy compositions.

  9. A model for asymmetric ballooning and analyses of ballooning behaviour of single rods with probabilistic methods

    International Nuclear Information System (INIS)

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

    1985-01-01

    Plastic deformation behaviour of Zircaloy cladding has been extensively examined in the past and can be described best by a model for asymmetric deformation. Slight displacement between the pellet and cladding will always exist and this will lead to the formation of azimuthal temperature differences. The ballooning process is strongly temperature dependent and, as a result of the built up temperature differences, differing deformation behaviours along the circumference of the cladding result. The calculated ballooning of cladding is mainly influenced by its temperature, the applied burst criterion and the parameters used in the deformation model. All these influencing parameters possess uncertainties. In order to quantify these uncertainties and to estimate distribution functions of important parameters such as temperature and deformation the response surface method was applied. For a hot rod the calculated standard deviation of cladding temperature amounts to 50 K. From this high value the large influence of the external cooling conditions on the deformation and burst behaviour of cladding can be estimated. In an additional statistical examination the parameters of deformation and burst models have been included and their influence on the deformation of the rod has been studied. (author)

  10. Recycling temperature elevation device and temperature control method for control rod driving system

    International Nuclear Information System (INIS)

    Okamura, Hajime.

    1996-01-01

    The present invention concerns a device for and a method of controlling a recycling temperature control device for control rod drives (CRD) of a nuclear power plant, which can prevent occurrence of cavitation and keep the amount of cooling water to be transferred to a water source transfer pipeline thereby improving maintenanciability, operationability and reliability. Namely, a supply pipeline supplies cooling water required for the control rod drives from a water source. A CRD pump elevates the pressure of the cooling water. A recycling pipeline is branched from the downstream of the CRD pump of the supply pipeline and connected to the supply pipeline at the upstream of the CRD pump. A first pressure element and a restricting valve disposed at the upstream thereof are connected to the upstream of the CRD pump and the water source transfer pipeline. The water source transfer pipeline is branched from the recycling pipeline and connected to the water source. A second pressure element is disposed to a recycling pipeline at the downstream of the branched point from the water source transfer pipeline. (I.S.)

  11. Method for calculating the critical heat flux in mixed rod assemblies based on the tables of crisis in bundles

    International Nuclear Information System (INIS)

    Bobkov, V.P.

    2000-01-01

    The method for calculating the critical heat flux in the mixed rod assemblies, for example RBMK, containing three-four angle and peripheral macrocells, is presented. The method is based on generalization of experimental data in form of tables for the rods beams. It is recommended for the areas of parameters both provided for by experimental data and for others, where the data are absent. The advantages of the table method as follows: it is acceptable within a wide range of parameters and provides for smooth description of dependence of critical heat fluxes on these parameters; it is characterized by clearness, high reliability and accuracy and is easy in application [ru

  12. Theoretical interpretations and experimental verifications of a radioelectric resonance method for measuring the electronic density and collision frequency in a discharge plasma in gases; Interpretations theoriques et verifications experimentales d'une methode de resonance radioelectrique pour la mesure de la densite d'une decharge dans les gaz

    Energy Technology Data Exchange (ETDEWEB)

    Nguyen Trong, Khoi [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

    1960-07-01

    Theoretical discussions and experimental verifications of one radioelectric resonance method for measuring plasma electronic density and collision frequency. (author) [French] Discussions theoriques et verifications experimentales sur une methode de resonance radioelectrique pour la mesure de la densite electronique et de la frequence de collision d'un plasma d'une decharge dans le gaz. (auteur)

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

    International Nuclear Information System (INIS)

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

    1981-01-01

    A technique is provided for engaging and disengaging burnable poison rods from a spider in a nuclear reactor fuel assembly. A cap on the end of each of the burnable poison rods is provided with a shank or stem that is received in a respective bore formed in the spider. A frangible flange secures the shank and rod to the spider. Pressing the shank in the direction of the bore axis by means, e.g., of a plate ruptures the frangible flange to release the rod from the spider. (author)

  14. 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.)

  15. Method and system for the production of fuel rods for nuclear reactors

    International Nuclear Information System (INIS)

    Meinhardt, V.; Schultz, R.; Gall, A.

    1975-01-01

    The invention deals with a method to produce nuclear reactor fuel rods which contain the fuel in a cladding tube and in which an alkali metal is put in the space between the fuel and cladding for better heat transfer from the fuel to the cladding. The bubble-free filling-in of the alkali metal causes difficulties; the invention suggests to carry out the filling-in of fuel and alkali metal in a horizontal position of the cladding to rinse through the free space firstly with alkali metal, then to fill with fuel and then with the alkali metal taking account of necessary space for the fission gas volume. Equipment to carry out this process in which all operations are carried out in a vacuum box is described. (UWI) [de

  16. CFD method research on characteristics cells in rod bundle fuel assembly

    International Nuclear Information System (INIS)

    Chen Jie; Chen Bingyan; Zhang Hong

    2011-01-01

    Two characteristic cells are in AFA-3G fuel assembly, that is typical cell and control rod guide cell. And there are some rules on the arrangement of mixing vanes. For the two characteristic cells, mixing capability is evaluated axially from the point of the first and second kind of sub-channel with CFD method. Mass mixing and heat mixing are interaction but different with each other. Although the mass mixing in the first kind of sub-channel is stronger, the thermal capability of the two is to some tune from the point of heat transfer. In the experiment research on thermal-hydraulic performance of AFA-3G fuel assembly, the arrangements of mixing vanes should refer to the two spacer grids of characteristic cells. (authors)

  17. Application of contact mechanics for fretting damage of fuel rod: part 1 influence functions and numerical method

    International Nuclear Information System (INIS)

    Kim, H. K.; Yoon, K. H.; Kang, H. S.; Song, G. N.

    1998-01-01

    For the analysis of the fretting problem of the fuel rods, present paper(Part I) shows the numerical method developed for evaluating the stresses on the contact surfaces between the fuel rods and the spacer grids. Theory of Contact Mechanics was incorporated. Contact area was regarded as a plane strain condition, so plane problem was taken into consideration. Normal stress profile on the contact surface was assumed to be Hertzian. As for the direction of the shear load, a closed load path, e.g. load increase in transverse increase in axial decrease in transverse decrease in axial increase in transverse increase in axial direction was considered for simulating the rod vibration in a reactor core. Partial slip problem was consulted. As for the numerical method, a triangular traction element was utilized and the corresponding influence functions were evaluated. Numerical program has been implemented for the present analysis, of which the validity was verified by comparing the Mindlin-Cattaneo solution

  18. Data Rods: High Speed, Time-Series Analysis of Massive Cryospheric Data Sets Using Object-Oriented Database Methods

    Science.gov (United States)

    Liang, Y.; Gallaher, D. W.; Grant, G.; Lv, Q.

    2011-12-01

    Change over time, is the central driver of climate change detection. The goal is to diagnose the underlying causes, and make projections into the future. In an effort to optimize this process we have developed the Data Rod model, an object-oriented approach that provides the ability to query grid cell changes and their relationships to neighboring grid cells through time. The time series data is organized in time-centric structures called "data rods." A single data rod can be pictured as the multi-spectral data history at one grid cell: a vertical column of data through time. This resolves the long-standing problem of managing time-series data and opens new possibilities for temporal data analysis. This structure enables rapid time- centric analysis at any grid cell across multiple sensors and satellite platforms. Collections of data rods can be spatially and temporally filtered, statistically analyzed, and aggregated for use with pattern matching algorithms. Likewise, individual image pixels can be extracted to generate multi-spectral imagery at any spatial and temporal location. The Data Rods project has created a series of prototype databases to store and analyze massive datasets containing multi-modality remote sensing data. Using object-oriented technology, this method overcomes the operational limitations of traditional relational databases. To demonstrate the speed and efficiency of time-centric analysis using the Data Rods model, we have developed a sea ice detection algorithm. This application determines the concentration of sea ice in a small spatial region across a long temporal window. If performed using traditional analytical techniques, this task would typically require extensive data downloads and spatial filtering. Using Data Rods databases, the exact spatio-temporal data set is immediately available No extraneous data is downloaded, and all selected data querying occurs transparently on the server side. Moreover, fundamental statistical

  19. Method for operating a nuclear reactor with scrammable part length rod

    International Nuclear Information System (INIS)

    Bevilacqua, F.

    1979-01-01

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

  20. The Study on Radioactive Nuclide Distributions within a Fuel Rod by Tomographic Gamma Scanning Method

    International Nuclear Information System (INIS)

    Quanhu, Zhang; Lee, H. K.; Hong, K. P.; Choo, Y. S.; Kim, D. S.

    2005-06-01

    Based on the specified need of the IMEF, the feasibility of Tomographic Gamma Scanning (TGS) technique has been investigated for its potential for non-destructive gamma scanning measurements of irradiated fuel rods. TGS technique has been developed for determining some radioactive isotopes' distributions of a fuel rod in hot cell. The results obtained from the simulation model extracting from real gamma scanning experimental condition in this work by new developed computer simulation codes confirmed that the gamma emission TGS technique has potential for determination of radioactive isotopes' distributions of a fuel rod. In order to verify the simulation codes, we have designed several computation schemes for both 3 by 3 and 10 by 10 fuel rod model under present situation at M1 hot cell in IMEF. The results which relative errors are less than 10% show that we have simulated and implemented determination of radioactive isotopes' distributions on simulated fuel rod by TGS technique successfully

  1. Study of development of non-destructive method for determining FGR from high burned PWR type fuel rod

    International Nuclear Information System (INIS)

    Yanagisawa, Kazuaki; Miyanishi, Hideyuki; Kitagawa, Isamu; Iida, Shozo; Ito, Tadaharu; Amano, Hidetoshi.

    1991-11-01

    Experimental study was made to evaluate the FGR (Fission Product Gas Release) from high burned PWR type fuel rods by means of non-destructive method through measurement of the gamma activity of 85 Kr isotope which was accumulated in the fuel top plenum. Experimental result shows that it is possible to know the amounts of FGR at fuel plenum by the equations given in the followings. FGR = 0.28C/V f or FGR = 0.07C where, FGR (%) is the amounts of Xe and Kr released from UO 2 fuel, C (counts/h) the radioactivity of 85 Kr at plenum of the tested fuel rod and V f (ml) the plenum volume of the tested fuel rod, respectively. The present study was made by using 14 x 14 PWR type fuel rods preirradiated up to the burn-up of 42.1 MWd/kgU, followed by the pulse irradiation at Nuclear Safety Research Reactor of Japan Atomic Energy Research Institute (JAERI). The FGR of the tested segmented fuel rods were measured by puncturing and found to range from 0.6% to 12% according to the magnitude of the deposited energy given by pulse. Estimated experimental error bands against the above equations were within plus minus 30%. (author)

  2. Study of an alternative method for inspection of rods with UO{sub 2} pellets early manufactured

    Energy Technology Data Exchange (ETDEWEB)

    Carnaval, João Paulo R.; Oliveira, Carlos A.; Beltran, Dalton J.M.C., E-mail: joaocarnaval@inb.gov.br, E-mail: carlossilva@inb.gov.br, E-mail: daltonbeltran@inb.gov.br [Indústrias Nucleares do Brasil S.A. (INB), Resende, RJ (Brazil). Gerência de Engenharia do Produto e Gerência de Análise do Combustível

    2017-07-01

    The inspection of the fuel rods manufactured at INB, for production of fuel assemblies, is based on a group of scintillators detectors in series scanning the products. These detectors capture the gamma rays emitted on the decay of uranium isotopes (passive measurement) and determine the enrichment level ({sup 235}U weight percent) of the UO{sub 2} pellets inside the fuel rods. During the inspection of fuel rods for Angra-1 21{sup st} Reload, it was found that the 2.6% {sup 235}U and 4.15% {sup 235}U pellets stacks behave as 2.6% {sup 235}U only. The investigation of this event allowed to conclude that the measurement of enrichment may be affected by the loss of the secular equilibrium among uranium isotopes and their decay products caused by the AUC precipitation during the UO{sub 2} powder and pellet fabrication. Therefore, the spectrum background created by Compton scattering, inside Rod Scanner detectors, from high energies of {sup 238}U products decay affect the {sup 235}U% measurement. After continuous measurements, the 2.6% {sup 235}U and 4.15% {sup 235}U pellets stacks became distinguished and the results were used to calculate an 'equilibrium factor'. It was concluded that after 35 days the UO{sub 2} powder should reach approximately 60% of secular equilibrium reinstatement and the rods assembled with the pellets produced from this powder would be adequate for inspection on Rod Scanner. It was concluded that would be possible to achieve the equilibrium factor by blending a lot of UO{sub 2} powder manufactured a long time ago (old powder) with another lot early manufactured (young powder) resulting in a lot which would provide pellets and, consequently, rods adequate for inspection by Rod Scanner. This work presents a study of an alternative method to perform the inspection of fuel rods with UO{sub 2} pellets early manufactured aiming to provide quality assurance for the product. (author)

  3. 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.)

  4. Calculational model based on influence function method for power distribution and control rod worth in fast reactors

    International Nuclear Information System (INIS)

    Toshio, S.; Kazuo, A.

    1983-01-01

    A model for calculating the power distribution and the control rod worth in fast reactors has been developed. This model is based on the influence function method. The characteristics of the model are as follows: 1. Influence functions for any changes in the control rod insertion ratio are expressed by using an influence function for an appropriate control rod insertion in order to reduce the computer memory size required for the method. 2. A control rod worth is calculated on the basis of a one-group approximation in which cross sections are generated by bilinear (flux-adjoint) weighting, not the usual flux weighting, in order to reduce the collapse error. 3. An effective neutron multiplication factor is calculated by adjoint weighting in order to reduce the effect of the error in the one-group flux distribution. The results obtained in numerical examinations of a prototype fast reactor indicate that this method is suitable for on-line core performance evaluation because of a short computing time and a small memory size

  5. Calculational model based on influence function method for power distribution and control rod worth in fast reactors

    International Nuclear Information System (INIS)

    Sanda, T.; Azekura, K.

    1983-01-01

    A model for calculating the power distribution and the control rod worth in fast reactors has been developed. This model is based on the influence function method. The characteristics of the model are as follows: Influence functions for any changes in the control rod insertion ratio are expressed by using an influence function for an appropriate control rod insertion in order to reduce the computer memory size required for the method. A control rod worth is calculated on the basis of a one-group approximation in which cross sections are generated by bilinear (flux-adjoint) weighting, not the usual flux weighting, in order to reduce the collapse error. An effective neutron multiplication factor is calculated by adjoint weighting in order to reduce the effect of the error in the one-group flux distribution. The results obtained in numerical examinations of a prototype fast reactor indicate that this method is suitable for on-line core performance evaluation because of a short computing time and a small memory size

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

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  7. 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.)

  8. The State of the Art of the Decladding Method for the Spent Nuclear Fuel Rods

    International Nuclear Information System (INIS)

    Jung, Jae Hoo; Park, B. S.; Kim, K. H.; Hong, D. H.; Yoon, J. S.; Lee, H. J.; Kim, S. H.; Song, T. G.; Lim, K. M.; Lee, J. K.

    2006-12-01

    Our country's energy consumption is increasing day after day even though it relies on imports for more than 95 percent of its energy needs. In this circumstances, the atomic energy is a promising alternative to solve the problem of an energy security and an environmental preservation simultaneously. However, nuclear power produces spent fuel which is a highly radioactive waste. For a reliable and effective management of the spent fuel, the ACP(Advanced Spent Fuel Conditioning Process) is being developed at the KAERI. As a first state equipment of the ACP, a decladding machine is used to separate spent fuel rod into the UO 2 pellets and hulls. This technical report aims to analyze existing decladding methods, and then, find a suitable decladding mechanism for the ACP. Many studies on the decladding of spent fuel can be categorized two approaches: chemical approach and mechanical one. In this report, we concentrated on the mechanical decladding approach. We developed engineering scale decladding device(20 kgHM/batch) and evaluated the performance through the verification experiments. We expect that this technical report helps in developing a scale-up equipment and technology

  9. Study on reactor power transient characteristics (reactor training experiments). Control rod reactivity calibration by positive period method and other experiment

    International Nuclear Information System (INIS)

    Ozaki, Yoshihiko; Sunagawa, Takeyoshi

    2014-01-01

    In this paper, it is reported about some experiments that have been carried out in the reactor training that targets sophomore of the department of applied nuclear engineering, FUT. Reactor of Kinki University Atomic Energy Research Institute (UTR-KINKI) was used for reactor training. When each critical state was achieved at different reactor output respectively in reactor operating, it was confirmed that the control rod position at that time does not change. Further, control rod reactivity calibration experiments using positive Period method were carried out for shim safety rod and regulating rod, respectively. The results were obtained as reasonable values in comparison with the nominal value of the UTR-KINKI. The measurement of reactor power change after reactor scram was performed, and the presence of the delayed neutron precursor was confirmed by calculating the half-life. The spatial dose rate measurement experiment of neutrons and γ-rays in the reactor room in a reactor power 1W operating conditions were also performed. (author)

  10. F.E.M. of PWR's control rod cluster. Parametrical study of vibrating behavior by an Experiment Design method

    International Nuclear Information System (INIS)

    Bosselut, D.; Soulier, B.

    1997-03-01

    Some finite element models have been performed at EDF to simulate the vibrations of rod cluster and to analyse the wear phenomenon of rods using parametrical studies. In the first part, one of the finite element models is presented. The location of excitation sources is described. The calculated values are: rod displacement in the guiding cards, shock forces on the guiding cards and wear power produced. In the second part, a parametrical study is presented for a given computer experiment domain with an Experimental Design method. The building of the computer experiment design is described. The used polynomial model has all linear, quadratic and interactive terms for each of the 6 parameters (26 coefficients), 34 polynomials have been built to approach the effective shock forces and the mean wear power at each of the 17 guiding points. In the last part, the influence of parameters on calculated mean wear power is shown along rods and some responses surfaces are visualized. Systematism and closeness of experiment design technique is underlined. Easy simulation of all the response domain by polynomial approach, allows comparison with experiment feedback. (author)

  11. Spring retainer apparatus and method for facilitating loading of fuel rods into a fuel assembly grid

    International Nuclear Information System (INIS)

    De Mario, E.E.

    1988-01-01

    For use with a fuel assembly having at least one grid formed of interleaved straps defining hollow cells for respectively receiving fuel rods, at least some of the straps being disposed in pairs thereof so as to form springs in pairs therof being positioned in back-to-back relationships between adjacent ones of the cells, the springs in each pair thereof being configured to normally assume expanded positions in which they are displaced away from one another to engage fuel rods received in the respective cells and being deflectible to retracted positions in which they are displaced toward one another to allow loading of the fuel rods in the respective cells without engaging the springs, a spring retainer apparatus for facilitating the loading of the fuel rods into the cells of the fuel assembly grid is described comprising: (a) elongated holder bars, each holder bar being alignable with one of the pairs of the straps of the grid which defines the pairs of springs and extendible along, and in spaced relation from, the one strap pair and between and spaced from positions occupied by fuel rods when received in the cells of the grid; and (b) supported by each of the holder bars corresponding to the pairs of springs defined by the pair of straps aligned with the holder bar

  12. Study and development of an optical method for the measurement of convection coefficients; Etude et developpement d'une methode optique pour la mesure du coefficient de convection

    Energy Technology Data Exchange (ETDEWEB)

    Crowther, David J.

    1990-03-06

    This research thesis addresses the field of fluid-wall thermal exchanges in which the notion of exchange coefficient is notably useful to design, size and optimise devices. A first part reports a bibliographical study which gives an overview of solutions envisaged to determine the convection coefficient in permanent regime with the use of flow sensors, as well as in transient regime. Then, the author reports the development of an unsteady method which is based on the analysis of the cooling kinetics of the front face of a convecting wall, after a unique energetic perturbation (an infinitely brief pulse, or a finite duration energy step). This method is applied to the general case (wall with finite thickness) and to the case of a semi-infinite wall which is typical of materials which are weak thermal conductors. This is extended to the case of good thermal conductors by considering a thermally thin wall. After a detailed description of the experimental bench, above-mentioned solutions are applied to insulating and good thermal conducting materials. In order to validate results of an analysis in transient regime, they are compared with measurements performed in permanent regime with a flow-metering technique. The study of the principle of the dissipation-based flow sensor, and its operation are reported. Experimental results are presented for both methods (pulse and flow sensor), and compared in order to highlight the interest of the unsteady method [French] Difficile a mesurer, le coefficient de convection reste cependant une grandeur necessaire au calcul et a l'optimisation de tout systeme thermique. L'amelioration des capteurs thermiques permet aujourd'hui de concevoir une methode optique, utilisable a distance, et non destructive. Nous proposons dans ce but, un procede de mesure en regime transitoire base sur la radiometrie photothermique impulsionnelle. L'analyse du regime de relaxation d'une paroi, apres une brusque elevation de temperature, permet de remonter

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

    Energy Technology Data Exchange (ETDEWEB)

    Jabsen, F S

    1972-10-16

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

  14. Measurement of control rods efficiency at the RB reactor by pulse method; Merenje efikasnosti kontrolnih sipki u reaktoru RB impulsnom metodom

    Energy Technology Data Exchange (ETDEWEB)

    Petrovic, M; Markovic, V; Velickovic, Lj [Boris Kidric Institute of nuclear sciences, Vinca, Belgrade (Yugoslavia)

    1963-07-01

    Pulse method was applied for measuring the efficiency of control rods at the RB reactor. This paper describes the theory of experiment, experimental procedure applied and results obtained. Results are considered to be useful for safety analysis. it was found that the influence of delayed neutrons is rather small and could be neglected in estimation of rods efficiency.

  15. The pseudo-harmonics method: an application involving perturbations caused by control rod insertion in PWR reactors

    International Nuclear Information System (INIS)

    Claro, L.H.; Alvim, A.C.M.; Thome, Z.D.

    1988-08-01

    The objective of this work is to stydy the effect of intense perturbations, such as control rod insertion in the core of PWR reactors, through a perturbation approach consisting of a modified version of the pseudo-harmonics method. A typical one-dimensional PWR reactor model was used as a reference state, from which two perturbations were imposed, simulation gray and black control rod insertion. In the first case, eigenvalue convergence was achieved with the eighth order of approximation approximation and perturbed fluxes and eigenvalue estimates agreed very well with direct calculation results. The second case tested represents a very intense localized perturbation. Oscillation in keff were observed er of approximation increased and the method failed to converge. Results obtained indicate that the pseudo-harmonics method can be used to compute 2 group fluxes and fundamental eigenvalue of perturbated states resulting from gray control rod insertion in PWR reactors. The method is limited, however, by perturbation intensity, as other perturbation methods are. (author) [pt

  16. Developpement d'une methode de Monte Carlo dependante du temps et application au reacteur de type CANDU-6

    Science.gov (United States)

    Mahjoub, Mehdi

    La resolution de l'equation de Boltzmann demeure une etape importante dans la prediction du comportement d'un reacteur nucleaire. Malheureusement, la resolution de cette equation presente toujours un defi pour une geometrie complexe (reacteur) tout comme pour une geometrie simple (cellule). Ainsi, pour predire le comportement d'un reacteur nucleaire,un schema de calcul a deux etapes est necessaire. La premiere etape consiste a obtenir les parametres nucleaires d'une cellule du reacteur apres une etape d'homogeneisation et condensation. La deuxieme etape consiste en un calcul de diffusion pour tout le reacteur en utilisant les resultats de la premiere etape tout en simplifiant la geometrie du reacteur a un ensemble de cellules homogenes le tout entoure de reflecteur. Lors des transitoires (accident), ces deux etapes sont insuffisantes pour pouvoir predire le comportement du reacteur. Comme la resolution de l'equation de Boltzmann dans sa forme dependante du temps presente toujours un defi de taille pour tous types de geometries,un autre schema de calcul est necessaire. Afin de contourner cette difficulte, l'hypothese adiabatique est utilisee. Elle se concretise en un schema de calcul a quatre etapes. La premiere et deuxieme etapes demeurent les memes pour des conditions nominales du reacteur. La troisieme etape se resume a obtenir les nouvelles proprietes nucleaires de la cellule a la suite de la perturbation pour les utiliser, au niveau de la quatrieme etape, dans un nouveau calcul de reacteur et obtenir l'effet de la perturbation sur le reacteur. Ce projet vise a verifier cette hypothese. Ainsi, un nouveau schema de calcul a ete defini. La premiere etape de ce projet a ete de creer un nouveau logiciel capable de resoudre l'equation de Boltzmann dependante du temps par la methode stochastique Monte Carlo dans le but d'obtenir des sections efficaces qui evoluent dans le temps. Ce code a ete utilise pour simuler un accident LOCA dans un reacteur nucleaire de type

  17. Operation method of the X-ray equipment for the investigation of the ballooning of LWR-fuel rod simulators

    International Nuclear Information System (INIS)

    Mueller, S.; Thun, G.

    1977-06-01

    An X-Ray-equipment is described which has been selected and assembled for the recording of fuel rod simulator-deformations during a loss of coolant accident using a movie technique. With this method it is possible to observe and record the ballooning of the simulator under conditions similar to those in a reactor. Some typical pictures are shown which show that the quality is high enough to allow a quantitative evaluation of the ballooning as a function of time. (orig.) [de

  18. Plasma Stability in Magnetic Mirror Machine with Stabilizing Rods; Stabilite du Plasma dans une Machine a Miroirs Magnetiques avec Barreaux de Stabilisation; Ustojchivost' plazmy v probkotrone so stabiliziruyushchimi sterzhnyami; Estabilidad del Plasma en una Trampa de Espejos con Barras Estabilizadoras

    Energy Technology Data Exchange (ETDEWEB)

    Trubnikov, B. A. [Institut Atomnoj Ehnergii Im. I.V. Kurchatova, Moskva, SSSR (Russian Federation)

    1966-04-15

    magnetic mirror machine. (In comparing theory with experiment it is assumed that the plasma is described by a Maxwellian distribution with a cut-out cone and it expands in length right up to the magnetic mirrors. To avoid numerical calculations, we discuss in this article the case x = z{sub max}/ Script-Small-L << 1; this is, however, extrapolated to x {approx} 1, which is admissible for semi-quantitative evaluations.) In addition to the infinite rods, we discuss stabilizing rods of finite length, which can be realized with multipole magnets of finite length. Stabilization can be achieved even in the extreme case of very short rods, where they can be regarded as a system of 2n magnetic point dipoles located in the central plane of the magnetic mirror machine on a circle of radius a, and directed along the radius away from the axis (or towards the axis - the directions alternate). Such systems, which we considered at the suggestion of Artsimovich, are very promising since their structure facilitates access to the plasma and makes it possible to install additional devices for diagnostics or for heating the plasma. (author) [French] L'auteur etend le critere bien connu Greek-Small-Letter-Delta {integral} Script-Small-L / B < 0 au cas arbitraire des champs a dissymetrie axiale. A cet effet, il convient d'exprimer ce critere par a formule: {integral}( Greek-Small-Letter-Delta B/B{sup 2}) (P{sub 0} + P{sub Up-Tack }) d Script-Small-L > 0; ou Greek-Small-Letter-Delta B = s x {Delta}B est l'accroissement du module du champ sur la ligne de force voisine (exterieure), s etant perpendicualire a B. Le critere mentionne est applique a l'etude de la stabilite du plasma dans une machine a miroirs magnetiques comportant des barreaux de stabilisation. Le champ au voisinage de l'axe est decrit par le potentiel scalaire: {psi} = {integral} B{sub 0} (z) dz - B{sub 0}{sup 1} (z) r 2/4 + g (z) r{sup n} cos n{phi} Pour le champ situe sur l'axe proprement dit, on admet l'approximation parabolique

  19. Calibration of langmuir probes by a microwave method; Etalonnage des sondes de langmuir par une methode hyperfrequence

    Energy Technology Data Exchange (ETDEWEB)

    Consoli, T [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

    1960-07-01

    Measurements of the electronic density of a plasma between 10{sup 6} and 10{sup 8} e/cm{sup 3}, made by the Langmuir probe and by resonance frequency shift of a cavity are compared. (author) [French] On compare les mesures de la densite electronique d'un plasma peu dense 10{sup 6} < ne < 10{sup 8} e/cm{sup 3}, par sonde de Langmuir et par glissement de la frequence de resonance d'une cavite contenant le plasma. (auteur)

  20. Study on heat transfer and hydraulic model of spiral-fin fuel rods based on equivalent annulus method

    International Nuclear Information System (INIS)

    Zhang Dan; Liu Changwen; Lu Jianchao

    2011-01-01

    Tight lattice fuel assembly usually adopts spiral-fin fuel elements. Compared with the traditional PWR fuel rods, the closely packed and spiral fin spacers make the heat transfer and hydraulic phenomena in sub-channels very complicated, and: there was no suitable model and correlation to study it. This paper studied the effect of spiral spacers on the channel geometry in the equivalent annulus and physical performance based on the Rehme equivalent annulus methods, and the heat transfer of the spiral fin fuel rods and hydraulic model were obtained. The new model was verified with the traditional one, and the verification showed that two new models agreed well, which could provide certain theoretical explanation to the effect of the spiral spacer on the thermal hydraulics. (authors)

  1. Replacement rod

    International Nuclear Information System (INIS)

    Hatfield, S.C.

    1989-01-01

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

  2. Correction method for critical extrapolation of control-rods-rising during physical start-up of reactor

    International Nuclear Information System (INIS)

    Zhang Fan; Chen Wenzhen; Yu Lei

    2008-01-01

    During physical start-up of nuclear reactor, the curve got by lifting the con- trol rods to extrapolate to the critical state is often in protruding shape, by which the supercritical phenomena is led. In the paper, the reason why the curve was in protruding was analyzed. A correction method was introduced, and the calculations were carried out by the practical data used in a nuclear power plant. The results show that the correction method reverses the protruding shape of the extrapolating curve, and the risk of reactor supercritical phenomena can be reduced using the extrapolated curve got by the correction method during physical start-up of the reactor. (authors)

  3. Uranium and thorium loadings determined by chemical and nondestructive methods in HTGR fuel rods for the Fort St. Vrain Early Validation Irradiation Experiment

    International Nuclear Information System (INIS)

    Angelini, P.; Rushton, J.E.

    1979-01-01

    The Fort St. Vrain Early Validation Irradiation Experiment is an irradiation test of reference and of improved High-Temperature Gas-Cooled Reactor fuels in the Fort St. Vrain Reactor. The irradiation test includes fuel rods fabricated at ORNL on an engineering scale fuel rod molding machine. Fuel rods were nondestructively assayed for 235 U content by a technique based on the detection of prompt-fission neutrons induced by thermal-neutron interrogation and were later chemically assayed by using the modified Davies Gray potentiometric titration method. The chemical analysis of the thorium content was determined by a volumetric titration method. The chemical assay method for uranium was evaluated and the results from the as-molded fuel rods agree with those from: (1) large samples of Triso-coated fissile particles, (2) physical mixtures of the three particle types, and (3) standard solutions to within 0.05%. Standard fuel rods were fabricated in order to evaluate and calibrate the nondestructive assay device. The agreement of the results from calibration methods was within 0.6%. The precision of the nondestructive assay device was established as approximately 0.6% by repeated measurements of standard rods. The precision was comparable to that estimated by Poisson statistics. A relative difference of 0.77 to 1.5% was found between the nondestructive and chemical determinations on the reactor grade fuel rods

  4. Cutting method and cutting device for spent fuel rod of nuclear reactor

    International Nuclear Information System (INIS)

    Komatsu, Masahiko; Ose, Toshihiko.

    1996-01-01

    A control rod transferred under water in a vertically suspended state is postured horizontally at such a water depth that radiations can be shielded, and then it is cut to a dropping speed limiting portion and a cross-like main body. The separated cross-like main body portion is further cut in the longitudinal direction and separated into a pair of cut pieces each having an L-shaped cross section. A disk like metal saw is used as a cutting tool. Alternatively, a plasma jet cutter or a melting-type water jet cutter is used as a cutting tool. Then, since the spent control rod to be cut is postured horizontally under water, the water depth for the cutting position can be reduced. As a result, the cutting state using the cutting tool can be observed by naked eyes from the position above the water surface thereby enabling to perform the cutting operation reliably. (N.H.)

  5. Method of forming magnetostrictive rods from rare earth-iron alloys

    Science.gov (United States)

    McMasters, O. Dale

    1986-09-02

    Rods of magnetrostructive alloys of iron with rare earth elements are formed by flowing a body of rare earth-iron alloy in a crucible enclosed in a chamber maintained under an inert gas atmosphere, forcing such molten rare-earth-iron alloy into a hollow mold tube of refractory material positioned with its lower end portion within the molten body by means of a pressure differential between the chamber and mold tube and maintaining a portion of the molten alloy in the crucible extending to a level above the lower end of the mold tube so that solid particles of higher melting impurities present in the alloy collect at the surface of the molten body and remain within the crucible as the rod is formed in the mold tube.

  6. Linear extrapolation distance for a black cylindrical control rod with the pulsed neutron method

    International Nuclear Information System (INIS)

    Loewenhielm, G.

    1978-03-01

    The objective of this experiment was to measure the linear extrapolation distance for a central black cylindrical control rod in a cylindrical water moderator. The radius for both the control rod and the moderator was varied. The pulsed neutron technique was used and the decay constant was measured for both a homogeneous and a heterogeneous system. From the difference in the decay constants the extrapolation distance could be calculated. The conclusion is that within experimental error it is safe to use the approximate formula given by Pellaud or the more exact one given by Kavenoky. We can also conclude that linear anisotropic scattering is accounted for in a correct way in the approximate formula given by Pellaud and Prinja and Williams

  7. METHODS FOR IMPROVING THE ENERGY EFFICIENCY OF WELL ROD PUMP UNITS

    Directory of Open Access Journals (Sweden)

    BRUNMAN1 Vladimir E.

    2016-11-01

    Full Text Available The concept of oil production energy efficiency improvement of good rod pumps by utilization of kinetic energy of the downward moving rod in capacitor bank is proposed. A mathematical model of the system is developed. Criteria of reduction of the peak values of current, consuming power and elimination of oscillations are obtained. It is shown that the developed system is capable of reducing the consumption of current twice and the peak power by three times. Thus it is possible to reduce operational and capital costs by reducing the cross-section of the feeder cables and decreasing the power of input transformers and diesel generator set if autonomous feeding of pumping units is used

  8. The photothermal camera - a new non destructive inspection tool; La camera photothermique - une nouvelle methode de controle non destructif

    Energy Technology Data Exchange (ETDEWEB)

    Piriou, M. [AREVA NP Centre Technique SFE - Zone Industrielle et Portuaire Sud - BP13 - 71380 Saint Marcel (France)

    2007-07-01

    techniques classiques de controle de surface (ressuage, magnetoscopie, courants de Foucault) pour: - detecter sans aucun contact, des defauts sous ligaments ou debouchants de quelques microns d'ouverture, sur des pieces metalliques non preparees (oxydees, usinees, soudees), - fonctionner sur des surfaces aux geometries variees, sur des pieces chaudes, sur des materiaux isolants (dielectriques), sans etre affectee par les proprietes magnetiques de la piece a examiner. Cette methode a permis, entre autre, de controler in situ les soudures 'tube/plaque' d'un echangeur intermediaire du reacteur rapide Phenix, de prouver qu'elle est une alternative au ressuage pour le controle des soudures de l'enceinte a vide d'ITER, de detecter des fissures dans les soudures (ex: J-weld d'adaptateurs de couvercles) et de reveler la fissuration amorcee par faiencage thermique. Les particularites de cette methode innovante sont: - de fonctionner a distance de la piece a controler, jusqu'a 2 metres, - d'etre totalement telecommandee a la distance de 15 metres (voire beaucoup plus par lien Ethernet), - d'etre une methode 'propre' puisqu'elle ne genere aucun dechet. Ces particularites en font une methode alternative au ressuage, afin de garantir la protection des operateurs et de l'environnement. (auteur)

  9. The pedicle screw-rod system is an acceptable method of reconstructive surgery after resection of sacroiliac joint tumours

    Directory of Open Access Journals (Sweden)

    Yi-Jun Zhou

    2016-03-01

    Full Text Available Hemipelvic resections for primary bone tumours require reconstruction to restore weight bearing along anatomic axes. However, reconstruction of the pelvic arch remains a major surgical challenge because of the high rate of associated complications. We used the pedicle screw-rod system to reconstruct the pelvis, and the purpose of this investigation was to assess the oncology, functional outcome and complication rate following this procedure. The purpose of this study was to investigate the operative indications and technique of the pedicle screw-rod system in reconstruction of the stability of the sacroiliac joint after resection of sacroiliac joint tumours. The average MSTS (Musculoskeletal Tumour Society score was 26.5 at either three months after surgery or at the latest follow-up. Seven patients had surgery-related complications, including wound dehiscence in one, infection in two, local necrosis in four (including infection in two, sciatic nerve palsy in one and pubic symphysis subluxation in one. There was no screw loosening or deep vein thrombosis occurring in this series. Using a pedicle screw-rod after resection of a sacroiliac joint tumour is an acceptable method of pelvic reconstruction because of its reduced risk of complications and satisfactory functional outcome, as well as its feasibility of reconstruction for type IV pelvis tumour resection without elaborate preoperative customisation. Level of evidence: Level IV, therapeutic study.

  10. Numerical Investigation of Cross Flow Phenomena in a Tight-Lattice Rod Bundle Using Advanced Interface Tracking Method

    Science.gov (United States)

    Zhang, Weizhong; Yoshida, Hiroyuki; Ose, Yasuo; Ohnuki, Akira; Akimoto, Hajime; Hotta, Akitoshi; Fujimura, Ken

    In relation to the design of an innovative FLexible-fuel-cycle Water Reactor (FLWR), investigation of thermal-hydraulic performance in tight-lattice rod bundles of the FLWR is being carried out at Japan Atomic Energy Agency (JAEA). The FLWR core adopts a tight triangular lattice arrangement with about 1 mm gap clearance between adjacent fuel rods. In view of importance of accurate prediction of cross flow between subchannels in the evaluation of the boiling transition (BT) in the FLWR core, this study presents a statistical evaluation of numerical simulation results obtained by a detailed two-phase flow simulation code, TPFIT, which employs an advanced interface tracking method. In order to clarify mechanisms of cross flow in such tight lattice rod bundles, the TPFIT is applied to simulate water-steam two-phase flow in two modeled subchannels. Attention is focused on instantaneous fluctuation characteristics of cross flow. With the calculation of correlation coefficients between differential pressure and gas/liquid mixing coefficients, time scales of cross flow are evaluated, and effects of mixing section length, flow pattern and gap spacing on correlation coefficients are investigated. Differences in mechanism between gas and liquid cross flows are pointed out.

  11. Water Assisted Growth of C60 Rods and Tubes by Liquid–Liquid Interfacial Precipitation Method

    Directory of Open Access Journals (Sweden)

    Cheuk-Wai Tai

    2012-06-01

    Full Text Available C60 nanorods with hexagonal cross sections are grown using a static liquid–liquid interfacial precipitation method in a system of C60/m-dichlorobenzene solution and ethanol. Adding water to the ethanol phase leads instead to C60 tubes where both length and diameter of the C60 tubes can be controlled by the water content in the ethanol. Based on our observations we find that the diameter of the rods/tubes strongly depends on the nucleation step. We propose a liquid-liquid interface growth model of C60 rods and tubes based on the diffusion rate of the good C60 containing solvent into the poor solvent as well as on the size of the crystal seeds formed at the interface between the two solvents. The grown rods and tubes exhibit a hexagonal solvate crystal structure with m-dichlorobenzene solvent molecules incorporated into the crystal structure, independent of the water content. An annealing step at 200 °C at a pressure < 1 kPa transforms the grown structures into a solvent-free face centered cubic structure. Both the hexagonal and the face centered cubic structures are very stable and neither morphology nor structure shows any signs of degradation after three months of storage.

  12. Direct determination of enthalpies of solid phase reactions by immersion method; Determination directe des enthalpies de reaction en phase solide par une methode de plongee

    Energy Technology Data Exchange (ETDEWEB)

    Roux, A; Richard, M; Eyraud, L; Stevanovic, M; Elston, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1967-07-01

    It is not generally possible to measure the enthalpy change corresponding to solid phase reactions using the dynamic differential thermal analysis method because these reactions are usually too slow at the temperature of operation of present equipment. A ballistic differential thermal analysis apparatus has been developed which is based on an immersion-compensation method; it overcomes the difficulties previously encountered. This apparatus has been used after calibration for determining the enthalpies of formation of calcium and cadmium titanates. and also the Wigner energies of BeO, MgO and Al{sub 2}O{sub 3} samples irradiated at variable dose at a temperature of under 100 deg. C. (authors) [French] Il n'est generalement pas possible de mesurer la variation d'enthalpie correspondant aux reactions en phase solide par la methode d'analyse thermique differentielle dynamique. En effet, ces reactions sont le plus souvent trop lentes aux temperatures d'utilisation des dispositifs actuels. Un appareil d'analyse thermique differentielle balistique, base sur une methode de plongee avec compensation, a ete mis au point et permet de surmonter les difficultes precedentes. Apres etalonnages, cet appareil a ete utilise pour la determination des enthalpies de formation du titanate de calcium et du titanate de cadmium ainsi que pour celle des energies Wigner emmagasinees dans des echantillons de BeO, MgO et Al{sub 2}O{sub 3} irradies a une temperature inferieure a 100 deg. C et a differentes doses. (auteurs)

  13. Water rod

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  14. Equilibrium Configurations of the Noncircular Cross-Section Elastic Rod Model with the Elliptic KB Method

    Directory of Open Access Journals (Sweden)

    Yongzhao Wang

    2015-01-01

    Full Text Available The mechanical deformation of DNA is very important in many biological processes. In this paper, we consider the reduced Kirchhoff equations of the noncircular cross-section elastic rod characterized by the inequality of the bending rigidities. One family of exact solutions is obtained in terms of rational expressions for classical Jacobi elliptic functions. The present solutions allow the investigation of the dynamical behavior of the system in response to changes in physical parameters that concern asymmetry. The effects of the factor on the DNA conformation are discussed. A qualitative analysis is also conducted to provide valuable insight into the topological configuration of DNA segments.

  15. Method for the protection of the cladding tubes of fuel rods

    International Nuclear Information System (INIS)

    Steinberg, E.

    1978-01-01

    To present stress crack corrosion and to protect the cladding tubes of the fuel rods made of a circonium alloy from attack by iodine, the inward surfaces are provided with protective coatings. Therefore the casting tubes already filled with fuel element pellets are put under over-pressure at a temperature range between 300 and 500 0 C, until almost yield-point is reached. A small amount of H 2 O or H 2 O 2 , filled in, reacts with the cladding tube material to form the Zr-O 2 protective coating. Afterwards comes a pressure relief, and the cladding tube reaches its original dimensions. (DG) [de

  16. Calculation of control rods in rectangular reactor, and applications (1960); Calcul des barres de conteole dans un reacteur rectangulaire et applications (1960)

    Energy Technology Data Exchange (ETDEWEB)

    Goshen, S; Pazy, A [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

    1960-07-01

    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) [French] Le but de ce rapport est de trouver une methode pour estimer l'antireactivite des barres de controle perpendiculaires a l'axe dans pile cylindrique. Le rapport se divise en deux parties. Dans la premiere nous donnons une methode de calcul des barres de controle dans une pile rectangulaire, analogue a la methode de Nordheim-Scalettar pour les piles cylindriques. A titre d'exemple, nous donnons les formules de theories a un et deux groupes de neutrons, la generalisation pour plusieurs groupes est evidente. Dans la deuxieme partie, nous trouvons, par une methode de variation, une formule qui permet d'estimer le laplacien d'une pile, qui peut etre divisee en parallelepipedes dont les laplaciens sont donnes. Nous utilisons enfin, cette formule pour calculer l'antireactivite des barres perpendiculaires a l'axe dans une pile cylindrique. (auteur)

  17. Determination of the most reactivity control rod by pseudo-harmonics perturbation method; Determinacao da barra de controle mais reativa usando o metodo de pseudo-harmonicos

    Energy Technology Data Exchange (ETDEWEB)

    Freire, Fernando S.; Silva, Fernando C.; Martinez, Aquilino S. [Universidade Federal, Rio de Janeiro, RJ (Brazil). Coordenacao dos Programas de Pos-graduacao de Engenharia. Programa de Engenharia Nuclear]. E-mail: ffreire@con.ufrj.br; fernando@con.ufrj.br; aquilino@.con.ufrj.br

    2005-07-01

    Frequently it is necessary to compute the change in core multiplication caused by a change in the core temperature or composition. Even when this perturbation is localized, such as a control rod inserted into the core, one does not have to repeat the original criticality calculation, but instead we can use the well-known pseudo-harmonics perturbation method to express the corresponding change in the multiplication factor in terms of the neutron flux expanded in the basis vectors characterizing the unperturbed core. Therefore we may compute the control rod worth to find the most reactivity control rod to calculate the fast shutdown margin. In this thesis we propose a simple and precise method to identify the most reactivity control rod. (author)

  18. Preparation of 1 Ci of europium 155 without carrier; Une methode de production d'europium 155 sans entraineur au niveau du curie

    Energy Technology Data Exchange (ETDEWEB)

    Falconi, N; Radicella, R [Commissariat a l' Energie Atomique, Dir. des Materiaux et Combustibles Nucleaires, Saclay (France). Centre d' Etudes Nucleaires

    1968-07-01

    High activity 'point' sources of {sup 155}Eu are used for medical and industrial applications. For this purpose we have studied a method of obtaining I Ci of carrier free {sup 155}Eu, with a solid residue smaller than 5 mg per Ci. In order to separate the {sup 155}Eu from several grams of a {sup 154}Sm enriched target we propose a procedure which is based on the work of Bouissieres and David, Onstott, and Takekoshi et al. The separation is carried out by electrolysis on a mercury cathode followed by purification on ion exchange resin. The yields of the europium separation and target recovery are 80 per cent and 90 per cent respectively. The time required for the procedure is three days. (authors) [French] L'obtention de sources ponctuelles de haute activite d'europium-l55 est interessante tant pour des usages medicaux qu'industriels. Nous avons cherche le moyen de preparer en une seule fois, une solution d'un curie d'europium-155 sans entraineur presentant un extrait sec inferieur a 5 mg par curie. En se basant sur des travaux de Bouissieres et Davis, d'Onstott, et de Takekoshi et coll., nous avons mis au point une methode permettant de separer l'europium-155 de sa cible de samarium enrichi en {sup 154}Sm par une electrolyse sur cathode de mercure suivie d'une purification sur resine echangeuse d'ions. Les rendements chimiques de separation de l'europium et de recuperation de la cible de samarium enrichi sont respectivement de 80 pour cent et 90 pour cent. La duree de l'operation est de 3 jours. (auteurs)

  19. 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.)

  20. The development of the measurement technique of the control rod worth with the inverse kinetics method considering the influence of the steady neutron source

    International Nuclear Information System (INIS)

    Takeuchi, Mitsuo; Wada, Shigeru; Takahashi, Hiroyuki; Hayashi, Kazuhiko; Murayama, Yoji

    2000-09-01

    At the research reactor such as JRR-3M, the operation management is carried out in order to ensure safe operation, for example, the excess reactivity is measured regularly and confirmed that it satisfies a safety condition. The excess reactivity is calculated using control rod position in criticality and control rod worth measured by a positive period method (P.P method), the conventional inverse kinetic method (IK method) and so on. The neutron source, however, influences measurement results and brings in a measurement error. A new IK method considering the influence of the steady neutron sources is proposed and applied to the JRR-3M. This report shows that the proposed IK method measures control rod worth more precisely than a conventional IK method. (author)

  1. Effects of TiO{sub 2} buffer layer on the photoelectrochemical properties of TiO{sub 2} Nano rods grown by modified chemical bath deposition method

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Tae-hyun; Ha, Jin-wook; Ryu, Hyukhyun [Inje University, Gimhae (Korea, Republic of); Lee, Won-Jae [Dong-Eui University, Busan (Korea, Republic of)

    2015-08-15

    In this study, we grew TiO{sub 2} nano rods on TiO{sub 2}-film buffered FTO substrate using modified chemical bath deposition (M-CBD). The TiO{sub 2} buffer layer was grown by spin coating method with different RPM (revolutions per minute) values and deposition cycles. We investigated the effects of the RPM values and the deposition cycles on the morphological, structural and photoelectrochemical properties of TiO{sub 2} nano rods. In this work, we have also found that the morphological and structural properties of TiO{sub 2} nano rods affected the photoelectrochemical properties of TiO{sub 2} nano rods. And the maximum photocurrent density of 0.34 mA/cm{sup 2} at 0.6V (vs.SCE) was obtained from the buffer layer deposition process condition of 4,000 RPM and two-times buffer layer depositions.

  2. Acceleration Test Method for Failure Prediction of the End Cap Contact Region of Sodium Cooled Fast Reactor Fuel Rod

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyung-Kyu; Lee, Young-Ho; Lee, Hyun-Seung; Lee, Kang-Hee [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2017-05-15

    This paper reports the results of an acceleration test to predict the contact-induced failure that could occur at the cylinder-to-hole joint for the fuel rod of a sodium-cooled fast reactor (SFR). To incorporate the fuel life of the SFR currently under development at KAERI (around 35,000 h), the acceleration test method of reliability engineering was adopted in this work. A finite element method was used to evaluate the flow-induced vibration frequency and amplitude for the test parameter values. Five specimens were tested. The failure criterion during the life of the SFR fuel was applied. The S-N curve of the HT-9, the material of concern, was used to obtain the acceleration factor. As a result, a test time of 16.5 h was obtained for each specimen. It was concluded that the B{sub 0.004} life would be guaranteed for the SFR fuel rods with 99% confidence if no failure was observed at any of the contact surfaces of the five specimens.

  3. An analytical method for the calculation of static characteristics of linear step motors for control rod drives in nuclear reactors

    International Nuclear Information System (INIS)

    Khan, S.H.; Ivanov, A.A.

    1995-01-01

    An analytical method for calculating static characteristics of linear dc step motors (LSM) is described. These multiphase passive-armature motors are now being developed for control rod drives (CRD) in large nuclear reactors. The static characteristics of such LSM is defined by the variation of electromagnetic force with armature displacement and it determines motor performance in its standing and dynamic modes of operation. The proposed analytical technique for calculating this characteristic is based on the permeance analysis method applied to phase magnetic circuits of LSM. Reluctances of various parts of phase magnetic circuit is calculated analytically by assuming probable flux paths and by taking into account complex nature of magnetic field distribution in it. For given armature positions stator and armature iron saturations are taken into account by an efficient iterative algorithm which gives fast convergence. The method is validated by comparing theoretical results with experimental ones which shows satisfactory agreement for small stator currents and weak iron saturation

  4. Calculation of the shock-wave in the region close to an underground nuclear explosion (method Cades); Calcul de l'onde de choc en zone proche d'une explosion nucleaire souterraine (methode cades)

    Energy Technology Data Exchange (ETDEWEB)

    Supiot, F; Brugies, J [Commissariat a l' Energie Atomique, Bruyeres-le-Chatel (France). Centre d' Etudes

    1969-07-01

    The outline of a method is presented for calculating the characteristics of a shock wave produced by an underground nuclear explosion (pressure, wave velocity, velocity of the medium, energy left in the medium by the shock, etc.). By means of an application to a granitic medium and of a comparison with results obtained during French nuclear explosions, it has been possible to show the good agreement existing between the calculations and the experimental results. The advantages of such a method for studying the industrial applications of underground nuclear explosions are stressed. (authors) [French] On expose les grandes lignes d'une methode de calcul des caracteristiques de l'onde de choc issue d'une explosion nucleaire souterraine (pression, vitesse de l'onde, vitesse du milieu, energie deposee par le choc dans le milieu...). Une application a un milieu granitique et une comparaison aux resultats obtenus au cours d'explosions nucleaires francaises permet de montrer la bonne concordance entre le calcul et les resultats experimentaux. On souligne l'interet d'une telle, methode pour l'etude d'applications industrielles des explosions nucleaires souterraines. (auteurs)

  5. Critical mass, rod values and reactivity coefficients for Rapsodie; Masse critique, valeur des barres et coefficients de reactivite de rapsodie

    Energy Technology Data Exchange (ETDEWEB)

    Stevens, L; Gourdon, J [Commissariat a l' Energie Atomique, Cadarache (France). Centre d' Etudes Nucleaires

    1967-07-01

    Besides a brief general description, the report contains a description and discussion of the aims, the methods used and the results of critical mass, rod worth and static reactivity coefficient measurements on the Rapsodie reactor. (authors) [French] Apres une breve description generale, le rapport decrit et discute le but, les methodes employees et les resultats des mesures de masse critique, de reactivite des barres et des coefficients de reactivite statiques du reacteur RAPSODIE. (auteurs)

  6. A fast position estimation method for a control rod guide tube inspection robot with a single camera

    International Nuclear Information System (INIS)

    Lee, Jae C.; Seop, Jun H.; Choi, Yu R.; Kim, Jae H.

    2004-01-01

    One of the problems in the inspection of control rod guide tubes using a mobile robot is accurate estimation of the robot's position. The problem is usually explained by the question 'Where am I?'. We can solve this question by a method called dead reckoning using odometers. But it has some inherent drawbacks such that the position error grows without bound unless an independent reference is used periodically to reduce the errors. In this paper, we presented one method to overcome this drawback by using a vision sensor. Our method is based on the classical Lucas Kanade algorithm for on image tracking. In this algorithm, an optical flow must be calculated at every image frame, thus it has intensive computing load. In order to handle large motions, it is preferable to use a large integration window. But a small integration window is more preferable to keep the details contained in the images. We used the robot's movement information obtained from the dead reckoning as an input parameter for the feature tracking algorithm in order to restrict the position of an integration window. By means of this method, we could reduce the size of an integration window without any loss of its ability to handle large motions and could avoid the trade off in the accuracy. And we could estimate the position of our robot relatively fast without on intensive computing time and the inherent drawbacks mentioned above. We studied this algorithm for applying it to the control rod guide tubes inspection robot and tried an inspection without on operator's intervention

  7. Fuel rods

    International Nuclear Information System (INIS)

    Hattori, Shinji; Kajiwara, Koichi.

    1980-01-01

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

  8. Rodding Surgery

    Science.gov (United States)

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

  9. 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.)

  10. The Carbapenem Inactivation Method (CIM), a Simple and Low-Cost Alternative for the Carba NP Test to Assess Phenotypic Carbapenemase Activity in Gram-Negative Rods

    NARCIS (Netherlands)

    Zwaluw, K. van der; Haan, A. de; Pluister, G.N.; Bootsma, H.J.; Neeling, A.J. de; Schouls, L.M.

    2015-01-01

    A new phenotypic test, called the Carbapenem Inactivation Method (CIM), was developed to detect carbapenemase activity in Gram-negative rods within eight hours. This method showed high concordance with results obtained by PCR to detect genes coding for the carbapenemases KPC, NDM, OXA-48, VIM, IMP

  11. An outline of the methods which will be used to evaluate shutdown margins, control rod worth and local criticality problems for the HTR - power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Dworak, A

    1971-09-28

    The methods to calculate control rod efficiency outlined in this paper were the results of an investigation carried out by A. Dworak and C. Hunt on the Dragon Charge I Core 1. There the theoretical results agreed very good with the experimental measurements. This paper gives a brief outline of the methods in question and summarizes the calculations done up to now.

  12. Deposition of very thin uniform indium sulfide layers over metallic nano-rods by the Spray-Ion Layer Gas Reaction method

    Energy Technology Data Exchange (ETDEWEB)

    Genduso, G. [Dipartimento di Ingegneria Chimica, Gestionale, Informatica, Meccanica, Università di Palermo, Viale delle Scienze, 90100 Palermo (Italy); Institut for Heterogeneous Material Systems, Helmholtz-Zentrum Berlin für Materialien und Energie GmbH, Hahn-Meitner-Platz 1, D-14109 Berlin (Germany); Inguanta, R.; Sunseri, C.; Piazza, S. [Dipartimento di Ingegneria Chimica, Gestionale, Informatica, Meccanica, Università di Palermo, Viale delle Scienze, 90100 Palermo (Italy); Kelch, C.; Sáez-Araoz, R. [Institut for Heterogeneous Material Systems, Helmholtz-Zentrum Berlin für Materialien und Energie GmbH, Hahn-Meitner-Platz 1, D-14109 Berlin (Germany); Zykov, A. [Institut for Heterogeneous Material Systems, Helmholtz-Zentrum Berlin für Materialien und Energie GmbH, Hahn-Meitner-Platz 1, D-14109 Berlin (Germany); present address: Institut für Physik, Humboldt-Universität zu Berlin, Newtonstr. 15,12489 Berlin (Germany); Fischer, Ch.-H., E-mail: fischer@helmholtz-berlin.de [Institut for Heterogeneous Material Systems, Helmholtz-Zentrum Berlin für Materialien und Energie GmbH, Hahn-Meitner-Platz 1, D-14109 Berlin (Germany); second affiliation: Free University Berlin, Chemistry Institute, Takustr. 3, D-14195 Berlin (Germany)

    2013-12-02

    Very thin and uniform layers of indium sulfide were deposited on nickel nano-rods using the sequential and cyclical Spray-ILGAR® (Ion Layer Gas Reaction) technique. Substrates were fabricated by electrodeposition of Ni within the pores of polycarbonate membranes and subsequent chemical dissolution of the template. With respect to the depositions on flat substrates, experimental conditions were modified and optimized for the present geometry. Our results show that nano-rods up to a length of 10 μm were covered uniformly along their full length and with an almost constant film growth rate, thus allowing a good control of the coating thickness; the effect of the deposition temperature was also investigated. However, for high numbers of process steps, i.e. thickness, the films became uneven and crusty, especially at higher temperature, mainly owing to the simultaneous side reaction of the metallic Ni forming nickel sulfide at the surface of the rods. However, such a problem occurs only in the case of reactive nano-rod materials, such as less noble metals. It could be strongly reduced by doubling the spray step duration and thereby sealing the metallic surface before the process step of the sulfurization. Thus, quite smooth, about 100 nm thick coatings could be obtained. - Highlights: • Ni nano-rod substrates were grown within polycarbonate membranes. • We can coat nano-rods uniformly by the Ion Layer Gas Reaction method. • As a model we deposited up to about 100 nm In{sub 2}S{sub 3} on Ni nanorods (250 nm × 10 μm). • Element mapping at insulated rods showed homogenous coating over the full length. • Parameter optimization reduced effectively the Ni sulfide formation.

  13. Great improvement on tetracycline removal using ZnO rod-activated carbon fiber composite prepared with a facile microwave method

    Energy Technology Data Exchange (ETDEWEB)

    Tran Thi, Viet Ha; Lee, Byeong-Kyu, E-mail: bklee@ulsan.ac.kr

    2017-02-15

    Highlights: • ZnO rod-ACF was prepared by a method involving a microwave within only 3 min. • ZnO rods (average diameter of 0.3–0.5 μm, length of 1.0–1.5 μm) were grown on ACF. • 99% of tetracycline was degraded and 90.7% TOC was removed within 1 h under UV light. • ZnO rod-ACF achieved high performances even after three cycles of uses. - Abstract: New composite materials of activated carbon fiber (ACF) coated with zinc oxide (ZnO) were obtained by applying a green, cost-effective and rapid synthetic route using a commercial microwave oven. ZnO rods with a uniform and stable structure and an average diameter of 0.3–0.5 μm and length of 1.0–1.5 μm were achieved after only 3-min microwave treatment. The properties of ZnO were efficiently transferred to ACF, such that the resulting material, termed ZnO rod-ACF, demonstrated a promising potential as an efficient photocatalyst and simultaneously as an adsorbent. Pharmaceutical tetracycline at a concentration of 40 mg/L was used to evaluate the organic pollutant removal capacity of the synthesized materials. At pH 8, ZnO rod-ACF exhibited excellent removal capacity (over 99%) and mineralization (90.7%) of tetracycline in aqueous solution within 1 h under UV irradiation. The stability of ZnO rod-ACF was maintained and the mineralization of tetracycline was also maintained at 81.35% after multiple usage cycles. The photodegradation pathways of tetracycline were proposed based on the identified reaction intermediates.

  14. 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.)

  15. 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.)

  16. A method based on a separation of variables in magnetohydrodynamics (MHD); Une methode de separation des variables en magnetohydrodynamique

    Energy Technology Data Exchange (ETDEWEB)

    Cessenat, M.; Genta, P.

    1996-12-31

    We use a method based on a separation of variables for solving a system of first order partial differential equations, in a very simple modelling of MHD. The method consists in introducing three unknown variables {phi}1, {phi}2, {phi}3 in addition of the time variable {tau} and then searching a solution which is separated with respect to {phi}1 and {tau} only. This is allowed by a very simple relation, called a `metric separation equation`, which governs the type of solutions with respect to time. The families of solutions for the system of equations thus obtained, correspond to a radial evolution of the fluid. Solving the MHD equations is then reduced to find the transverse component H{sub {Sigma}} of the magnetic field on the unit sphere {Sigma} by solving a non linear partial differential equation on {Sigma}. Thus we generalize ideas due to Courant-Friedrichs and to Sedov on dimensional analysis and self-similar solutions. (authors).

  17. A new deconvolution method applied to ultrasonic images; Etude d'une methode de deconvolution adaptee aux images ultrasonores

    Energy Technology Data Exchange (ETDEWEB)

    Sallard, J

    1999-07-01

    This dissertation presents the development of a new method for restoration of ultrasonic signals. Our goal is to remove the perturbations induced by the ultrasonic probe and to help to characterize the defects due to a strong local discontinuity of the acoustic impedance. The point of view adopted consists in taking into account the physical properties in the signal processing to develop an algorithm which gives good results even on experimental data. The received ultrasonic signal is modeled as a convolution between a function that represents the waveform emitted by the transducer and a function that is abusively called the 'defect impulse response'. It is established that, in numerous cases, the ultrasonic signal can be expressed as a sum of weighted, phase-shifted replicas of a reference signal. Deconvolution is an ill-posed problem. A priori information must be taken into account to solve the problem. The a priori information translates the physical properties of the ultrasonic signals. The defect impulse response is modeled as a Double-Bernoulli-Gaussian sequence. Deconvolution becomes the problem of detection of the optimal Bernoulli sequence and estimation of the associated complex amplitudes. Optimal parameters of the sequence are those which maximize a likelihood function. We develop a new estimation procedure based on an optimization process. An adapted initialization procedure and an iterative algorithm enables to quickly process a huge number of data. Many experimental ultrasonic data that reflect usual control configurations have been processed and the results demonstrate the robustness of the method. Our algorithm enables not only to remove the waveform emitted by the transducer but also to estimate the phase. This parameter is useful for defect characterization. At last the algorithm makes easier data interpretation by concentrating information. So automatic characterization should be possible in the future. (author)

  18. A method to determine the dampening system of control rod drop mechanism for PWR reactors

    International Nuclear Information System (INIS)

    Trindade, C.E.; Mattos, J.R.L. de; Perrotta, J.A.

    1988-08-01

    A method to determine the Control Assembly damping drop system (dashpot/guide tube) was developed. It's presented a theoretical model, an experimental device and the procedures to determine this system, which is used in PWR reactors. (author) [pt

  19. A high-order boundary integral method for surface diffusions on elastically stressed axisymmetric rods

    OpenAIRE

    Li, Xiaofan; Nie, Qing

    2009-01-01

    Many applications in materials involve surface diffusion of elastically stressed solids. Study of singularity formation and long-time behavior of such solid surfaces requires accurate simulations in both space and time. Here we present a high-order boundary integral method for an elastically stressed solid with axi-symmetry due to surface diffusions. In this method, the boundary integrals for isotropic elasticity in axi-symmetric geometry are approximated through modified alternating quadratu...

  20. A high-order boundary integral method for surface diffusions on elastically stressed axisymmetric rods.

    Science.gov (United States)

    Li, Xiaofan; Nie, Qing

    2009-07-01

    Many applications in materials involve surface diffusion of elastically stressed solids. Study of singularity formation and long-time behavior of such solid surfaces requires accurate simulations in both space and time. Here we present a high-order boundary integral method for an elastically stressed solid with axi-symmetry due to surface diffusions. In this method, the boundary integrals for isotropic elasticity in axi-symmetric geometry are approximated through modified alternating quadratures along with an extrapolation technique, leading to an arbitrarily high-order quadrature; in addition, a high-order (temporal) integration factor method, based on explicit representation of the mean curvature, is used to reduce the stability constraint on time-step. To apply this method to a periodic (in axial direction) and axi-symmetric elastically stressed cylinder, we also present a fast and accurate summation method for the periodic Green's functions of isotropic elasticity. Using the high-order boundary integral method, we demonstrate that in absence of elasticity the cylinder surface pinches in finite time at the axis of the symmetry and the universal cone angle of the pinching is found to be consistent with the previous studies based on a self-similar assumption. In the presence of elastic stress, we show that a finite time, geometrical singularity occurs well before the cylindrical solid collapses onto the axis of symmetry, and the angle of the corner singularity on the cylinder surface is also estimated.

  1. A new nodal kinetics method for analyzing fast control rod motions in nuclear reactor cores

    International Nuclear Information System (INIS)

    Kaya, S.; Yavuz, H.

    2001-01-01

    A new nodal kinetics approach is developed for analyzing large reactivity accidents in nuclear reactor cores. This method shows promising that it has capability of inspecting promt criticality transients and it gives comparable results with respect to those of other techniques. (orig.)

  2. Sucker rods

    Energy Technology Data Exchange (ETDEWEB)

    Rylov, B M; Kostur, I N; Shcheigiy, B I; Sukhanov, V S

    1983-01-01

    As an addendum to A.s. USSR patent No 769087, this particular sucker rod utilizes a differential piston spring that has been attached outside the body of the auxiliary pump. The pump cylinder is attached to the intake line of the main pump. The lower part of the auxiliary pump is equipped with vertical slits, while the differential piston is equipped with a perforated pusher and support under the spring; it can also be shifted as necessary with respect to the vertical slits.

  3. An extraction method of uranium 233 from the thorium irradiates in a reactor core; Une methode d'extraction de l'uranium-233 a partir du thorium irradie dans une pile

    Energy Technology Data Exchange (ETDEWEB)

    Chesne, A; Regnaut, P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1955-07-01

    Description of the conditions of separation of the thorium, of the uranium 233 and of the protactinium 233 in hydrochloric solution by absorption then selective elution on anion exchange resin. A precipitation of the thorium by the oxalic acid permits the recuperation of the hydrochloric acid which is recycled, the main, raw material consumed being the oxalic acid. (authors) [French] Description des conditions de separation du thorium, de l'uranium 233 et du protactinium 233 en solution chlorhydrique par absorption puis elution selective sur resine echangeuse d'anions. Une precipitation du thoriun par l'acide oxalique permet la recuperation de l'acide chlorhydrique qui est recycle, la principale matiere premiere consommee etant l'acide oxalique. (auteurs)

  4. Non-destructive methods of control of thermo-physical properties of fuel rods

    International Nuclear Information System (INIS)

    Kruglov, A B; Kruglov, V B; Kharitonov, V S; Struchalin, P G; Galkin, A G

    2017-01-01

    Information about the change of thermal properties of the fuel elements needed for a successful and safe operation of the nuclear power plant. At present, the existing amount of information on the fuel thermal conductivity change and “fuel-shell” thermal resistance is insufficient. Also, there is no technique that would allow for the measurement of these properties on the non-destructive way of irradiated fuel elements. We propose a method of measuring the thermal conductivity of the fuel in the fuel element and the contact thermal resistance between the fuel and the shell without damaging the integrity of the fuel element, which is based on laser flash method. The description of the experimental setup, implementing methodology, experiments scheme. The results of test experiments on mock-ups of the fuel elements and their comparison with reference data, as well as the results of numerical modeling of thermal processes that occur during the measurement. Displaying harmonization of numerical calculation with the experimental thermograms layout shell portions of the fuel cell, confirming the correctness of the calculation model. (paper)

  5. Growth of Yb3+-doped Y2O3 single crystal rods by the micro-pulling-down method

    International Nuclear Information System (INIS)

    Mun, J.H.; Novoselov, A.; Yoshikawa, A.; Boulon, G.; Fukuda, T.

    2005-01-01

    The rare-earth sesquioxides (RE 2 O 3 , RE = Lu, Y and Sc) are very promising host crystals for advanced laser diode (LD)-pumped Yb 3+ -doped solid-state lasers due to unusual combination, almost unique of favourable structural, thermal and spectroscopic properties which are described. In spite of these favourable properties, the bulk single crystal growth technology for the rare-earth sesquioxides has not been established yet. The extremely high melting temperature at around 2400 deg. C has prevented it. However, we shall show that yttrium oxide crystals (Yb x Y 1-x ) 2 O 3 , x = 0.0, 0.005, 0.05, 0.08 and 0.15 of cylindrical shape as laser rods with 4.2 mm in diameter and 15-20 mm in length have been grown from rhenium crucibles by the micro-pulling-down method. The crystal quality characterisation of undoped Y 2 O 3 crystal was determined using X-ray rocking curve (XRC) analysis. Yb were homogeneously distributed in Y 2 O 3 host crystal

  6. Improvements in or relating to methods of and apparatus for coating wire, rod or strip material by sputtering

    International Nuclear Information System (INIS)

    Wareing, J.B.

    1976-01-01

    A method and apparatus are described for coating wire, rod or strip material comprising first subjecting the material to electron bombardment in a glow discharge to heat and activate the surface and then subjecting it to sputtering by use of a soft cathode discharge. The apparatus comprises a low pressure gas chamber through which the material is passed, and containing a glow discharge electron gun having a tubular cathode shaped so that the material can be passed axially through it, and an anode surrounding the cathode. The cathode is formed in two parts, the first part at one end, being made of material of low sputtering yield, and the second part being formed at least partially of the required coating material. The first part of the cathode may be of stainless steel or Al. The two parts of the cathode are electrically isolated with means provided for applying a lower negative potential, with respect to the anode, to the second part compared with the first part. The voltage applied to the second part may be controlled so as to control the sputtering rate. The gas pressure in the chamber is also controllable. The coating material may be arranged as inserts in the fixed cathode structure or as segments around the surface to be coated, and may be composed of Pb, Zn or Cu. (U.K.)

  7. 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

  8. Fuel rods

    International Nuclear Information System (INIS)

    Adachi, Hajime; Ueda, Makoto

    1985-01-01

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

  9. 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

  10. Two-dimensional transient thermal analysis of a fuel rod by finite volume method

    Energy Technology Data Exchange (ETDEWEB)

    Costa, Rhayanne Yalle Negreiros; Silva, Mário Augusto Bezerra da; Lira, Carlos Alberto de Oliveira, E-mail: ryncosta@gmail.com, E-mail: mabs500@gmail.com, E-mail: cabol@ufpe.br [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil). Departamento de Energia Nuclear

    2017-07-01

    One of the greatest concerns when studying a nuclear reactor is the warranty of safe temperature limits all over the system at all time. The preservation of core structure along with the constraint of radioactive material into a controlled system are the main focus during the operation of a reactor. The purpose of this paper is to present the temperature distribution for a nominal channel of the AP1000 reactor developed by Westinghouse Co. during steady-state and transient operations. In the analysis, the system was subjected to normal operation conditions and then to blockages of the coolant flow. The time necessary to achieve a new safe stationary stage (when it was possible) was presented. The methodology applied in this analysis was based on a two-dimensional survey accomplished by the application of Finite Volume Method (FVM). A steady solution is obtained and compared with an analytical analysis that disregard axial heat transport to determine its relevance. The results show the importance of axial heat transport consideration in this type of study. A transient analysis shows the behavior of the system when submitted to coolant blockage at channel's entrance. Three blockages were simulated (10%, 20% and 30%) and the results show that, for a nominal channel, the system can still be considerate safe (there's no bubble formation until that point). (author)

  11. Sensitivity and rapidity of the evaporation method for the measurement of the radioactivity of residual water; Sensibilite et rapidite de la methode d'evaporation pour la mesure de la radioactivite d'une eau residuaire

    Energy Technology Data Exchange (ETDEWEB)

    Cohen, P; Reiffsteck, A; Wormser, G [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1958-07-01

    The authors have used the evaporation method for counting the radioactivity of water polluted to the tolerance limit by Sr{sup 90}. Total duration of the manipulation is 2 hours. (author) [French] Les auteurs ont utilise la methode d'evaporation pour le comptage de la radioactivite d'une eau polluee, a la limite de tolerance, par du {sup 90}Sr. La duree de la manipulation totale est de 2 h. (auteur)

  12. Modelling of pressurized water reactor fuel, rod time dependent radial heat flow with boundary element method; Modeliranje spremenljivega radijalnega toplotnega toka tlacnovodne gorivne palice z metodo robnih elementov

    Energy Technology Data Exchange (ETDEWEB)

    Sarler, B [Institut Jozef Stefan, Ljubljana (Yugoslavia)

    1987-07-01

    The basic principles of the boundary element method numerical treatment of the radial flow heat diffusion equation are presented. The algorithm copes the time dependent Dirichlet and Neumann boundary conditions, temperature dependent material properties and regions from different materials in thermal contact. It is verified on the several analytically obtained test cases. The developed method is used for the modelling of unsteady radial heat flow in pressurized water reactor fuel rod. (author)

  13. A non-destructive, ultrasonic method for the determination of internal pressure and gas composition in an LWR fuel rod on-going and future programme

    International Nuclear Information System (INIS)

    Ferrandis, J.; Leveque, G.; Villard, J.

    2006-01-01

    Several possible non-destructive methods have been investigated in the past to measure the internal gas pressure e.g., measurement of 85 Kr directly, or after accumulation in the plenum by freezing with liquid nitrogen. However no satisfactory resolution to the problem has been found, so at present there is no rapid and accurate method of determining the fission gas pressure in a fuel rod without puncturing the cladding. This procedure is time-consuming and expensive and as a consequence a relatively small number of measurements are generally made compared with the number of fuel rods irradiated. In this paper it is proposed a new method for the measurement of pressure that is: Non-destructive; Non-invasive (i.e., allows re-irradiation of the measured rod); Easy to operate - directly in the reactor pool; Can be used on the critical path; Is inexpensive compared with the methods currently in use. This method is also being adapted to the on line measurement of fission gas release on fuel irradiation in research reactors. This method is based on the application of acoustic technology

  14. Development of a method for studying non-linear phenomena in plasma; Mise au point d'une methode d'etude des phenomenes non lineaires dans un plasma

    Energy Technology Data Exchange (ETDEWEB)

    Gonfalone, A [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

    1965-03-01

    gaz d'electrons est a la base des meilleures methodes de mesure des frequences de collisions electroniques dans l'ionosphere. Dans les plasmas de laboratoire l'apparition de ces phenomenes est liee a un champ critique de valeur elevee. Tout en n'utilisant qu'une source electromagnetique de puissance relativement faible, nous avons mis au point une methode hyperfrequence simple qui permet de mettre ces phenomenes en evidence, de les mesurer et d'en donner une interpretation elementaire. Une onde electromagnetique, fournie par un klystron de 1 W de puissance, interagit avec une decharge HF (25 MHz) dans le volume d'une cavite TE{sub 111}. Le tube contenant la decharge et l'axe de la cavite cylindrique sont colineaires a un champ magnetique pouvant atteindre une valeur telle que la frequence de l'onde soit egale a la frequence gyromagnetique des electrons. La courbe de resonance de la cavite, qui depend de la densite electronique et de la frequence des collisions, devient d'autant plus dissymetrique que la puissance absorbee est grande et que la frequence gyromagnetique des electrons est voisine de la frequence de l'onde incidente. L'etude de la resonance permet de calculer les coefficients de proportionnalite qui relient les variations de la densite et de la frequence de collisions a la puissance absorbee. Les experiences ont ete faites en faisant varier separement: la puissance UHF incidente, la densite electronique initiale, la pression du gaz neutre ambiant, ainsi que le champ magnetique axial. La variation de la densite electronique en fonction du champ magnetique pour une puissance UHF forte, montre une resonance de forme dissymetrique avec quelquefois un pic aigu, au voisinage de {omega}{sub H}. L'application eventuelle des proprietes mises en evidence a la realisation de dispositifs pratiques est envisagee. (auteur)

  15. A calculation and uncertainty evaluation method for the effective area of a piston rod used in quasi-static pressure calibration

    Science.gov (United States)

    Gu, Tingwei; Kong, Deren; Shang, Fei; Chen, Jing

    2018-04-01

    This paper describes the merits and demerits of different sensors for measuring propellant gas pressure, the applicable range of the frequently used dynamic pressure calibration methods, and the working principle of absolute quasi-static pressure calibration based on the drop-weight device. The main factors affecting the accuracy of pressure calibration are analyzed from two aspects of the force sensor and the piston area. To calculate the effective area of the piston rod and evaluate the uncertainty between the force sensor and the corresponding peak pressure in the absolute quasi-static pressure calibration process, a method for solving these problems based on the least squares principle is proposed. According to the relevant quasi-static pressure calibration experimental data, the least squares fitting model between the peak force and the peak pressure, and the effective area of the piston rod and its measurement uncertainty, are obtained. The fitting model is tested by an additional group of experiments, and the peak pressure obtained by the existing high-precision comparison calibration method is taken as the reference value. The test results show that the peak pressure obtained by the least squares fitting model is closer to the reference value than the one directly calculated by the cross-sectional area of the piston rod. When the peak pressure is higher than 150 MPa, the percentage difference is less than 0.71%, which can meet the requirements of practical application.

  16. Fuel rods

    International Nuclear Information System (INIS)

    Fukushima, Kimichika.

    1984-01-01

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

  17. 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.)

  18. 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.)

  19. Status of rod consolidation, 1988

    International Nuclear Information System (INIS)

    Bailey, W.J.

    1989-01-01

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

  20. Single-crystal SrTiO3 fiber grown by laser heated pedestal growth method: influence of ceramic feed rod preparation in fiber quality

    Directory of Open Access Journals (Sweden)

    D. Reyes Ardila

    1998-10-01

    Full Text Available The rapidly spreading use of optical fiber as a transmission medium has created an interest in fiber-compatible optical devices and methods for growing them, such as the Laser Heated Pedestal Growth (LHPG. This paper reports on the influence of the ceramic feed rod treatment on fiber quality and optimization of ceramic pedestal processing that allows improvements to be made on the final quality in a simple manner. Using the LHPG technique, transparent crack-free colorless single crystal fibers of SrTiO3 (0.50 mm in diameter and 30-40 mm in length were grown directly from green-body feed rods, without using external oxygen atmosphere.

  1. Atomization in graphite-furnace atomic absorption spectrometry. Peak-height method vs. integration method of measuring absorbance: carbon rod atomizer 63

    International Nuclear Information System (INIS)

    Sturgeon, R.E.; Chakrabarti, C.L.; Maines, I.S.; Bertels, P.C.

    1975-01-01

    Oscilloscopic traces of transient atomic absorption signals generated during continuous heating of a Carbon Rod Atomizer model 63 show features which are characteristic of the element being atomized. This research was undertaken to determine the significance and usefulness of the two analytically significant parameters, absorbance maximum and integrated absorbance. For measuring integrated absorbance, an electronic integrating control unit consisting of a timing circuit, a lock-in amplifier, and a digital voltmeter, which functions as a direct absorbance x second readout, has been designed, developed, and successfully tested. Oscilloscopic and recorder traces of the absorbance maximum and digital display of the integrated absorbance are simultaneously obtained. For the elements studied, Cd, Zn, Cu, Al, Sn, Mo, and V, the detection limits and the precision obtained are practically identical for both methods of measurements. The sensitivities by the integration method are about the same as, or less than, those obtained by the peak-height method, whereas the calibration curves by the former are generally linear over wider ranges of concentrations. (U.S.)

  2. RESEARCH OF INFLUENCE OF THE RODS CONSTRUCTION ON THEIR COOLING ABILITY AT FROSTING OF SILUMINS BY METHOD OF NUMERICAL MODELING

    Directory of Open Access Journals (Sweden)

    V. Yu. Stetsenko

    2012-01-01

    Full Text Available Numerical modeling of heat transfer coefficient on the surface of the water-cooled rod with a slotted and jet cooling was made.  calculations were carried out in a free, open  source  CFD software package OpenFOAM. it is shown that jet cooling is more uniform and intense compared to the slotted cooling. 

  3. 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.)

  4. Calculation of hydrogen and oxygen uptake in fuel rod cladding during severe accidents using the integral diffusion method -- Preliminary design report

    International Nuclear Information System (INIS)

    Siefken, L.J.

    1999-01-01

    Preliminary designs are described for models of hydrogen and oxygen uptake in fuel rod cladding during severe accidents. Calculation of the uptake involves the modeling of seven processes: (1) diffusion of oxygen from the bulk gas into the boundary layer at the external cladding surface, (2) diffusion from the boundary layer into the oxide layer, (3) diffusion from the inner surface of the oxide layer into the metallic part of the cladding, (4) uptake of hydrogen in the event that the cladding oxide layer is dissolved in a steam-starved region, (5) embrittlement of cladding due to hydrogen uptake, (6) cracking of cladding during quenching due to its embrittlement and (7) release of hydrogen from the cladding after cracking of the cladding. An integral diffusion method is described for calculating the diffusion processes in the cladding. Experimental results are presented that show a rapid uptake of hydrogen in the event of dissolution of the oxide layer and a rapid release of hydrogen in the event of cracking of the oxide layer. These experimental results are used as a basis for calculating the rate of hydrogen uptake and the rate of hydrogen release. The uptake of hydrogen is limited to the equilibrium solubility calculated by applying Sievert's law. The uptake of hydrogen is an exothermic reaction that accelerates the heatup of a fuel rod. An embrittlement criteria is described that accounts for hydrogen and oxygen concentration and the extent of oxidation. A design is described for implementing the models for hydrogen and oxygen uptake and cladding embrittlement into the programming framework of the SCDAP/RELAP5 code. A test matrix is described for assessing the impact of the proposed models on the calculated behavior of fuel rods in severe accident conditions. This report is a revision and reissue of the report entitled; ''Preliminary Design Report for Modeling of Hydrogen Uptake in Fuel Rod Cladding During Severe Accidents.''

  5. 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.)

  6. 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

  7. Monte-Carlo method for studying the slowing down of neutrons in a thin plate of hydrogenated matter; Methode de Monte-Carlo pour l'etude du ralentissement des neutrons dans une plaque mince de matiere hydrogenee

    Energy Technology Data Exchange (ETDEWEB)

    Ribon, P; Michaudon, A [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1965-07-01

    The studies of interaction of slow neutrons with atomic nuclei by means of the time of flight methods are made with a pulsed neutron source with a broad energy spectrum. The measurement accuracy needs a high intensity and an output time as short as possible and well defined. If the neutrons source is a target bombarded by the beam of a pulsed accelerator, it is usually required to slow down the neutrons to obtain a sufficient intensity at low energies. The purpose of the Monte-Carlo method which is described in this paper is to study the slowing down properties, mainly the intensity and the output time distribution of the slowed-down neutrons. The choice of the method and parameters studied is explained as well as the principles, some calculations and the program organization. A few results given as examples were obtained in the line of this program, the limits of which are principally due to simplifying physical hypotheses. (author) [French] l'etude de l'interaction des neutrons lents avec les noyaux atomiques par la methode du temps de vol s'effectue avec une source pulsee de neutrons dont le spectre en energie est assez etendu. La precision des mesures demande que la source soit intense et que la duree d'emission des neutrons soit breve et bien definie. Si la source est une cible bombardee par le faisceau de particules d'un accelerateur pulse, il est generalement indispensable de ralentir les neutrons pour avoir une intensite suffisante a basse energie. Nous presentons ici une methode de Monte-Carlo pour l'etude detaillee de ce ralentissement, notamment l'intensite et la distribution des temps de sortie des neutrons ralentis. Cette presentation comprend: la justification du choix de la methode de Monte-Carlo, les principes generaux, les differentes etapes du calcul et du programme ecrit pour le calculateur electronique IBM 7090. Nous indiquons aussi les restrictions qui sont apportees au domaine d'application de ce programme et qui proviennent surtout des

  8. ELSY neutronic analysis by deterministic and Monte Carlo methods. An innovative concept for the control rod systems

    International Nuclear Information System (INIS)

    Artioli, Carlo; Sarotto, Massimo; Grasso, Giacomo; Krepel, Jiri

    2009-01-01

    This paper deals with the neutronic design of ELSY (the European Lead-cooled SYstem), a 600 MW e Fast Reactor developed within the 6th EURATOM Framework Programme. ELSY aims at being an 'adiabatic' system (as far as possible) in order to fulfill both the requirements of sustainability and proliferation resistance. It represents the European solution for the Lead Fast Reactor (LFR), one of the six candidate typologies proposed by the Generation-IV International Forum (GIF). The analysis of the ELSY reference configuration, with typical pure MOX loading, is here presented. An introductory investigation of the adiabatic and, possibly, the burner options viability is also achieved by providing a rough estimate of the Minor Actinides (MAs) equilibrium concentrations and time constants. One of the main challenge-points in the design of the core, made up of wrapper-less square Fuel Assemblies (FAs) according to the common scheme of PWRs, is the small delta-T between the coolant average outlet temperature (480degC) and the allowable cladding one (550degC): it requires a rather flat radial power distribution, obtained by segmenting the core in three zones with different enrichments. Three different control sets have been introduced in order to achieve the required reliability for reactor shutdown and safety systems: eight traditional concept Control Rod (CR) assemblies together with two independent systems of sparse control 'Finger Absorber' Rods (FARs), small B 4 C rods that can be inserted, in principle, in the center of each FA. One of the two finger absorber systems includes a subset of rods devoted to the regulation of the criticality swing during the cycle: their number can be limited indeed since the small reactivity swing (some hundreds pcm) due to the about unitary breeding ratio. Such an innovative solution can also be positioned in order to maintain an optimal power flattening during the fuel cycle. To verify the feasibility of this solution, a very detailed

  9. Distribution of Al and in impurities along homogeneous Ge-Si crystals grown by the Czochralski method using Si feeding rod

    Science.gov (United States)

    Kyazimova, V. K.; Alekperov, A. I.; Zakhrabekova, Z. M.; Azhdarov, G. Kh.

    2014-05-01

    A distribution of Al and In impurities in Ge1 - x Si x crystals (0 ≤ x ≤ 0.3) grown by a modified Czochralski method (with continuous feeding of melt using a Si rod) have been studied experimentally and theoretically. Experimental Al and In concentrations along homogeneous crystals have been determined from Hall measurements. The problem of Al and In impurity distribution in homogeneous Ge-Si single crystals grown in the same way is solved within the Pfann approximation. A set of dependences of Al and In concentrations on the crystal length obtained within this approximation demonstrates a good correspondence between the experimental and theoretical data.

  10. Automated nuclear fuel rod pattern loading system

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  11. 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.)

  12. UNE-EN ISO/IEC 17025:2005-accredited method for the determination of pesticide residues in fruit and vegetable samples by LC-MS/MS.

    Science.gov (United States)

    Camino-Sánchez, F J; Zafra-Gómez, A; Oliver-Rodríguez, B; Ballesteros, O; Navalón, A; Crovetto, G; Vílchez, J L

    2010-11-01

    A rapid, simple and sensitive multi-residue method was developed and validated for the simultaneous quantification and confirmation of 69 pesticides in fruit and vegetables using liquid chromatography-tandem mass spectrometry (LC-MS/MS). The samples were extracted following the quick, easy, cheap, effective, rugged and safe method known as QuEChERS. Mass spectrometric conditions were individually optimised for each analyte in order to achieve maximum sensitivity in multiple reaction monitoring (MRM) mode. Using the developed chromatographic conditions, 69 pesticides can be separated in less than 17 min. Two selected reaction monitoring (SRM) assays were used for each pesticide to obtain simultaneous quantification and identification in one run. With this method in SRM mode, more than 150 pesticides can be analysed and quantified, but their confirmation is not possible in all cases according to the European regulations on pesticide residues. Nine common representative matrices (zucchini, melon, cucumber, watermelon, tomato, garlic, eggplant, lettuce and pepper) were selected to investigate the effect of different matrices on recovery and precision. Mean recoveries ranged from 70% to 120%, with relative standard deviations (RSDs) lower than 20% for all the pesticides. The proposed method was applied to the analysis of more than 2000 vegetable samples from the extensive greenhouse cultivation in the province of Almeria, Spain, during one year. The methodology combines the advantages of both QuEChERS and LC-MS/MS producing a very rapid, sensitive, accurate and reliable procedure that can be applied in routine analytical laboratories. The method was validated and accredited according to UNE-EN-ISO/IEC 17025:2005 international standard (accreditation number 278/LE1027).

  13. Determination of Optimal Imaging Mode for Ultrasonographic Detection of Subdermal Contraceptive Rods: Comparison of Spatial Compound, Conventional, and Tissue Harmonic Imaging Methods

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sung Jin; Seo, Kyung; Song, Ho Taek; Park, Ah Young; Kim, Yaena; Yoon, Choon Sik [Gangnam Severance Hospital, Yonsei University College of Medicine, Seoul (Korea, Republic of); Suh, Jin Suck; Kim, Ah Hyun [Dept. of Radiology and Research Institute of Radiological Science, Severance Hospital, Yonsei University College of Medicine, Seoul (Korea, Republic of); Ryu, Jeong Ah [Dept. of Radiology, Guri Hospital, Hanyang University College of Medicine, Guri (Korea, Republic of); Park, Jeong Seon [Dept. of Radiology, Hanyang University Hospital, Hanyang University College of Medicine, Seoul (Korea, Republic of)

    2012-09-15

    To determine which mode of ultrasonography (US), among the conventional, spatial compound, and tissue-harmonic methods, exhibits the best performance for the detection of Implanon with respect to generation of posterior acoustic shadowing (PAS). A total of 21 patients, referred for localization of impalpable Implanon, underwent US, using the three modes with default settings (i.e., wide focal zone). Representative transverse images of the rods, according to each mode for all patients, were obtained. The resulting 63 images were reviewed by four observers. The observers provided a confidence score for the presence of PAS, using a five-point scale ranging from 1 (definitely absent) to 5 (definitely present), with scores of 4 or 5 for PAS being considered as detection. The average scores of PAS, obtained from the three different modes for each observer, were compared using one-way repeated measure ANOVA. The detection rates were compared using a weighted least square method. Statistically, the tissue harmonic mode was significantly superior to the other two modes, when comparing the average scores of PAS for all observers (p < 0.00-1). The detection rate was also highest for the tissue harmonic mode (p < 0.001). Tissue harmonic mode in US appears to be the most suitable in detecting subdermal contraceptive implant rods.

  14. Determination of Optimal Imaging Mode for Ultrasonographic Detection of Subdermal Contraceptive Rods: Comparison of Spatial Compound, Conventional, and Tissue Harmonic Imaging Methods

    International Nuclear Information System (INIS)

    Kim, Sung Jin; Seo, Kyung; Song, Ho Taek; Park, Ah Young; Kim, Yaena; Yoon, Choon Sik; Suh, Jin Suck; Kim, Ah Hyun; Ryu, Jeong Ah; Park, Jeong Seon

    2012-01-01

    To determine which mode of ultrasonography (US), among the conventional, spatial compound, and tissue-harmonic methods, exhibits the best performance for the detection of Implanon with respect to generation of posterior acoustic shadowing (PAS). A total of 21 patients, referred for localization of impalpable Implanon, underwent US, using the three modes with default settings (i.e., wide focal zone). Representative transverse images of the rods, according to each mode for all patients, were obtained. The resulting 63 images were reviewed by four observers. The observers provided a confidence score for the presence of PAS, using a five-point scale ranging from 1 (definitely absent) to 5 (definitely present), with scores of 4 or 5 for PAS being considered as detection. The average scores of PAS, obtained from the three different modes for each observer, were compared using one-way repeated measure ANOVA. The detection rates were compared using a weighted least square method. Statistically, the tissue harmonic mode was significantly superior to the other two modes, when comparing the average scores of PAS for all observers (p < 0.00-1). The detection rate was also highest for the tissue harmonic mode (p < 0.001). Tissue harmonic mode in US appears to be the most suitable in detecting subdermal contraceptive implant rods.

  15. The Technological Consolidation of UNED in Spain

    Directory of Open Access Journals (Sweden)

    Lorenzo Garcia Aretio

    2001-07-01

    Full Text Available This article discusses the role of the technologies that have been utilized to advance distance teaching and learning by the National Distance Education University (Universidad Nacional de Educación a Distancia – UNED of Spain. Following a description of UNED's historical development and organizational structure, UNED's experience with various educational media is discussed. Printed teaching materials, in the form of didactic units, were one of the first methods to be utilized when UNED began its operations in 1972. In turn, the role of radio and audio recordings, television and video recordings, telephone, videoconferencing, computer systems and computer-mediated communications are also described. UNED's pioneering projects, including the virtual classroom, virtual campus, and a program for the physically handicapped, are also detailed. Recent experiments include providing access to radio and television programs on the Internet and adoption of WebCT. On the horizon for UNED are portals for cellular phones using WAP technology and gearing up for multiple applications in accordance with Universal Mobile Telecommunications Technology (UMTS.

  16. 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

  17. Nuclear fuel rod end plug weld inspection

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  18. Simulation of leaking fuel rods

    International Nuclear Information System (INIS)

    Hozer, Z.

    2006-01-01

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

  19. Fuel rod leak detector

    International Nuclear Information System (INIS)

    Womack, R.E.

    1978-01-01

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

  20. Developpement et implementation d'une methode pour resoudre les equations de la couche limite laminaire et turbulente

    Science.gov (United States)

    Leuca, Maxim

    CFD (Computational Fluid Dynamics) is a computational tool for studying flow in science and technology. The Aerospace Industry uses increasingly the CFD modeling and design phase of the aircraft, so the precision with which phenomena are simulated boundary layer is very important. The research efforts are focused on optimizing the aerodynamic performance of airfoils to predict the drag and delay the laminar-turbulent transition. CFD codes must be fast and efficient to model complex geometries for aerodynamic flows. The resolution of the boundary layer equations requires a large amount of computing resources for viscous flows. CFD codes are commonly used to simulate aerodynamic flows, require normal meshes to the wall, extremely fine, and, by consequence, the calculations are very expensive. . This thesis proposes a new approach to solve the equations of boundary layer for laminar and turbulent flows using an approach based on the finite difference method. Integrated into a code of panels, this concept allows to solve airfoils avoiding the use of iterative algorithms, usually computing time and often involving convergence problems. The main advantages of panels methods are their simplicity and ability to obtain, with minimal computational effort, solutions in complex flow conditions for relatively complicated configurations. To verify and validate the developed program, experimental data are used as references when available. Xfoil code is used to obtain data as a pseudo references. Pseudo-reference, as in the absence of experimental data, we cannot really compare two software together. Xfoil is a program that has proven to be accurate and inexpensive computing resources. Developed by Drela (1985), this program uses the method with two integral to design and analyze profiles of wings at low speed (Drela et Youngren, 2014), (Drela, 2003). NACA 0012, NACA 4412, and ATR-42 airfoils have been used for this study. For the airfoils NACA 0012 and NACA 4412 the calculations

  1. Use of a Monte-Carlo method for studying the statistical distribution of electric fields around an ion in a one-component plasma; Etude, par une methode de Monte-Carlo de la repartition statistique des champs electriques au niveau d'un ion, dans un plasma a une composante

    Energy Technology Data Exchange (ETDEWEB)

    Rossignol-Guzzi, D [Commissariat a l' Energie Atomique, 94 - Limeil-Brevannes (France). Centre d' Etudes Nucleaires

    1968-11-01

    A Monte-Carlo simulation has been made of the equilibrium configurations taken by a plasma of equally charged punctual ions, immersed in a uniform neutralizing background of electrons. The statistical repartition of the electric field acting on one ion, needed to obtain Stark effect, was specially obtained. Comparison for dense plasmas, was made with the former works of Holtzmark, Mayer, Broyles. (author) [French] On simule sur ordinateur, suivant une methode de Monte-Carlo, les configurations prises a l'equilibre thermodynamique par un plasma d'ions ponctuels et de meme charge, places dans un milieu d'electrons uniformement distribues. On etudie, en particulier, la repartition statistique des champs electriques au niveau d'un ion, utilisee dans les calculs d'effets Stark. On compare, dans le cadre des plasmas denses, les resultats obtenus aux travaux precedents de Holtzmark, Mayer, Broyles. (auteur)

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

    International Nuclear Information System (INIS)

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

    1981-11-01

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

  3. 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.)

  4. Safety rod driving device

    International Nuclear Information System (INIS)

    Murakami, Kiyonobu; Kurosaki, Akira.

    1988-01-01

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

  5. X-spectrographic method for plutonium detection. Application to contamination measurements in humans; Etude d'une methode de detection du plutonium par spectrographie X. Application a la mesure des contaminations sur l'homme

    Energy Technology Data Exchange (ETDEWEB)

    Trouble, Michel [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1967-07-01

    After reviewing the radio-toxicology of plutonium 239 and conventional detection methods using its {alpha}-radiation, the author considers the measurement of the X emission spectrum of plutonium 239 using a proportional counter filled with argon under pressure. This preliminary work leads to the third part of this research involving the detailed study of the possibilities of applying thin alkali halide crystal scintillators to the detection of soft plutonium X-rays; there follows a systematic study of all the parameters liable to render the detection as sensitive as possible: movement due to the photomultiplier itself and its accessory electronic equipment, nature and size of the crystal scintillator as well as its mode of preparation, shielding against external parasitic radiation. Examples of some applications to the measurement of contamination in humans give an idea of the sensitivity of this method. (author) [French] Apres un apercu de la radiotoxicologie du plutonium 239 et des methodes classiques de detection par son rayonnement {alpha}, on etudie le spectre d'emission X du plutonium 239 avec un compteur proportionnel rempli avec de l'argon sous pression. Ce travail preliminaire permet d'aborder la troisieme partie de cette etude dans laquelle nous examinons d'une fagon approfondie les possibilites d'application des cristaux scintillateurs minces d'halogenure alcalin a la detection du rayonnement X mou du plutonium; suit une etude systematique de tous les parametres susceptibles de rendre la detection aussi sensible que possible: mouvement propre du photomultiplicateur et de l'electronique associee, nature et dimensions du cristal scintillateur ainsi que son mode de fabrication, blindage contre les rayonnements parasites exterieurs. Quelques applications a la mesure des contaminations sur l'homme permettent d'apprecier la sensibilite de cette methode. (auteur)

  6. Measurement of a neutral particle flux by a thermal method using the junction temperature effect; Mesure d'un flux de particules neutres par une methode thermique mettant a contribution l'effet de temperature des jonctions

    Energy Technology Data Exchange (ETDEWEB)

    Caron, Anthime [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires. Services Scientifiques

    1966-07-01

    Among all the methods suitable for measuring neutral particle fluxes obtained by proton charge exchange in an organic gas, the thermal method has been chosen. The energy imparted by the neutral particles to the target in the form of heat leads to the latter temperature increasing; this temperature is usually followed with a thermocouple. In order to increase the sensitivity and the elegance of the apparatus the thermocouple has been replaced by a junction whose characteristics are known to vary with temperature. A calibration is carried out using a beam of charged particles. The response obtained is linear. Measurements have been made with a power of up to 1 mW; the accuracy increases with the energy provided; for 4 joules an accuracy of 10 per cent is obtained. The apparatus may be improved in particular by extending the measurement range towards low power values, and by increasing the accuracy. (author) [French] Parmi toutes les methodes utilisees pour la mesure d'un flux de particules neutres, obtenues par echange de charge de protons dans un gaz organique, nous avons choisi la methode thermique. L'energie cedee par les particules neutres a la cible sous forme de chaleur provoque une elevation de temperature de celle-ci; cette temperature est habituellement reperee par thermocouple. Pour accroitre la sensibilite et la finesse de l'appareillage, nous avons substitue au thermocouple une jonction dont on sait que les caracteristiques varient avec la temperature. Un etalonnage est realise par un faisceau de particules chargees. La reponse obtenue est lineaire. Des puissances de l'ordre du mW ont ete mesurees; la precision croit avec l'energie apportee; elle est de 10 pour cent quand celle-ci est de 4 joules. L'appareillage peut etre notablement perfectionne, pour reculer la gamme des mesures vers les basses puissances et accroitre la precision. (auteur)

  7. Modeling and simulation performance of sucker rod beam pump

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-30

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

  8. Modeling and simulation performance of sucker rod beam pump

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  9. Status of rod consolidation

    International Nuclear Information System (INIS)

    Bailey, W.J.

    1985-04-01

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

  10. Adjustment of a direct method for the determination of man body burden in Pu-239 on by X-ray detection of U-235; Mise au point d'une methode directe de determination de la charge corporelle en plutonium 239 chez l'homme par detection X de l'uranium 235

    Energy Technology Data Exchange (ETDEWEB)

    Boulay, P [Commissariat a l' Energie Atomique, Bruyeres-le-Chatel (France). Centre d' Etudes

    1968-04-01

    The use of Pu-239 on a larger scale sets a problem about the contamination measurement by aerosol at lung level. A method of direct measurement of Pu-239 lung burden is possible, thanks to the use of a large area window proportional counter. A counter of such pattern, has been especially carried out for this purpose. The adjustment of the apparatus allows an adequate sensibility to detect a contamination at the maximum permissible body burden level. Besides, a method for individual 'internal calibration', with a plutonium mock: the protactinium-233, is reported. (author) [French] L'utilisation a une echelle de plus en plus large du plutonium-239 pose un probleme de la mesure de la contamination par aerosol au niveau du poumon. Une methode de mesure directe de la charge pulmonaire en plutonium-239 est possible grace a l'utilisation d'un compteur proportionnel a fenetre de grande surface. Un compteur de ce type a specialement ete realise dans ce but. La mise au point de l'appareillage permet une sensibilite suffisante pour deceler une contamination au niveau de la Q.M.A (quantite maximale admissible). D'autre part, une methode 'd'etalonnage interne' de l'individu a l'aide d'un simulateur de plutonium, le protactinium-233, est decrite. (auteur)

  11. Low fluid level in pulse rod shock absorber

    International Nuclear Information System (INIS)

    Aderhold, H.C.

    1974-01-01

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

  12. Low fluid level in pulse rod shock absorber

    Energy Technology Data Exchange (ETDEWEB)

    Aderhold, H. C.

    1974-07-01

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

  13. 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.)

  14. 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.)

  15. Why Rods and Cocci

    Indian Academy of Sciences (India)

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

  16. Control rod drives

    International Nuclear Information System (INIS)

    Oonuki, Koji.

    1981-01-01

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

  17. Control of the neutronic and thermohydraulic conditions of power ramps in an irradiation loop for PWR fuel rod; Controle des conditions neutroniques et thermohydrauliques des rampes de puissance dans une boucle d`irradiation de combustibles de reacteur a eau pressurisee

    Energy Technology Data Exchange (ETDEWEB)

    Moulin, D J.F.

    1993-09-10

    In order to study the power transients effects on PWR fuel rod clad, ramp tests in a pressurized water loop, are carried out at OSIRIS reactor. The present thesis deals with the on-line control of the device, during power ramp and conditioning irradiation. Based on a convolution-type resolution of the kinetics equations, a dynamic compensation of the Silver self-powered neutron detector was developed. With this method, the uncertainty of the ramp end-point is lower than 1%, thus it is very suited for monitoring both transient, as well as steady state conditions. Furthermore, a thermohydraulic model of the irradiation device is described: heat transfer equations, including gamma heating in materials, are solved to obtain temperatures and thermal fluxes of steady states. Results from the model and temperature measurements of the coolant are used together for fuel power determination, in real time. The clad external temperature profile is also calculated and displayed, to improve the irradiation monitoring. (author), 51 refs., 12 annexes, 66 figs.

  18. Failed fuel rod detector

    Energy Technology Data Exchange (ETDEWEB)

    Uchida, Katsuya; Matsuda, Yasuhiko

    1984-05-02

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

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

  20. Fabrication Methods of Fullerenes. a Critical Review Méthodes de fabrication des fullerènes. Une étude critique

    Directory of Open Access Journals (Sweden)

    Emberson S. C.

    2006-11-01

    Full Text Available The industrial use of fullerenes will require a significant up-scaling of their production. The five actually known techniques are discussed : 1. The electric arc (Krätschmer-Huffman. 2. The vaporization of carbon by pulsed lasers. 3. The direct vaporization of carbon in focused sunlight. 4. The direct inductive heating of carbon. 5. Sooting hydrocarbon flames. The difficulty of up-scaling the electric arc set-up because of the photochemical destruction of fullerenes will be demonstrated. The use of sooting hydrocarbon flames for an industrial production of fullerenes is suggested. L'utilisation industrielle de fullerènes nécessitera une augmentation significative de leur production. Les cinq techniques actuellement connues sont discutées : 1. L'arc électrique (Krätschmer-Huffman. 2. L'évaporation de carbone par laser pulsé. 3. L'évaporation de carbone par lumière solaire focalisée. 4. L'évaporation thermique de carbone. 5. Une flamme d'hydrocarbures formant des suies. On met en évidence la difficulté d'une montée en échelle des installations à base d'arcs électriques à cause de la destruction photochimique des fullerènes. L'utilisation de flammes produisant des suies est proposée pour la production industrielle de fullerènes.

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

    CSIR Research Space (South Africa)

    Shatalov, M

    2011-07-01

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

  2. Space synthesis: an application of synthesis method to two and three dimensional multigroup neutron diffusion equations; Synthese spatiale: une application de la methode de synthese aux equations de diffusion neutronique multigroupe a deux et trois dimensions

    Energy Technology Data Exchange (ETDEWEB)

    Nguyen-Ngoc, H [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1969-07-01

    In order to reduce computing time, two and three-dimensional multigroup neutron diffusion equations in cylindrical, rectangular (X, Y), (X, Y, Z) and hexagonal geometries are solved by the method of synthesis using an appropriate variational principle (stationary principle). The basic idea is to reduce the number of independent variables by constructing two or three-dimensional solutions from solutions of fewer variables, hence the name 'synthesis method'. Whatever the geometry, we are led to solve a system of ordinary differential equations with matrix coefficients to which one can apply well-known numerical methods: CHEBYSHEV's polynomial method, Gaussian elimination. Numerical results furnished by synthesis programs written for the IBM 7094, the IBM 360-75 and the CDC 6600 computers, are confronted with those which are given by programs employing the classical finite difference method. [French] En vue de reduire le-temps de calcul, les equations de diffusion neutronique, multigroupe, a deux et trois dimensions d'espace dans les geometries cylindrique, rectangulaire (X, Y), (X, Y, Z) et hexagonale sont resolues par la methode de synthese utilisant un principe variationnel approprie (principe stationnaire). L'idee consiste a reduire le nombre de variables independantes par construction d'une solution bi ou tridimensionnelle au moyen de solutions dependant d'un nombre inferieur de variables, d'ou le nom de la methode. Dans tous les cas de geometrie, nous sommes conduits a resoudre un systeme d'equations differentielles a coefficients matriciels auquel peuvent s'appliquer les methodes numeriques courantes; methode polynomiale de TCHEBYCHEFF et methode d'elimination de GAUSS. Les resultats numeriques obtenus par nos codes de synthese programmes sur IBM 7094, IBM 360-75 et CDC 6600, sont confrontes avec ceux que fournissent les programmes adoptant la methode classique des differences finies. (auteur)

  3. 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.)

  4. Multiple fuel rod gripper

    International Nuclear Information System (INIS)

    Shields, E.P.

    1987-01-01

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

  5. Fuel rod technology

    International Nuclear Information System (INIS)

    Bezold, H.; Romeiser, H.J.

    1979-07-01

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

  6. Control rod testing apparatus

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  7. Method of uranium prospecting in a mining division: development and results; La methode de prospection de l'uranium dans une division miniere: sa mise au point - ses resultats

    Energy Technology Data Exchange (ETDEWEB)

    Carrat, G. [Commissariat a l' energie atomique et aux energies alternatives - Service des Recherches a la Division Grury CEA (France)

    1959-07-01

    The main object of this report is to present the development of the prospecting method in a given region, the Morvan, carried out by the Grury Mining Division of the C.E.A.; with regard to the uraniferous mineral distribution of which the existence only came to light progressively as the work advanced. After a description of the various techniques which follow on one from the other finishing up at mine workings and the specification of a workable tonnage of uranium, an overall aspect of the Job accomplished in the last twelve years is presented. The prospecting method has been profoundly modified since the beginning of the work. Over the years it has evolved as a function of the knowledge progressively acquired, of the way the indications and the uraniferous deposits lie. In addition it has been varied by adapting to the ground in question the remarkable new technique known as radiometry or the study of surface radioactivity. lt has also made use of certain geophysical or geochemical techniques, thus producing a range of field tests which enable an advanced reconnaissance of the under soil to be made before mining is begun. However al no time has it excluded the classical and fundamental concept of geological ground sampling using the hammer and the compass. In this field an attempt has been made to use information provided by a precise geomorphological and tectonic test. Most of this work was carried out on the granitic ground of the Morvan, and the deposits considered in this study are all typically hydrothermal. Reprint of a paper published in 'Annales des Mines', March 1959 [French] Le but principal de cet expose est de presenter la mise au point de la methode de prospection d'une region determinee, le Morvan, suivie par la Division miniere de Grury du Commissariat a l' energie atomique, en fonction de la repartition de la mineralisation uranifere dont la realite n'est apparue que tres progressivement au fur et a mesure de l'avancement des travaux. Apres l

  8. Burnable poison rod

    International Nuclear Information System (INIS)

    Natsume, Tomohiro.

    1988-01-01

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

  9. 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.)

  10. Diffraction of an Electromagnetic Wave on a Dielectric Rod in a Rectangular Waveguide. A Method of Partial Waveguide Filling

    Science.gov (United States)

    Zav'yalov, A. S.

    2018-04-01

    A variant of the method of partial waveguide filling is considered in which a sample is put into a waveguide through holes in wide waveguide walls at the distance equal to a quarter of the wavelength in the waveguide from a short-circuiter, and the total input impedance of the sample in the waveguide is directly measured. The equivalent circuit of the sample is found both without and with account of the hole. It is demonstrated that consideration of the edge effect makes it possible to obtain more exact values of the dielectric permittivity.

  11. Development of interface tracking method. Two-phase flows applications; Developpement d'une methode de suivi d'interface. Applications aux ecoulements diphasiques

    Energy Technology Data Exchange (ETDEWEB)

    Tanguy, S.

    2004-11-15

    Spray formation mechanisms study from a liquid-gas flow is a fundamental research subject, which industrial applications are large, especially in combustion and propulsion field. Numerical simulation of such flows appear as an essential complement to experimental and theoretical studies, for comprehension and accurate prediction of such physical processes. In this study we developed an numerical interface tracking technique with a Navier-Stokes solver to study accurately the liquid-gas interface dynamics. We describe Level Set method which has been used to track interface motion, and numerical methods for solving Navier-Stokes equations. Different numerical schemes have been tested to improve the computation accuracy. Ghost Fluid Method enables a robust and accurate treatment of discontinuities across the liquid-gas interface. The codes developed (2D, 3D, parallelization MPI) are then used to study droplets collisions. Comparisons with experimental results show that simulations are realistic and predictive. Next, feasibility studies are done on more complex configurations. Droplets spray formation from primary atomization of a liquid jet seems to be especially a promising investigation field for such simulations. Finally, reactive interfaces propagation, as liquid vaporization and premixed combustion have also been studied using Ghost Fluid Method to impose specific jump conditions. (author)

  12. Une méthode précise pour la mise en évidence et l'étude de l'anisotropie dans les roches An Accurate Method for Detecting and Analyzing Anisotropy in Rocks

    Directory of Open Access Journals (Sweden)

    Talebi S.

    2006-11-01

    , nous avons tracé les isovitesses par interpolation linéaire entre les points de mesure. Trois plans de symétrie ont été observés dans la plupart des cas. Cette méthode pratique et originale de mesure sur les échantillons de cubes tronqués permet, outre une économie sur les blocs de roche, une mise en évidence nette et précise et une étude plus approfondie de l'anisotropie des roches, ainsi que de leurs éléments de symétrie éventuels. The main goal of this research was to develop an experimental method for analyzing the anisotropy of the velocity of longitudinal (P waves in rocks. This anistropy may have two main causes: (i the existence of a network of pores, cracks and microcracks distributed in an anisotropic way in the rock, and (ii the structure itself, whose anisotropy results from the nature and arrangement of the minerais. In a first phase, we began with tests on core samples drilled in three perpendicular directions, X, Y and Z, in a block of two types of sandstone: Vosges sandstone and Fontainebleau sandstone. The velocity of (P waves was measured in three to four core-sample sections with 30° intervals (6 directions. These velocities were clearly distributed along an ellipse, called the anisotropy ellipse. These ellipses were plotted by minicomputer, thus revealing the anisotropy state and homogeneity of the core samples. The method is so accurate that an axis rotation error in a sample (Y of Vosges sandstone was found and verified by the direction of the micas. We performed the same test on the same vacuum-saturated samples. A general tendency toward an increase in velocities, a decrease in the degree of anisotropy (the Vmax/Vmin ratio and a change in the position of the anisotropy axis was observed. A comparison of dry and saturated results gave a gond picture of the nature of the voids and their rote in overall anisotropy. In the next phase, we developed a measurement method using truncated-face cubes (maximum of 66 faces of different

  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 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.)

  15. Determination of detection limits for a VPD ICPMS method of analysis; Determination des limites de detection d'une methode d'analyse VPD ICPMS

    Energy Technology Data Exchange (ETDEWEB)

    Badard, M.; Veillerot, M

    2007-07-01

    This training course report presents the different methods of detection and quantifying of metallic impurities in semiconductors. One of the most precise technique is the collection of metal impurities by vapor phase decomposition (VPD) followed by their analysis by ICPMS (inductively coupled plasma mass spectrometry). The study shows the importance of detection limits in the domain of chemical analysis and the way to determine them for the ICPMS analysis. The results found on detection limits are excellent. Even if the detection limits reached with ICPMS performed after manual or automatic VPD are much higher than detection limits of ICPMS alone, this method remains one of the most sensible for ultra-traces analysis. (J.S.)

  16. 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

  17. Some dynamic aspects of the thyroid function studied by a method of isotopic balance; Quelques aspects dynamiques de la fonction thyroidienne etudies par une methode d''equilibre isotopique'

    Energy Technology Data Exchange (ETDEWEB)

    Simon, Claude

    1958-06-15

    This report addresses the use of a method which aims at directly measuring quantities of steady iodine involved by the thyroid metabolism in specific physiological conditions. The experiment is based on the use of male rats weighting 180 to 200 g which received an almost iodine-free diet. The daily dose of iodine is administered to each experimental group through beverage under the form of potassium iodide. After a delay, the usual beverage is replaced by a marked one. Iodine renewal by thyroid is monitored in vivo by external detection of radiations emitted by iodine 131. At the end of the experiment, rats are killed, and their blood is collected on heparin, centrifuged and processed to separate mineral iodine and hormonal iodine. Radioactivity is then measured with a scintillator and compared with that of a beverage sample [French] Le but de la methode utilisee est de mesurer directement les quantites d'iode stable mises en jeu par le metabolisme thyroidien dans des conditions physiologiques precises. Les travaux anterieurs ne permettent d'atteindre ces quantites que par des microdosages chimiques longs et delicats. En nous basant sur le principe du renouvellement d'un pool en etat dynamique stationnaire, nous decrivons une cinetique et mesurons en valeur absolue les differents parametres de ce systeme en equilibre.

  18. 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

  19. Study and comparison of some methods for calculating the transmission factor of a potential barrier in quantum mechanics (1963); Etude et comparaison de quelques methodes de calcul du facteur de transmission d'une barriere de potentiel en mecanique quantique (1963)

    Energy Technology Data Exchange (ETDEWEB)

    Jamet, P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    The author formulates two accurate methods for the calculation of the transmission coefficient of a one-dimensional potential barrier. The principles of these methods are. expressed in a symmetrical form with respect to the two sides of the potential barrier; this constitutes a proof of the fact that the transmission coefficient is path direction independent. The numerical application is carried out on several examples and the results are compared to those provided by the WKB method. (author) [French] L'auteur formule deux methodes exactes pour calculer le coefficient de transmission d'une barriere de potentiel unidimensionnelle. Les principes de ces methodes s'enoncent sous forme symetrique par rapport aux deux cotes de la barriere de potentiel, ce qui constitue une demonstration du fait que le coefficient de transmission est independant du sens de parcours. L'application numerique est faite sur quelques exemples et les resultats sont compares a ceux fournis par la methode B KW. (auteur)

  20. 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

  1. Developpement d'une methode d'analyse du cycle de vie consequentielle prospective macroscopique : Evaluation d'une politique de bioenergie dans L'Union Europeenne a l'horizon 2025

    Science.gov (United States)

    Dandres, Thomas

    Most of the time, the increase in human activities harms the environment. Due to the capacity limit of the Earth to bear such impacts and considering the growth of world population and its demands, there is a need to manage the future growth of human society in order to mitigate its impacts on the environment. Life cycle assessment (LCA) seems to be a great methodology for this purpose because it assesses different types of environmental impacts of a product or service based on all of its life cycle stages. However, it appears that LCA is not adapted to the evaluation of large-scale international policies required for some environmental issues. Especially, current LCA methodology fails to properly model indirect consequences on the environment of a major change within the energy sector in order to reduce greenhouse gas emissions and mitigate climate change. To model these indirect consequences on the environment, global economy modeling is required as a major change in the energy sector is expected to affect the rest of the economy and therefore to cause indirect environmental impacts. This is the goal of this university thesis: Develop a new decision tool "macroscopic life cycle assessment" using the LCA methodology and the GTAP macroeconomic model in order to assess environmental impacts caused by significant changes in human society. Application of macroscopic LCA to two different European energy policies for the 2005-2025 period shows its ability to model some rebound effects and to compare environmental benefits obtained from a bioenergy policy versus environmental impacts caused by the economic growth during 2005-2025. Additionally, the method allows to identify regions where and time periods when environmental impacts are expected to occur. Macroscopic LCA has important application opportunities for LCA: while the methodology is usually used to study the environmental profile of a product or service, it now becomes possible to evaluate environmental

  2. 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.)

  3. 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

  4. 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.)

  5. Sucker rod motor

    Energy Technology Data Exchange (ETDEWEB)

    Radzalov, N N; Radzhabov, N A

    1983-01-01

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

  6. Burnable poison rod

    International Nuclear Information System (INIS)

    Natsume, Tomohiro.

    1988-01-01

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

  7. Methods of obtaining an inert atmosphere for plutonium metal treatment installations; Modes d'obtention d'une atmosphere inerte dans les installations d'elaboration du plutonium metallique

    Energy Technology Data Exchange (ETDEWEB)

    Riolfo, R; Barbier, M

    1962-07-01

    Plutonium is a very pyrophoric metal (heat of combustion: 253 kCal/mole). The operations and manipulations involved in its treatment have thus to be carried out in an inert atmosphere. Several methods designed to eliminate the oxygen from the manipulation chamber have been tried: absorption by titanium - zirconium or copper turnings, or bubbling through potassium pyrogallate. They are not satisfactory. The reaction C + O{sub 2} -> CO{sub 2} has been chosen. Graphite is used. By this method, which is flexible and which has a negligible cost price, it is possible to absorb O{sub 2} within a wide range of concentrations (from 0.1 to 20 per cent) at a temperature 600 - 800 deg C compatible with the use of a conventional material. This report describes the trials carried out, the method selected and experimented, with the experimental details, the results obtained, and the extension of the method for a slightly different use. (authors) [French] Le plutonium est un metal tres pyrophorique (chaleur de combustion 253 kCal/mole). Les operations et manipulations qui decoulent de son elaboration doivent donc s'effectuer sous atmosphere inerte. Plusieurs methodes, dont le but etait d'eliminer l'oxygene de l'enceinte d'elaboration ont ete essayees: absorption par des copeaux de titane-zirconium, de cuivre, ou barbotage sur le pyrogallate de potasse. Elles n'ont pas donne satisfaction. La reaction C + O{sub 2} {yields} CO{sub 2} a ete retenue. On utilise le graphite. D'une tres grande souplesse d'utilisation, d'un prix de revient nul, il permet d'absorber l'oxygene dans une fourchette etendue de concentration (0,1 a 20 pour cent) a une temperature (600 a 800 degres C) compatible avec l'emploi d'un materiel classique. Cet expose decrit les essais effectues, la methode retenue et experimentee, le mode operatoire, les resultats obtenus, et l'extension de la methode a un probleme legerement different. (auteurs)

  8. An overview on rod-bundle thermal-hydraulic analyses

    International Nuclear Information System (INIS)

    Sha, W.T.

    1980-01-01

    Three methods used in rod-bundle thermal-hydraulic analysis are summarized. These methods are: (1) subchannel analysis, (2) porous medium formulation with volume porosity, surface permeability, distributed resistance and distributed heat source (sink) and, (3) bench-mark rod-bundle thermal-hydraulic analysis using a boundary-fitted coordinate system. Basic limitations and merits of each method are delineated. (orig.)

  9. Flow in rod bundles

    International Nuclear Information System (INIS)

    Hazi, G.; Mayer, G.

    2005-01-01

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

  10. Une perspective interactionniste

    Directory of Open Access Journals (Sweden)

    Joëlle Morrissette

    2010-02-01

    Full Text Available Cet article vise à montrer l’intérêt de puiser à la sociologie pour conduire des recherches dans le domaine de l’éducation. Plus précisément, il sera question de la contribution d’une perspective interactionniste pour appréhender un objet attaché à l’évaluation des apprentissages des élèves. Comme on le verra, s’inspirer en particulier d’auteurs attachés à la tradition de l’interactionnisme symbolique amène à se situer en marge des manières de dire et de faire habituelles des investigations portant sur cet objet, de la phase de problématisation à celle de l’analyse, au profit d’un point de vue (resocialisant et contextualisant. Pour illustrer le propos, je prendrai appui sur le format d’une recherche ayant documenté le savoir-faire d’un groupe d’enseignantes du primaire en matière d’évaluation formative, et ayant adopté une perspective interactionniste comme posture générale de recherche.An Interactionist PerspectiveAn Alternative Approach to Learning AssessmentThis article aims to show the usefulness of drawing from sociology to conduct research in the field of education. Specifically, it discusses the contribution of an interactionist perspective in understanding the objects attached to student learning assessment. As we shall see, drawing especially from authors working in the tradition of Symbolic Interactionism leads us outside the usual ways of thinking and doing in investigations related to assessment objects, from problematization to analysis, in favour of a (resocializing and contextualizing perspective. To illustrate this point, I will examine the format of a study documenting the expertise of a group of elementary school teachers with regard to formative assessment, and having an interactionist perspective as its basis of research.Una perspectiva interaccionista: otro punto de vista sobre la evaluación del aprendizajeEste artículo tiene como objetivo el demostrar el interés de

  11. 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.)

  12. UNE APPROCHE REGULATIONNISTE DE LA ...

    African Journals Online (AJOL)

    Administrateur

    en ressources naturelles ont enregistré des performances économiques, moins bonnes par .... de gestion», à financer des augmentations de salaires; augmentations qui, très ... entreprises), dans la mesure où, d'une façon générale, elle introduit une plus .... Outre le contrôle strict de la création monétaire, un autre élément.

  13. 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.)

  14. 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.)

  15. Trunnion Rod Microcrack Detection

    Science.gov (United States)

    2013-08-01

    Richard W. Haskins, Joseph A. Padula , and John E. Hite BACKGROUND: Post-tensioned rods are used to anchor spillway gates and transfer the forces...email: James.A.Evans@usace.army.mil). This technical note should be cited as follows: Evans, J. A., Haskins, R. W., Padula , J. A., and Hite, J. E. 2013

  16. On the Wave Stresses in the Rods of Anvil Hammers

    Directory of Open Access Journals (Sweden)

    V. M. Sinitskiy

    2014-01-01

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

  17. Contribution to the study of influences in emission spectrography on solutions. Application to a general analysis method for stainless steels (1961); Contribution a l'etude des influences en spectographie d'emission sur solution. Application a une methode generale d'analyse des aciers inoxydables (1961)

    Energy Technology Data Exchange (ETDEWEB)

    Baudin, G [Commissariat a l' Energie Atomique, Grenoble (France). Centre d' Etudes Nucleaires

    1961-11-15

    In order to establish a general method of analysis of stainless steels, by means of spark spectroscopy on solutions, a systematic study has been made of the factors involved. The variations in acidity of the solutions, or in the ratio of concentrations of two acids at constant pH, lead to a displacement of the calibration curve. Simple relations have been established between the concentration of the extraneous elements, and the effects produced, for the constituents Fe, Ti, Ni, Cr, Mn; a general method using abacus is proposed for steels containing only these elements. The interactions in the case of the elements Mo, Nb, Ta, W, were more complex, so that the simultaneous separation was studied with the help of ion-exchange resins. A general method of analysis is proposed for stainless steels. (author) [French] En vue d'etablir une methode generale d'analyse des aciers inoxydables par spectrographie d'etincelles sur solution, on a effectue une etude systematique des influences. Les variations de l'acidite des solutions ou du rapport des concentrations de deux acides a pH constant, entrainent un deplacement des courbes d'etalonnage. On a etabli des relations simples entre la teneur des tiers elements et les effets produits pour les constituants Fe, Ti, Ni, Cr, Mn; une methode generale avec abaques est proposee pour les aciers contenant ces seuls elements. Les influences dans le cas des elements Mo, Nb, Ta, W etant plus complexes, on eut a etudier la separation simultanee a l'aide de resines echangeuses d'ions. On propose une methode generale d'analyse des aciers inoxydables. (auteur)

  18. Morphoelastic rods. Part I: A single growing elastic rod

    KAUST Repository

    Moulton, D.E.

    2013-02-01

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

  19. 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

  20. Morphoelastic rods. Part I: A single growing elastic rod

    KAUST Repository

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

    2013-01-01

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

  1. 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)

  2. 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

  3. The study on the methods for improving the gredibility of NDT equipment for the gap of pellets of nuclear fuel rods

    International Nuclear Information System (INIS)

    Zhang Lei; Liu Ming; Wang Changhong; Ma Jinbo

    2014-01-01

    In order to improve the credibility of the new generation of automatic online non-destructive testing equipment for the gap of the pellets of nuclear fuel rods the researchers have done a lot of work in the development of the device. Such measures as multi-thread synchronization, precise timing, upper and lower computer communication control, antijamming processing are adopted such that the detecting device can accurately detecte the size of the gap between pellets, the position and length of the spring cavity at the front end of the nuclear fuel rods at a detection rate of 8 m/min. The detection credibility for the 0.5 mm gap is over 95%, reaching the international advanced level. At present, the device is put into use in the nuclear fuel element production line. (authors)

  4. 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

  5. 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)

  6. 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

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

    International Nuclear Information System (INIS)

    Lavicka, David; Polansky, Jiri

    2010-01-01

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

  8. {gamma} activity and heating of rods in EL2 and EL3; Activitiy {gamma} et echauffement des barres de EL2 et EL3

    Energy Technology Data Exchange (ETDEWEB)

    Lalere, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1959-07-01

    A method is described for calculating the {gamma} activity of uranium rods, given the mean flux in which they are irradiated, the time they remain in the pile and the duration of deactivation. This calculation leads to numerical formulae which may be applied to the rods of the two reactors. It allows the saturation activities to be foreseen both for EL2 and for EL3, taking into recount the minimum times necessary for extraction. Measurements have been carried out, and the results are in good agreement with those foreseen by calculation. In the last section this method is used to calculate the heating of the irradiated rods. (author) [French] Une methode est indiquee ici, qui permet de calculer l'activite {gamma} des barres d'uranium connaissant le flux moyen dans lequel elles ont ete irradiees, leur temps de sejour en pile et la duree de la desactivation. Ce calcul conduit a des formules numeriques que l'on peut appliquer aux barres des deux reacteurs. Il permet de prevoir les activites atteintes a saturation, tant a EL2 qu'a EL3, compte tenu des temps minima necessaires a l'extraction. Des mesures ont ete faites: les resultats sont en bon accord avec les previsions du calcul. Enfin, en derniere partie, cette methode est utilisee pour calculer l'echauffement des barres irradiees. (auteur)

  9. Fuel rod fixing system

    International Nuclear Information System (INIS)

    Christiansen, D.W.

    1982-01-01

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

  10. 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.)

  11. Fuel rod attachment system

    International Nuclear Information System (INIS)

    Christiansen, D.W.

    1982-01-01

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

  12. 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.)

  13. Nuclear fuel rod loading apparatus

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  14. 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.)

  15. Rod drive and latching mechanism

    International Nuclear Information System (INIS)

    Veronesi, L.; Sherwood, D.G.

    1982-01-01

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

  16. 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

  17. 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)

  18. 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.)

  19. 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

  20. Comparison of methods for measuring the ion exchange capacity of a soil. Development of a quick method; Comparaison des methodes de mesure de la capacite d'echange d'ions d'un sol. Mise au point d'une methode rapide

    Energy Technology Data Exchange (ETDEWEB)

    Amavis, R [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

    1959-07-01

    In the course of a study on the movement of radioactive ions in soil we had to measure the cationic exchange capacity of various soil samples, this parameter being one of the most important in the appreciation of the extent of fixation of radioactive ions in the ground. The object of this report is to describe the various methods used and to compare the results obtained. A colorimetric method, using Co(NH{sub 3}){sub 6}{sup 3+} as exchangeable ion, was developed. It gives results comparable to those obtained with conventional methods, whilst considerably reducing the time necessary for the operations. (author) [French] A l'occasion de l'etude du mouvement des ions radio-actifs dans un sol, nous avons ete amenes a mesurer la capacite d'echange cationique de differents echantillons de sols; ce parametre etant un des plus importants pour apprecier la valeur de la fixation des ions radioactifs dans un terrain. L'objet de ce rapport est d'exposer les diverses methodes utilisees et de comparer les resultats obtenus. Une methode calorimetrique, utilisant Co(NH{sub 3}){sub 6}{sup 3+} comme ion echangeable, a ete mise au point: elle donne des resultats comparables a ceux obtenus avec les methodes habituelles et permet de reduire considerablement la duree des manipulations. (auteur)

  1. 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

  2. Methods of Particle Detection in Free Neutron Decay; Methode de detection des particules dans une desintegration de neutrons libres; Metod obnaruzheniya chastits pri raspade svobodnogo nejtrona; Metodo para la deteccion de particulas en la desintegracion de neutrones libres

    Energy Technology Data Exchange (ETDEWEB)

    Novey, T B [Argonne National Laboratory, Lemont, IL (United States)

    1960-06-15

    A number of experimental studies have recently been completed by the Argonne Group on the decay of polarized neutrons in order to elucidate the structure of the weak nuclear interaction. These studies have taken the form of the measurement of the. angular distributions of electrons and protons with respect to the spin direction of the decaying free neutrons. The basic components of the apparatus which will be discussed are: 1. The one-meter iron-cobalt mirror which selects a beam of highly polarized neutrons and the methods for determination of the polarization. 2. The electron detector comprising a 10 cm diameter, 6 mm thick mosaic of anthracene crystals, and its light pipe system. 3. The proton detector, a 14-stage electron multiplier system, the first stage with a 15x15 cm opening, tapering in four stages to a standard 10-stage multiplier structure, and its entrance baffles for angular resolution. 4. The electronic system which selects pulses from the detectors having the proper time sequence, relative time-delay and pulse-height to allow indentification of a neutron decay. (author) [French] Le groupe Argonne a termine recemment des etudes experimentales sur la desintegration des neutrons polarises, en vue de mettre en lumiere le processus de l'interaction nucleaire faible. Ces etudes consistaient a mesurer les distributions angulaires d'electrons et de protons par rapport a la direction du spin des neutrons libres en voie de desintegration. Les principaux elements de l'appareillage qui sera decrit sont: 1. Le miroir de ferro-cobalt d'un metre, qui isole un faisceau de neutrons hautement polarises, ainsi que les methodes permettant de determiner la polarisation. 2. Le detecteur electronique qui comprend une mosaique de 10 cm de diametre et de 6 mm d'epaisseur en cristaux d'anthracene, et son systeme selectif. 3. Le detecteur de protons, un systeme multiplicateur electronique a 14 etages, le premier ayant une ouverture de 15x15 cm se retrecissant en 4 etages pour

  3. Cone rod dystrophies

    Science.gov (United States)

    Hamel, Christian P

    2007-01-01

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

  4. Cone rod dystrophies

    Directory of Open Access Journals (Sweden)

    Hamel Christian P

    2007-02-01

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

  5. Nuclear reactor fuel rod

    International Nuclear Information System (INIS)

    Busch, H.; Mindnich, F.R.

    1973-01-01

    The fuel rod consists of a can with at least one end cap and a plenum spring between this cap and the fuel. To prevent the hazard that a eutectic mixture is formed during welding of the end cap, a thermal insulation is added between the end cap and plenum spring. It consists of a comical extension of the end cap with a terminal disc against which the spring is supported. The end cap, the extension, and the disc may be formed by one or several pieces. If the disc is separated from the other parts it may be manufactured from chrome steel or VA steel. (DG) [de

  6. Temperature measurement in cans of fuel rods and fuel rod simulators

    International Nuclear Information System (INIS)

    Tschoeke, H.; Moeller, R.

    1977-01-01

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

  7. Une proposition de nouveau cartogramme

    Directory of Open Access Journals (Sweden)

    Raymond Badel

    1990-05-01

    Full Text Available Le cartogramme proposé est une technique de recherche-exploration rapide grâce aux possibilités offertes par le micro-ordinateur et sa diffusion. La propriété foncière en vallée d'Aure lui sert de support.

  8. Fuel rod pellet loading head

    International Nuclear Information System (INIS)

    Howell, T.E.

    1975-01-01

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

  9. 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

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

    International Nuclear Information System (INIS)

    Bogdan, G.

    1996-12-01

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

  11. 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.)

  12. 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.)

  13. 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)

  14. Use of the Neutron Die-Away Technique to Test Control Rod Effectiveness Theories; Emploi de la Methode d'Absorption des Neutrons pour Verifier les Theories sur l'Efficacite des Barres de Commande; Ispol'zovanie metoda spada potoka nejtronov dlya proverki teorij ehffektivnosti reguliruyushchikh sterzhnej; Aplicacion de la Tecnica de Extincion Neutronica a la Verificacion de las Teorias sobre la Eficacia de las Barras de Control

    Energy Technology Data Exchange (ETDEWEB)

    Perez, R. B. [University of Florida, Gainesville, FL (United States); De Saussure, G.; Silver, E. G. [University of Florida, Gainesville, FL (United States); Oak Ridge National Laboratory, Oak Ridge, TN (United States)

    1964-04-15

    extremely accurate. The disagreement found for the thick rods is to be expected when diffusion theory is used to describe the effect of absorbers with cross-sectional dimensions comparable to the neutron mean free path in the moderator. Measurements with rods of intermediate sizes are being carried out to determine the point at which diffusion theory becomes inadequate. (author) [French] Le calcul de l'efficacite des barres de commande se trouve complique par le fait qu'elle depend a la fois de la repartition energetique des neutrons et de la geometrie de l'ensemble. La comparaison entre la theorie et les resultats experimentaux obtenus sur des reacteurs ou des systemes sous- critiques souleve des difficultes, en raison de la complexite intrinseque de ces systemes. La methode d'absorption des neutrons permet d'utiliser un modele a neutrons exclusivement thermiques, dans lequel on peut dissocier la repartition energetique des neutrons.des effets spatiaux. Il s'ensuit que l'on peut etudier ie facteur geometrique de l'efficacite de la barre de commande sans ten ir compte des details du spectre des neutrons, et comparer les resultats a un dispositif experimental simple non empoisonne. La methode est fondee sur le fait que, dans une experience d'absorption des neutrons du genre de celle que decrivent les autetits, le laplacien de l'ensemble est fonction de la constante de decroissance du mode fondamental. B{sup 2} = ({lambda} - {lambda}{sub a})/D {lambda}{sub a} = inverse de la periode des neutrons dans le moderateur (s{sup -1}) D = constante de diffusion (cm{sup 2}/s) Le milieu ralentisseur utilise pour ces experiences etait constitue par des prismes rectangulaires de. beryllium, de dimensions diverses, dont les plus petits etaient des blocs de 2,54 cm de haut et 7,3 de cote. Trois types de barres de cadmium ont ete utilises: des barres minces de 0,476 cm de diametre, une barre de section cruciforme, et des barres creuses et 'epaisses' ayant une section carree de 7,3 de ctfte

  15. Nuclear fuel rods

    International Nuclear Information System (INIS)

    Wada, Toyoji.

    1979-01-01

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

  16. 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.

  17. 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.

  18. 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

  19. A cw 4-rod RFQ linac

    International Nuclear Information System (INIS)

    Fujisawa, Hiroshi

    1994-01-01

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

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

    International Nuclear Information System (INIS)

    Dias, A.F.V.

    1980-10-01

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

  1. A novel culture method reveals unique neural stem/progenitors in mature porcine iris tissues that differentiate into neuronal and rod photoreceptor-like cells.

    Science.gov (United States)

    Royall, Lars N; Lea, Daniel; Matsushita, Tamami; Takeda, Taka-Aki; Taketani, Shigeru; Araki, Masasuke

    2017-11-15

    Iris neural stem/progenitor cells from mature porcine eyes were investigated using a new protocol for tissue culture, which consists of dispase treatment and Matrigel embedding. We used a number of culture conditions and found an intense differentiation of neuronal cells from both the iris pigmented epithelial (IPE) cells and the stroma tissue cells. Rod photoreceptor-like cells were also observed but mostly in a later stage of culture. Neuronal differentiation does not require any additives such as fetal bovine serum or FGF2, although FGF2 and IGF2 appeared to promote neural differentiation in the IPE cultures. Furthermore, the stroma-derived cells were able to be maintained in vitro indefinitely. The evolutionary similarity between humans and domestic pigs highlight the potential for this methodology in the modeling of human diseases and characterizing human ocular stem cells. Copyright © 2017 Elsevier B.V. All rights reserved.

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

    Directory of Open Access Journals (Sweden)

    Zeman V.

    2011-06-01

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

  3. Fuel rod simulator effects in flooding experiments single rod tests

    International Nuclear Information System (INIS)

    Nishida, M.

    1984-09-01

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

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

    International Nuclear Information System (INIS)

    Lin Yibing; Pan Liangcai; Yin Shunjiu

    1991-09-01

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

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

    International Nuclear Information System (INIS)

    Wand, H.

    1982-01-01

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

  6. 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

  7. 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

  8. 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

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

    Directory of Open Access Journals (Sweden)

    S.Ul. Islam

    2016-09-01

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

  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. 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.)

  12. Cuisenaire Rods Go to College.

    Science.gov (United States)

    Chinn, Phyllis; And Others

    1992-01-01

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

  13. 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

  14. 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.)

  15. Une offre publique de documents ?

    OpenAIRE

    Tesnière, Valérie

    2017-01-01

    TOUT NUMÉRIQUE ? LE LIVRE CONCURRENCÉ ? LA BIBLIOTHÈQUE CONCURRENCÉE ? La fracture numérique reste une réalité comme l’attestent encore des chiffres cités par Le Monde : une évolution de la couverture Internet dans le monde à l’horizon 2009-2030 donnerait la projection suivante : 24,7 % des 6,8 milliards d’habitants de la planète sont connectés en 2009 et l’on atteindrait 50 % pour 8,2 milliards d’habitants en 2030. Cela interroge indirectement le statut futur du papier : sera-t-il l’apanage ...

  16. Une introduction à MATLAB c

    OpenAIRE

    CREMONA, Christian; LABORATOIRE CENTRAL DES PONTS ET CHAUSSEES - LCPC

    2002-01-01

    MATLAB c présente toutes les fonctionnalités des approches récentes de la programmation : programmation objet basée sur des hiérarchies de classes, programmation événementielle du graphisme. MATLAB c présente une aide en ligne très complète sous format html des différentes fontions accessibles. COMPTE RENDU DE RECHERCHE

  17. Neural signal processing for identifying failed fuel rods in nuclear reactors

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  18. 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

  19. Ultrasonic inspection for testing the PWR fuel rod endplug welds

    International Nuclear Information System (INIS)

    Pillet, C.; Destribats, M.T.; Papezyk, F.

    1976-01-01

    A method of ultrasonic testing with local immersion and transversal waves was developed. It is possible to detect defects as the lacks of fusion and penetration and porosity in the PWR fuel rod endplug welds [fr

  20. 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.)

  1. A method for scientific code coupling in a distributed environment; Une methodologie pour le couplage de codes scientifiques en environnement distribue

    Energy Technology Data Exchange (ETDEWEB)

    Caremoli, C; Beaucourt, D; Chen, O; Nicolas, G; Peniguel, C; Rascle, P; Richard, N; Thai Van, D; Yessayan, A

    1994-12-01

    This guide book deals with coupling of big scientific codes. First, the context is introduced: big scientific codes devoted to a specific discipline coming to maturity, and more and more needs in terms of multi discipline studies. Then we describe different kinds of code coupling and an example of code coupling: 3D thermal-hydraulic code THYC and 3D neutronics code COCCINELLE. With this example we identify problems to be solved to realize a coupling. We present the different numerical methods usable for the resolution of coupling terms. This leads to define two kinds of coupling: with the leak coupling, we can use explicit methods, and with the strong coupling we need to use implicit methods. On both cases, we analyze the link with the way of parallelizing code. For translation of data from one code to another, we define the notion of Standard Coupling Interface based on a general structure for data. This general structure constitutes an intermediary between the codes, thus allowing a relative independence of the codes from a specific coupling. The proposed method for the implementation of a coupling leads to a simultaneous run of the different codes, while they exchange data. Two kinds of data communication with message exchange are proposed: direct communication between codes with the use of PVM product (Parallel Virtual Machine) and indirect communication with a coupling tool. This second way, with a general code coupling tool, is based on a coupling method, and we strongly recommended to use it. This method is based on the two following principles: re-usability, that means few modifications on existing codes, and definition of a code usable for coupling, that leads to separate the design of a code usable for coupling from the realization of a specific coupling. This coupling tool available from beginning of 1994 is described in general terms. (authors). figs., tabs.

  2. Le poids de l'histoire: A la recherche d'une pedagogie (The Weight of History: In Search of a Method).

    Science.gov (United States)

    Bourbeau-Walker, Micheline

    1984-01-01

    It is proposed that while the sciences have progressed steadily, language teaching methods have swung like a pendulum between two broad approaches: formal and functional. The history of this pattern is outlined, current practices are discussed, and the possibility of escaping from this polarizing cycle is examined. (MSE)

  3. Etude de la mesure des emittances transverses d'un faisceau par la methode des gradients. Application au cas d'une focalisation par solenoide

    CERN Document Server

    M'Garrech, S

    2002-01-01

    Within the framework of the ALTO project (Linear Accelerator near the Orsay Tandem), IPNO will recover the LAL station NEPAL, which will be used as the ALTO injecting system. To calculate the beam optics through the linear accelerator, it is necessary to determine the electron beam emittance at the exit of the buncher station. There are several methods to determine this emittance: direct methods, like the Pepper Pot technique, and indirect ones, like the three distances method and the three gradients one. The latter requires a variable optic element (quadrupole, solenoid...). In the case of the use of a solenoid, the horizontal and vertical motions are coupled, which implies an additional difficulty for the analysis of the measurements. The main goal of this report is to identify and to solve these mathematical difficulties, so as to determine at last the initial emittance. The treated example comes from a paper by R.Chehab et al [12], the code used is BETA, and resolution is done using the last square method...

  4. Discrete Charge Effects on an Infinitely Long Cylindrical Rod Model

    OpenAIRE

    Agung, Ahmad A. J; Jesudason, Christopher G.

    2011-01-01

    Two methods for determining the potential (\\psi) around a discretely charged rod have been devised. The methods utilize the potential around the continuously charged rod (\\bar{\\psi}) as the reference where \\bar{\\psi} isdetermined by the Poisson-Boltzmann equation. The potential data are used to determine the theoretical radial distribution function (RDF) which is compared with MD simulation data. It is shown that the magnitude of the charge and size parameters very strongly affects the shape ...

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

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  6. A novel combined method of osteosynthesis in treatment of tibial fractures: a comparative study on sheep with application of rod-through-plate fixator and bone plating.

    Science.gov (United States)

    Tralman, G; Andrianov, V; Arend, A; Männik, P; Kibur, R T; Nõupuu, K; Uksov, D; Aunapuu, M

    2013-04-01

    The study compares the efficiency of a new bone fixator combining periostal and intramedullary osteosynthesis to bone plating in treatment of tibial fractures in sheep. Experimental osteotomies were performed in the middle third of the left tibia. Animals were divided into two groups: in one group (four animals) combined osteosynthesis (rod-through-plate fixator, RTP fixator) was applied, and in the other group (three animals) bone plating was used. The experiments lasted for 10 weeks during which fracture union was followed by radiography, and the healing process was studied by blood serum markers reflecting bone turnover and by histological and immunohistochemical investigations. In the RTP fixator group, animals started to load body weight on the operated limbs the next day after the surgery, while in the bone plating group, this happened only on the seventh day. In the RTP fixator group, consolidation of fractures was also faster, as demonstrated by radiographical, histological, and immunohistochemical investigations and in part by blood serum markers for bone formation. It can be concluded that application of RTP fixation is more efficient than plate fixation in the treatment of experimental osteotomies of long bones in sheep. © 2012 Blackwell Verlag GmbH.

  7. A new method for studying iodine metabolism; the isotopic equilibrium method - kinetic and quantitative aspects of measurements made on rats; Une nouvelle methode d'etude du metabolisme de l'iode: la methode d'equilibre isotopique - aspects cinetiques et quantitatifs obtenus chez le rat

    Energy Technology Data Exchange (ETDEWEB)

    Simon, C. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-05-15

    plasma avec une sensibilite de 0,001 {mu}g {sup 127}I. Cette sensibilite a permis de doser des pools aussi petits que l'iodure et les iodotyrosines libres de la thyroide ainsi que de demontrer l'absence des iodotyrosines libres dans le plasma du rat normal. In vivo, la methode d'equilibre isotopique a permis de mesurer le contenu en iode de la thyroide et de calculer l'intensite de la secretion de cette glande sans qu'il soit necessaire de la prelever. Par double marquage avec {sup 125}I et {sup 131}I, la methode d'equilibre isotopique a permis de mesurer l'intensite du flux de recyclage intrathyroidien ainsi que les vitesses de renouvellement de tous les composes iodes de la thyroide. Pour cette glande, aucune relation precurseur-produit n'a ete trouvee entre les iodotyrosines (MIT et DIT) et les iodothyronines (T{sub 4} et T{sub 3}). L'absence de cette relation resulte de l'heterogeneite de renouvellement de la thyroglobuline. D'autre part, il a ete montre qu'il existait dans le plasma une fraction d'iode organique differente de la thyroglobuline et se renouvelant plus rapidement que les hormones circulantes T{sub 4} et T{sub 3}. La methode d'equilibre isotopique est tres favorable aux dosages d'iode en grandes series. Elle permet de plus d'ameliorer les fractionnements biochimiques en ajoutant des entraineurs sans que le dosage ulterieur d'iode {sup 127}I soit perturbe. (auteur)

  8. A new method for studying iodine metabolism; the isotopic equilibrium method - kinetic and quantitative aspects of measurements made on rats; Une nouvelle methode d'etude du metabolisme de l'iode: la methode d'equilibre isotopique - aspects cinetiques et quantitatifs obtenus chez le rat

    Energy Technology Data Exchange (ETDEWEB)

    Simon, C [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-05-15

    The isotopic equilibrium method which has been developed in the case of the rat has made it possible to measure the absolute values of the principal parameters of iodine metabolism in this animal. The quantities and concentrations of iodine have been measured in the thyroid gland and in the plasma with a sensitivity of 0.001 {mu}g of {sup 127}I. This sensitivity has made it possible to measure pools as small as the iodide and the free iodotyrosines of the thyroid and to demonstrate the absence of free iodotyrosines in the plasma of the normal rat. In vivo, the isotopic equilibrium method has made it possible to measure the iodine content of the thyroid gland and to calculate the intensity of this gland's secretion without removing it. By double labelling with {sup 125}I and {sup 131}I the isotopic equilibrium method has made it possible to measure the flux, intensity of the intrathyroidal recycling as well as the turnover rates of all the iodine containing compounds of the thyroid gland. For this gland no precursor-product relationship has been found between The iodotyrosines (MIT and DIT) and the iodothyronines (T{sub 4} and T{sub 3}). The absence of this relationship is due to the heterogeneity of the thyroglobulin turnover. It has been shown furthermore that there exists in the plasma an organic fraction of the iodine which is different to thyroglobulin and which is renewed more rapidly than the circulating hormones T{sub 3} and T{sub 4}. The isotopic equilibrium method is very useful for series measurements of iodine. It makes it possible furthermore to improve the biochemical fractionations by adding carriers without affecting the subsequent {sup 127}I measurements. (author) [French] La methode d'equilibre isotopique, mise au point chez le rat, a permis de mesurer en valeur absolue les principaux parametres du metabolisme de l'iode chez cet animal. Les quantites ou les concentrations d'iode ont ete mesurees pour la thyroide et pour le plasma avec une

  9. 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.)

  10. Spacers for fuel rod clusters

    International Nuclear Information System (INIS)

    Jabsen, F.S.

    1978-01-01

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

  11. Gamma-ray spectroscopy on irradiated fuel rods

    International Nuclear Information System (INIS)

    Terremoto, Luis Antonio Albiac

    2009-01-01

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

  12. 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

  13. 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.)

  14. 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.)

  15. Variation in sensitivity, absorption and density of the central rod distribution with eccentricity.

    Science.gov (United States)

    Tornow, R P; Stilling, R

    1998-01-01

    To assess the human rod photopigment distribution and sensitivity with high spatial resolution within the central +/-15 degrees and to compare the results of pigment absorption, sensitivity and rod density distribution (number of rods per square degree). Rod photopigment density distribution was measured with imaging densitometry using a modified Rodenstock scanning laser ophthalmoscope. Dark-adapted sensitivity profiles were measured with green stimuli (17' arc diameter, 1 degrees spacing) using a T ubingen manual perimeter. Sensitivity profiles were plotted on a linear scale and rod photopigment optical density distribution profiles were converted to absorption profiles of the rod photopigment layer. Both the absorption profile of the rod photopigment and the linear sensitivity profile for green stimuli show a minimum at the foveal center and increase steeply with eccentricity. The variation with eccentricity corresponds to the rod density distribution. Rod photopigment absorption profiles, retinal sensitivity profiles, and the rod density distribution are linearly related within the central +/-15 degrees. This is in agreement with theoretical considerations. Both methods, imaging retinal densitometry using a scanning laser ophthalmoscope and dark-adapted perimetry with small green stimuli, are useful for assessing the central rod distribution and sensitivity. However, at present, both methods have limitations. Suggestions for improving the reliability of both methods are given.

  16. AU-DELA DES METHODES TRADITIONNELLES DANS LACQUISITION DES COMPETENCES LINGUISTIQUES DANS UNE LANGUE ETRANGERE – LE BLOG, UN OUTIL PRATIQUE?

    Directory of Open Access Journals (Sweden)

    BOGDAN Rodica

    2014-12-01

    Full Text Available Although the classical and traditional methods have proven effective in seizing on foreign language skills by pupils and students, new means of communication and dissemination of information - in the virtual world via Internet - are continuous and opportune challenges towards progress. The purpose of this paper is to question the effectiveness of new instruments - namely the blog - for conveying and assimilating the skills necessary for communicating in another language. We live in a world in which a wide range of public figures stemming from diverse environments, such as the academia, mass-media, political and cultural environment, etc. use this tool - the blog - in order to present in a personal manner their knowledge or opinions, to promote a different kind of dialogue with their readers, with the audience, with their constituency or with those interested in the proposed topics. The journalist and the blogger are often tantamount. Great online publications have a blog button on their site. In such conditions, is the blog a practical tool for both the teacher and the economy student learning a new language? A series of pertinent questions are posed: to which extent can the blog be used as an instrument in teaching and learning a foreign language? Is the virtual platform a proper place for designing, displaying and posting both practical and theoretical topics? If preparing such a topic - at an economy student's hands - entails a significant effort as a self-taught, is this effort best assessed when presenting a paper before a professor, in the interaction between the two, which coincides with the evaluation and assessment of the student's work? While the information content is taught the same for all students, regarding one on one assimilation and evaluation, could an individual blog - moderated by the student - or a collective blog in which all students have access be more appropriate? Is the use of a virtual platform more effective than

  17. Scattering of electromagnetic waves by an non-uniform cylindrical plasma; Diffusion coherente d'une onde electromagnetique par un cylindre de plasma inhomogene

    Energy Technology Data Exchange (ETDEWEB)

    Faugeras, P E [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires. Groupe de recherches sur la fusion controlee

    1967-07-01

    The problem of the scattering of plane electromagnetic waves from a non-uniform, cylindrically symmetrical plasma is solved analytically, by a self-consistent field method, for a wave with the electric field parallel to the cylinder axis. Numerical results for the diffracted field are plotted for interesting ranges of the parameters involved: diameter, density on the axis, radial profile of the density, and collision frequencies. The case where the incident field is cylindric (waves surfaces parallel to the cylinder axis) is examined - this permits to connect theoretical calculations and experimental diffraction patterns, and also to explain the diffraction effects observed in a classical microwave interferometry experiment. These results, and the possibility of measuring exactly the diffracted field (showed by experiments with dielectric and metallic rods) lead to a new plasma diagnostic method, based on the diffraction, which has no theoretical limitations and it usable when the classical free-space wave methods are not (plasma diameter lower than 10 wave lengths). The feasibility of this method is tested with a plasma at atmospheric pressure and a 2 mm incident wavelength. The plasma is obtained by the laminar flow of a plasma torch, with a working gas (He or Ar) seeded by potassium (density continuously variable between 10{sup 11} and 10{sup 15} e/cm{sup 3}. Some diffraction patterns by this plasma and for various incident waves, are also given and explained with theoretical calculations. (author) [French] On etudie la diffusion coherente d'une onde electromagnetique par un cylindre de plasma inhomogene par une methode de champ self-consistant, et pour une onde de vecteur electrique parallele a l'axe du cylindre. On a calcule le champ diffracte en faisant varier le diametre du cylindre, la densite sur l'axe, le profil de densite et les frequences de collisions, et on donne ici les principaux resultats. On examine ensuite le cas d'une onde incidente cylindrique

  18. 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.)

  19. Segmented fuel and moderator rod

    International Nuclear Information System (INIS)

    Doshi, P.K.

    1987-01-01

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

  20. The testing of a method for dosing plutonium by {alpha}-counting in the presence of strong concentrations of salts or of uranium; Essai d'une methode de dosage du plutonium par comptage {alpha} en presence de fortes concentrations en sels ou en uranium

    Energy Technology Data Exchange (ETDEWEB)

    Fontaine, A M; Baude-Malafosse, L M; Cunq, M J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1961-07-01

    This report describes a method for dosing small quantities of plutonium in a solution having a high concentration of salts. It shows the possibility of dosing up to 5.10{sup -3} {mu}g of Pu in the presence of 10 mg of NaNO{sub 3} with out decreasing the counting-rate. The only error possible is that in the counting. It is also possible to dose 10{sup -3} {mu}g of Pu in the presence of 1,7 mg of uranyl nitrate. (author) [French] Ce rapport decrit une methode de dosage de faibles quantites de plutonium dans une solution de forte concentration en sels. Il montre la possibilite de doser jusqu'a 5.10{sup -3} {mu}g de Pu en presence de 10 mg de NO{sub 3}Na sans diminution du taux de comptage. La seule erreur que l'on puisse faire est l'erreur de comptage. On peut aussi doser 10{sup -3} {mu}g de Pu en presence de 1,7 mg de nitrate d'uranyle. (auteur)

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

    Directory of Open Access Journals (Sweden)

    Zafer Özdemir

    2000-06-01

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

  2. Application de la methode de la reponse frequentielle a l'arret "SSFR", sur une machine synchrone a poles saillants de grande puissance

    Science.gov (United States)

    Belqorchi, Abdelghafour

    , being influenced by the measurement current magnitude, the latter was maintained constant in the range 1mHz-20Hz. Other problems such as the rotation impact on damper circuits or the difficulty of rotor positioning are eliminated or attenuated by the intrinsic characteristics of the machine. Regarding the data analysis, the Maximum Likelihood Estimation (MLE) method was used to determine the third and second order equivalent circuit from SSFR measurements. In d-axis, the approaches of adjustment to two and three transfer functions (Ld(s), sG(s) and Lafo(s)) were explored. The second order model, derived from (Ld( s) and G(s)), was used to deduce the machine standard parameters. The latter were compared with the values given by the manufacturer and by conventional on-site tests: Instantaneous three-phase short-circuit, Dalton-Cameron and the d-axis transient time constant at open stator (T'do). The comparison showed the good accuracy of SSFR values. Subsequently, a machine model was built in EMTP-RV based on SSFR standard parameters. The model was able to reproduce stator and rotor currents measured during instantaneous three-phase short-circuit test. Some adjustments, to SSFR parameters, were needed to reproduce stator voltage and rotor current acquired during load rejection d-axis test. It is worthwhile noting that the load rejection d-axis test, recently added to IEEE 115-2009 annex, must be modified to take into account the saturation and excitation impedance impact on deduced parameters. Regarding this issue, some suggestions are proposed by the author. The obtained SSFR results, contribute to raise confidence on SSFR application on large salient pole machines. In addition, it shows the aptitude of the SSFR model to represent the machine in both cases of weak and strong disturbances, at least on machines similar the one studied. Index Terms: Salient pole, frequency response, SSFR, equivalent circuit, operational inductance.

  3. Fabrication of internally instrumented reactor fuel rods

    International Nuclear Information System (INIS)

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

    1975-01-01

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

  4. 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

  5. Fuel rod behavior of a PWR during load following

    International Nuclear Information System (INIS)

    Perrotta, J.A.; Andrade, G.G. de

    1982-01-01

    The behavior of a PWR fuel rod when operating in normal power cycles, excluding in case of accidents, is analysed. A computer code, that makes the mechanical analysis of the cladding using the finite element method was developed. The ramps and power cycles were simulated suposing the existence of cracks in pellets when the cladding-pellet interaction are done. As a result, an operation procedure of the fuel rod in power cycle is recommended. (E.G.) [pt

  6. Photonic mesophases from cut rod rotators

    Energy Technology Data Exchange (ETDEWEB)

    Stelson, Angela C.; Liddell Watson, Chekesha M., E-mail: cml66@cornell.edu [Materials Science and Engineering, Cornell University, Ithaca, New York 14853 (United States); Avendano, Carlos [Chemical Engineering and Analytical Science, The University of Manchester, Manchester M13 9PL (United Kingdom)

    2016-01-14

    The photonic band properties of random rotator mesophases are calculated using supercell methods applied to cut rods on a hexagonal lattice. Inspired by the thermodynamic mesophase for anisotropic building blocks, we vary the shape factor of cut fraction for the randomly oriented basis. We find large, stable bandgaps with high gap isotropy in the inverted and direct structures as a function of cut fraction, dielectric contrast, and filling fraction. Bandgap sizes up to 34.5% are maximized at high dielectric contrast for rods separated in a matrix. The bandgaps open at dielectric contrasts as low as 2.0 for the transverse magnetic polarization and 2.25 for the transverse electric polarization. Additionally, the type of scattering that promotes the bandgap is correlated with the effect of disorder on bandgap size. Slow light properties are investigated in waveguide geometry and slowdown factors up to 5 × 10{sup 4} are found.

  7. Fuel Rod Flow-Induced Vibration Overview

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-10-15

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

  8. 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)

  9. 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.)

  10. Insuffler une énergie nouvelle : donner une voix et une visibilité aux ...

    International Development Research Centre (IDRC) Digital Library (Canada)

    22 oct. 2010 ... J'ai accepté, car cette méthode nous permettait de tenir une véritable consultation auprès des jeunes du Brésil. CRDI – Pouvez-vous nous expliquer en quoi consiste cette méthode ? Grzybowski – Les sondages d'opinion sont importants, mais c'est un peu comme si on tâtait le pouls d'un patient. De même ...

  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. Vibration characteristics of a long flexible rod supported with multiple gaps

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  13. 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.

  14. Radioactive lightning rods waste treatment

    International Nuclear Information System (INIS)

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

    2008-01-01

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

  15. La résistance de vague des carènes. Calcul de la fonction de Green par intégration numérique et par une méthode asymptotique. 1° Partie Hull Resistance to Wave? Computing the Green Function by Numerical Integration and by an Asymptotic Method. Part One

    Directory of Open Access Journals (Sweden)

    Carou A.

    2006-11-01

    Full Text Available Le calcul de la résistance de vague d'une carène par éléments finis concentrés sur un ouvert borné nécessite la connaissance de la fonction de Green du problème à grande distance. Cette fonction est très difficile à calculer numériquement. On justifie dans ce travail une méthode asymptotique rapide, remplaçant avantageusement l'intégration numérique. Computing wave resistance -by finite elements concentrated on a bounded open set requires the prior knowledge of the Green function of the problem at a great distance. Computing this function is numerically very difficult. A fast asymptotic method is iustified in this article, and it can be used ta advantage as a replacemenf for numerical integration.

  16. Vers une cartographie géo-lexicale

    Directory of Open Access Journals (Sweden)

    William Martinez

    2011-06-01

    Full Text Available L’analyse statistique de la distribution du vocabulaire dans des guides de tourisme du XIXe et XXe siècle permet d’identifier la terminologie descriptive essentielle telle qu’elle est employée dans ces textes. Une interprétation plus structurée de ces données est possible grâce aux méthodes de cooccurrence qui produisent des cartes de mots associés décrivant l’usage préféré de noms, adjectifs, adverbes, etc. à propos d’une ville, une région ou un itinéraire. À partir de cette visualisation originale de structures de mots, nous envisagerons la convergence des données lexicales et cartographiques dans une base de données de type SIG (Système d’Information Géographique.The statistical analysis of vocabulary distribution in French tourist guides of the 19th and 20th century reveals the essential descriptive terminology used in these texts. A more structured interpretation of this data is made possible by way of co-occurrence methods that produce associated word maps describing the preferred usage of nouns, adjectives, adverbs etc. regarding a given town, region or route. Based on this original visualizing of word structures we will consider the convergence of lexical and cartographic data in a GIS-type database (Geographic Information System.

  17. The Third ATLAS ROD Workshop

    CERN Multimedia

    Poggioli, L.

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

  18. Une herboristerie ethnique à Paris

    OpenAIRE

    Hamaïdi , Maurad

    2012-01-01

    Place de la Chapelle à Paris, dans le 18e arrondissement. L'enseigne de ce magasin est peu explicite, mais la vitrine laisse deviner le type de produits vendus. L'information est un peu plus développée dans la langue arabe car il y est précisé que la vente concerne tous types d'encens, ainsi que des plantes arabes : il s'agit d'une herboristerie. Il est également écrit que le magasin exporte vers le Maroc : le mot en arabe est ambigu puisqu'il s'agit de "Maghreb", mais en général, utilisé seu...

  19. Przegląd najczęściej stosowanych metod fizjoterapii w zespołach bólowych kręgosłupa wśród pacjentów NZOZ ŚROD-MED w Policach = The review of most commonly used physiotherapy methods in spineaches based on analysis of patient documents in NZOZ ŚROD-MED Police

    Directory of Open Access Journals (Sweden)

    Katarzyna Korabiusz

    2016-05-01

                        Results: This article is based on analysis of patient documents in NZOZ ŚROD- MED. As we can see the problems with spine are more connected to woman that man.                    Conclusions: There are more physical therapy than kinesiotherapy or massage. The most commonly physiotherapy method used in 2008 was TP and traction, and LASER and manual therapy in the 2013.

  20. 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.)

  1. 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.)

  2. Maximum/minimum asymmetric rod detection

    International Nuclear Information System (INIS)

    Huston, J.T.

    1990-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-07-15

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

  4. Refabricated and instrumented fuel rods

    International Nuclear Information System (INIS)

    Silberstein, K.

    2005-01-01

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

  5. Prototypical Rod Consolidation Demonstration Project

    International Nuclear Information System (INIS)

    1993-05-01

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

  6. Prototypical Rod Consolidation Demonstration Project

    International Nuclear Information System (INIS)

    1993-05-01

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

  7. 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.)

  8. 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.

  9. Lifting device for drilling rods

    Energy Technology Data Exchange (ETDEWEB)

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

    1982-01-01

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

  10. Prototypical Rod Consolidation Demonstration Project

    International Nuclear Information System (INIS)

    1993-05-01

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

  11. Contrast vaginography is more accurate than the radiopaque rod for localization of the vagina

    International Nuclear Information System (INIS)

    Wiggenraad, Ruud G.; Coerkamp, Emile G.; Tamminga, Reinder I.; Wiersma, Tjeerd G.; Sorge, Adriaan A. von

    2000-01-01

    Purpose: To compare the radiopaque vaginal rod method with contrast vaginography in localization of the vagina. Methods and Materials: In 25 female patients who needed pelvic radiotherapy, both our standard localization procedure using the vaginal rod and a localization procedure using contrast vaginography were performed. As a rod can change the position of the vagina, contrast vaginography was considered to display the true anatomic position of the vagina. The corresponding rod and nonrod X-rays of each patient were compared. The distance from the true vaginal apex to the displaced vaginal apex (= the top of the rod) was measured in the sagittal plane. This distance was called the inaccuracy of the rod method. Furthermore, the size of the vaginal vault was measured using the contrast vaginography. Results: The median inaccuracy of the rod method was 13 mm (range 2 to 24 mm). The maximal width of the vagina ranged from 24 to 68 mm in the frontal plane (median 39 mm) and from 3 to 22 mm in the sagittal plane (median 10 mm). Conclusion: The rod method is not accurate to localize the vagina. Furthermore, the rod gives no information on the actual size of the vaginal vault. Contrast vaginography is the method of choice to localize the vagina.

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

    International Nuclear Information System (INIS)

    Sauer, A.

    1989-01-01

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

  13. Cadmium safety rod thermal tests

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  14. 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

  15. 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)

  16. Skin bridge versus rod colostomy in children - comparison between complications.

    Science.gov (United States)

    Askarpour, Shahnam; Peyvasteh, Mehran; Changai, Bahram; Javaherizadeh, Hazhir

    2012-10-01

    Due to economic problems, sigmoid loop colostomy using glass rod may cause problems for our patients for finding glass rod and several visits. The aim of the study was to compare rod versus skin bridge colostomy. In this study, 42 cases who are candidate for colostomy were included. Cases were randomly placed in skin bridge and rod colostomy group. Independent sample t-test and Chi-square were used for comparison. SPSS version 16.0 (SPSS Inc, Chicago, IL, USA) was used for analysis. Of 42 cases, 20 were male and 22 were female. Hirschsprung's disease was the indication of colostomy in 33 cases. In nine cases, imperforate anus was the indication of colostomy. Mean time of surgery was 79.4 and 82.5 minute for the rod and skin bridge group respectively (P>0.05). Retraction was seen in 2 case of rod group, and no case of skin bridge group. Prolapse was seen in 2 (9.5%) case of rod group and 1(4.7%) case in skin bridge. There were no reports of necrosis, stenosis, and hernia in both groups. In the skin bridge group the rates of complications were lower but the groups are too small for statistical analysis. Colostomy with a skin bridge method may decrease number of revision and expenses and may be appropriate option. Sigmoid loop colostomy using skin bridge flap may be appropriate choice in developing country. Another study with more samples is recommended to better comparison of Skin Bridge versus rod colostomy.

  17. Commissioning of a passive rod scanner at INB

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-01

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

  18. Chitin Fiber and Chitosan 3D Composite Rods

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  19. Chitin Fiber and Chitosan 3D Composite Rods

    Directory of Open Access Journals (Sweden)

    Zhengke Wang

    2010-01-01

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

  20. 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

  1. 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)

  2. Xanthomatose normolipidemique a localisation nasale chez une ...

    African Journals Online (AJOL)

    Introduction : Exposer un cas de xanthomatose normolipidémique. Observation : il s'agit d'une adolescente de 18 ans qui a présenté une masse des cavités nasales dont le bilan paraclinique a plaidé en faveur d'un xanthogranulome juvénile. L'évolution à court terme a été satisfaisante après l'exérèse chirurgicale par une ...

  3. Flow resistance in rod assemblies

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  4. Nuclear fuel rod loading apparatus

    International Nuclear Information System (INIS)

    King, H.B.

    1981-01-01

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

  5. 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.)

  6. Generation of heat on fuel rod in cosine pattern by using induction heating

    International Nuclear Information System (INIS)

    Keettikkal, Felix; Sajeesh, Divya; Rao, Poornima; Hande, Shashank; Dakave, Ganesh; Kute, Tushar; Mahajan, Akshay; Kulkarni, R.D.

    2017-01-01

    Fuel rods are used in a nuclear reactor for fission process. When these rods are cooled by water during the heat transfer, the temperature stress causes undesirable defects in the fuel rod. Studying these defects occurring in the fuel rod in the nuclear cluster during nuclear reaction is a difficult task because fission reaction makes it difficult to analyse the changes in the rod. Hence there is a need to use a replica of the rod with similar thermal stress to study and analyse the rod for the defects. Normally the heat generated on the fuel rod follows a cosine pattern which is an inherent characteristic inside a nuclear reactor. In view of this, in this paper induction heating method is used on a rod to create an exact replica of the cosine pattern of heat by varying the pitch of the coil. First, a MATLAB simulation is done using simulink. Then a prototype of the model has been developed comprising of carbon steel pipe, with length and outside diameter of 1 meter and 48.2 mm, respectively. Instead of using water as coolant, rod is simulated in air. Therefore, the heat generated is lost by normal convection and radiation. Non-nuclear testing can be a valuable tool in the development or in some kind of experiment using nuclear reactor. Induction heating becomes an alternative to classical heating technologies because of its advantages such as efficiency, quickness, safety, clean heating and accurate power control. (author)

  7. Experiment and numerical simulation of bubbly two-phase flow across horizontal and inclined rod bundles

    International Nuclear Information System (INIS)

    Serizawa, A.; Huda, K.; Yamada, Y.; Kataoka, I.

    1997-01-01

    Experimental and numerical analyses were carried out on vertically upward air-water bubbly two-phase flow behavior in both horizontal and inclined rod bundles with either in-line or staggered array. The inclination angle of the rod bundle varied from 0 to 60 with respect to the horizontal. The measured phase distribution indicated non-uniform characteristics, particularly in the direction of the rod axis when the rods were inclined. The mechanisms for this non-uniform phase distribution is supposed to be due to: (1) Bubble segregation phenomenon which depends on the bubble size and shape: (2) bubble entrainment by the large scale secondary flow induced by the pressure gradient in the horizontal direction which crosses the rod bundle; (3) effects of bubble entrapment by vortices generated in the wake behind the rods which travel upward along the rod axis; and (4) effect of bubble entrainment by local flows sliding up along the front surface of the rods. The liquid velocity and turbulence distributions were also measured and discussed. In these speculations, the mechanisms for bubble bouncing at the curved rod surface and turbulence production induced by a bubble were discussed, based on visual observations. Finally, the bubble behaviors in vertically upward bubbly two-phase flow across horizontal rod bundle were analyzed based on a particle tracking method (one-way coupling). The predicted bubble trajectories clearly indicated the bubble entrapment by vortices in the wake region. (orig.)

  8. A study of the deformability of composite rods with end fittings for truss structures

    Energy Technology Data Exchange (ETDEWEB)

    Stepanychev, E I; Novikov, V V; Sukhanov, A V; Lapotkin, V A; Postnov, A N [Moskovskii Aviatsionnyi Tekhnologicheskii Institut, Moscow (USSR)

    1989-04-01

    A method is proposed for studying the deformability characteristics of composite rods with end fittings under static loading. Three types of connectors for tubular composite rods are examined from the standpoint of deformability, and differences in the deformability of the central part of a rod and of the connection zone are discussed. The effect of thermal cycling on the deformability of structures of this type is analyzed. 7 refs.

  9. Inspection system for Zircaloy clad fuel rods

    International Nuclear Information System (INIS)

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

    1975-10-01

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

  10. 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

  11. A study on the thermal hydraulics in rod bundles

    International Nuclear Information System (INIS)

    Chung, Moon Ki; Yang, Sun Kyu

    1989-03-01

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

  12. Une masse palpébrale révélant une fistule carotidocaverneuse ...

    African Journals Online (AJOL)

    ... huit mois une masse palpébrale droite, avec une discrète exophtalmie et hémorragie sous conjonctivale, l'angio IRM a permis de confirmer le diagnostique d'une fistule carotidocaverneuse à haut débit, qui est responsable de cette symptomatologie. L'objectif de cet article est de mettre la lumière sur cette pathologie rare, ...

  13. Rod internal pressure quantification and distribution analysis using Frapcon

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-01

    /MTU is determined to be the total fuel rod void volume and the amount of released fission gas in the fuel rod, respectively. Cumulative distribution functions (CDFs) are prepared from the distribution of RIP and CHS predictions for all standard and IFBA rods. The provided CDFs allow for the determination of the portion of WBN1 fuel rods that exceed a specified RIP or CHS limit. Results are separated into IFBA and standard rods so that the two groups may be analyzed individually. FRAPCON results are provided in sufficient detail to enable the recalculation of the RIP while considering any desired plenum gas temperature, total void volume, or total amount of gas present in the void volume. A method to predict the CHS from a determined or assumed RIP is also proposed, which is based on the approximately linear relationship between the CHS and the RIP. Finally, improvements to the computational methodology of FRAPCON are proposed.

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  15. 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)

  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. Solitary waves on nonlinear elastic rods. II

    DEFF Research Database (Denmark)

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

    1987-01-01

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

  18. ELECTRIC FIELD MEASUREMENT IN ROD-DISCONTINUED ...

    African Journals Online (AJOL)

    2014-06-30

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

  19. 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.)

  20. 21 CFR 876.4270 - Colostomy rod.

    Science.gov (United States)

    2010-04-01

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

  1. Spider and burnable poison rod combinations

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  2. Self-Assembly of Rod-Coil Block Copolymers

    National Research Council Canada - National Science Library

    Jenekhe, S

    1999-01-01

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

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

    International Nuclear Information System (INIS)

    Wilson, W.; Comstock, R.J.

    1999-01-01

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

  4. Oxide nano-rod array structure via a simple metallurgical process

    International Nuclear Information System (INIS)

    Nanko, M; Do, D T M

    2011-01-01

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

  5. 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)

  6. 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)

  7. Synthesis of Vertically Aligned ZnO Nano rods on Various Substrates

    International Nuclear Information System (INIS)

    Hassan, J.J.; Hassan, Z.; Abu Hassan, H.; Mahdi, M.A.

    2011-01-01

    We successfully synthesized vertically aligned ZnO nano rods on Si, GaN, Sic, Al 2 O 3 , ITO, and quartz substrates using microwave assisted chemical bath deposition (MA-CBD) method. All these types of substrates were seeded with PVA-ZnO nano composites layer prior to the nano rods growth. The effect of substrate type on the morphology of the ZnO nano rods was studied. The diameter of grown ZnO nano rods ranged from 50 nm to 200 nm. Structural quality and morphology of ZnO nano rods were determined by x-ray diffraction and scanning electron microscopy, which revealed hexagonal wurtzite structures perpendicular to the substrate along the z-axis in the direction of (002). Photoluminescence measurements of grown ZnO nano rods on all substrates exhibited high UV peak intensity. Raman scattering studies were conducted to estimate the lattice vibration modes. (author)

  8. 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

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

    International Nuclear Information System (INIS)

    Boehm, H.H.; Foerch, H.

    1984-01-01

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

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

    International Nuclear Information System (INIS)

    Gadelha, J.A.M.

    1982-10-01

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

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

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  12. Nuclear reactor fuel rod behavior modelling and current trends

    International Nuclear Information System (INIS)

    Colak, Ue.

    2001-01-01

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

  13. 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

  14. 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

  15. Individual nuclear fuel rod weighing system

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  16. Individual nuclear fuel rod weighing system

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  17. Automated nuclear fuel rod pattern loading system

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  18. ELECTROMAGNETIC APPARATUS FOR MOVING A ROD

    Science.gov (United States)

    Young, J.N.

    1958-04-22

    An electromagnetic apparatus for moving a rod-like member in small steps in either direction is described. The invention has particular application in the reactor field where the reactor control rods must be moved only a small distance and where the use of mechanical couplings is impractical due to the high- pressure seals required. A neutron-absorbing rod is mounted in a housing with gripping uaits that engage the rod, and coils for magnetizing the gripping units to make them grip, shift, and release the rod are located outside the housing.

  19. 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

  20. Rod cluster having improved vane configuration

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  1. Prototypical Rod Consolidation Demonstration Project

    International Nuclear Information System (INIS)

    1993-05-01

    The objective of Phase 3 of the Prototypical Rod consolidation Demonstration Project (PRCDP) was to procure, fabricate, assemble, and test the Prototypical Rod consolidation System as described in the NUS Phase 2 Final Design Report. This effort required providing the materials, components, and fabricated parts which makes up all of the system equipment. In addition, it included the assembly, installation, and setup of this equipment at the Cold Test Facility. During the Phase 3 effort the system was tested on a component, subsystem, and system level. This volume 1, discusses the PRCDP Phase 3 Test Program that was conducted by the HALLIBURTON NUS Environmental Corporation under contract AC07-86ID12651 with the United States Department of Energy. This document, Volume 1, Book 1 discusses the following topics: the background of the project; test program description; summary of tests and test results; problem evaluation; functional requirements confirmation; recommendations; and completed test documentation for tests performed in Phase 3

  2. Prototypical Rod Consolidation Demonstration Project

    International Nuclear Information System (INIS)

    1993-05-01

    The objective of Phase 3 of the Prototypical Rod consolidation Demonstration Project (PRCDP) was to procure, fabricate, assemble, and test the Prototypical Rod consolidation System as described in the NUS Phase 2 Final Design Report. This effort required providing the materials, components, and fabricated parts which makes up all of the system equipment. In addition, it included the assembly, installation, and setup of this equipment at the Cold Test Facility. During the Phase 3 effort the system was tested on a component, subsystem, and system level. This volume 1, discusses the PRCDP Phase 3 Test Program that was conducted by the HALLIBURTON NUS Environmental Corporation under contract AC07-86ID12651 with the United States Department of Energy. This document, Volume 1, Book 9 discusses the following topics: Integrated System Normal Operations Test Results and Analysis Report; Integrated System Off-Normal Operations Test Results and Analysis Report; and Integrated System Maintenance Operations Test Results and Analysis Report

  3. Prototypical Rod Consolidation Demonstration Project

    International Nuclear Information System (INIS)

    1993-05-01

    The objective of Phase 3 of the Prototypical Rod consolidation Demonstration Project (PRCDP) was to procure, fabricate, assemble, and test the Prototypical Rod consolidation System as described in the NUS Phase 2 Final Design Report. This effort required providing the materials, components, and fabricated parts which makes up all of the system equipment. In addition, it included the assembly, installation, and setup of this equipment at the Cold Test Facility. During the Phase 3 effort the system was tested on a component, subsystem, and system level. This volume 1, discusses the PRCDP Phase 3 Test Program that was conducted by the HALLIBURTON NUS Environmental Corporation under contract AC07-86ID12651 with the United States Department of Energy. This document, Volume 1, Book 8 discusses Control System SOT Tests Results and Analysis Report. This is a continuation of Book 7

  4. Prototypical Rod Construction Demonstration Project

    International Nuclear Information System (INIS)

    1993-05-01

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

  5. 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

  6. 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.)

  7. 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.)

  8. Hollow rods for the oil producing industry

    Energy Technology Data Exchange (ETDEWEB)

    Khalimova, L M; Elyasheva, M A

    1970-01-01

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

  9. Study of the machining of uranium carbide rods obtained by continuous casting under electronic bombardment; Etude de l'usinage de barreaux de carbure d'uranium obtenus par coulee continue sous bombardement electronique

    Energy Technology Data Exchange (ETDEWEB)

    Rousset, P; Accary, A [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1965-07-01

    The authors consider the various methods of machining uranium mono-carbide and compare them critically in the case of their application to uranium carbide obtained by fusion under an electronic bombardment and continuous casting. This study leads them to propose two mechanical machining methods: cylindrical rectification and center-less rectification, preceded by a preliminary roughing out of a cylinder, the latter appearing more suitable. A study of the machining yields as a function of the diameter of the rough bars and of the diameter of the finished rods has shown that an optimum value of the rough bar diameter exists for each value of the finished rod diameter. It is found that the yield increases as the diameter itself increases, this yield rising from 45 per cent to around 70 per cent as the diameter of the rough bars increases from 25-26 mm to 37-38 mm. (authors) [French] Les auteurs envisagent les differentes methodes d'usinage du monocarbure d'uranium et se livrent a une etude critique de celles-ci, dans le cas de leur application a l'usinage de barreaux de carbure d'uranium obtenus par fusion sous bombardement electronique et coulee continue. Cette etude les conduit a proposer deux methodes d'usinage mecanique: la rectification cylindrique et la rectification 'centerless', precedee d'un ebauchage par carottage, la seconde paraissant la plus appropriee. L'etude des rendements d'usinage en fonction du diametre des barreaux bruts et du diametre des barreaux finis, a mis en evidence une valeur optimale du diametre des barreaux bruts pour chaque valeur du diametre des barreaux usines. Elle a montre que le rendement croit lorsque le diametre croit lui-meme, ce rendement passant d'environ 45 pour cent a environ 70 pour cent, lorsque le diametre des barreaux bruts passe de 25-26 mm a 37-38 mm.

  10. Nuclear Fuel Test Rod Fabrication for Data Acquisition Test

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  11. Nuclear Fuel Test Rod Fabrication for Data Acquisition Test

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

  12. Comparison of different Maxwell solvers coupled to a PIC resolution method of Maxwell-Vlasov equations; Evaluation de differents solveurs Maxwell pour la resolution de Maxwell-Vlasov par une methode PIC

    Energy Technology Data Exchange (ETDEWEB)

    Fochesato, Ch. [CEA Bruyeres-le-Chatel, Dept. de Conception et Simulation des Armes, Service Simulation des Amorces, Lab. Logiciels de Simulation, 91 (France); Bouche, D. [CEA Bruyeres-le-Chatel, Dept. de Physique Theorique et Appliquee, Lab. de Recherche Conventionne, Centre de Mathematiques et Leurs Applications, 91 (France)

    2007-07-01

    The numerical solution of Maxwell equations is a challenging task. Moreover, the range of applications is very wide: microwave devices, diffraction, to cite a few. As a result, a number of methods have been proposed since the sixties. However, among all these methods, none has proved to be free of drawbacks. The finite difference scheme proposed by Yee in 1966, is well suited for Maxwell equations. However, it only works on cubical mesh. As a result, the boundaries of complex objects are not properly handled by the scheme. When classical nodal finite elements are used, spurious modes appear, which spoil the results of simulations. Edge elements overcome this problem, at the price of rather complex implementation, and computationally intensive simulations. Finite volume methods, either generalizing Yee scheme to a wider class of meshes, or applying to Maxwell equations methods initially used in the field of hyperbolic systems of conservation laws, are also used. Lastly, 'Discontinuous Galerkin' methods, generalizing to arbitrary order of accuracy finite volume methods, have recently been applied to Maxwell equations. In this report, we more specifically focus on the coupling of a Maxwell solver to a PIC (Particle-in-cell) method. We analyze advantages and drawbacks of the most widely used methods: accuracy, robustness, sensitivity to numerical artefacts, efficiency, user judgment. (authors)

  13. 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.)

  14. Prototypical Rod Consolidation Demonstration Project

    International Nuclear Information System (INIS)

    1993-05-01

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

  15. 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)

  16. Axial-flow-induced vibration for a rod supported by translational springs at both ends

    International Nuclear Information System (INIS)

    Kang, H.S.; Song, K.N.; Kim, H.K.; Yoon, K.H.

    2003-01-01

    An axial-flow-induced vibration model was proposed for a rod supported by two translational springs at both ends in order to evaluate the sensitivity to spring stiffness on the FIV for a PWR fuel rod. For developing the model, a one-mode approximation was made based on the assumption that the first mode was dominant in vibration behavior of the single span rod. The first natural frequency and mode shape functions for the flow-induced vibration, called the FIV, model were derived by using Lagrange's method. The vibration displacements were calculated by both of the spring-supported rod and the simple-supported (SS) one. As a result, the vibration displacement for the spring-supported (50 kN m -1 ) rod was 15-20% larger than that of the SS rod when the rods are in axial flow of 5-8 m s -1 velocity. The discrepancy between both displacements became much larger as flow velocity increased, and that of the rod having the short span length was larger than that of the rod having the long span length although the displacement value itself of the long span rod was larger than that of the short one. The vibration displacement for the spring-supported rod appeared to decrease with the increase of the spring constant. Since single span beam supported by the two translational springs are focused on in this paper, further study will be needed to reflect more realistic supporting conditions of the PWR fuel rod such as two springs and four dimples and cross or swirling flow caused by the mixing vane of the spacer grid

  17. 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)

  18. Considerations about the utilization of electrically heated rods used for simulation of nuclear fuel pins

    International Nuclear Information System (INIS)

    Lima, R. de C.F. de; Carajilescov, P.

    1987-01-01

    The dinamic behavior of electrically heated rods used for simulation of nuclear fuel pins in nuclear power transients, is analysed by the application of the lumped parameter and the finite difference methods. Deviations of the rods surface conditions, for extreme accidental transient conditions are presented and discussed. (author) [pt

  19. 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)

  20. Evaluation of a method for correction of scatter radiation in thorax cone beam CT; Evaluation d'une methode de correction du rayonnement diffuse en tomographie du thorax avec faisceau conique

    Energy Technology Data Exchange (ETDEWEB)

    Rinkel, J.; Dinten, J.M. [CEA Grenoble (DTBS/STD), Lab. d' Electronique et de Technologie de l' Informatique, LETI, 38 (France); Esteve, F. [European Synchrotron Radiation Facility (ESRF), 38 - Grenoble (France)

    2004-07-01

    Purpose: Cone beam CT (CBCT) enables three-dimensional imaging with isotropic resolution. X-ray scatter estimation is a big challenge for quantitative CBCT imaging of thorax: scatter level is significantly higher on cone beam systems compared to collimated fan beam systems. The effects of this scattered radiation are cupping artefacts, streaks, and quantification inaccuracies. The beam stops conventional scatter estimation approach can be used for CBCT but leads to a significant increase in terms of dose and acquisition time. At CEA-LETI has been developed an original scatter management process without supplementary acquisition. Methods and Materials: This Analytical Plus Indexing-based method (API) of scatter correction in CBCT is based on scatter calibration through offline acquisitions with beam stops on lucite plates, combined to an analytical transformation issued from physical equations. This approach has been applied with success in bone densitometry and mammography. To evaluate this method in CBCT, acquisitions from a thorax phantom with and without beam stops have been performed. To compare different scatter correction approaches, Feldkamp algorithm has been applied on rough data corrected from scatter by API and by beam stops approaches. Results: The API method provides results in good agreement with the beam stops array approach, suppressing cupping artefact. Otherwise influence of the scatter correction method on the noise in the reconstructed images has been evaluated. Conclusion: The results indicate that the API method is effective for quantitative CBCT imaging of thorax. Compared to a beam stops array method it needs a lower x-ray dose and shortens acquisition time. (authors)

  1. 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)

  2. 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)

  3. Radiography inspection of weld for nuclear fuel rod

    International Nuclear Information System (INIS)

    Zhang Kai; Zhang Xichang

    1995-05-01

    The survey of radiography inspection, advantages, disadvantages and applications of main kinds of radiography inspection methods are presented. Emphasis is put upon the structure and functions of X-ray flaw detecting device for nuclear fuel rod welds, the actuating program of the device, as well as the structure of some key mechanism and the functions of them. The analysis is made upon the actuating principles. Finally, the test of long-term operation proves the device to be stable in operation, reliable in action, to possess high level of automation and high sensitivity and it can simultaneously perform on-line X-ray inspection of 25 nuclear fuel rods with a diameter less than 10 mm, and meet the requirements of large-scale production of nuclear fuel rods (5 figs.)

  4. Torsion of DNA modeled as a heterogeneous fluctuating rod

    Science.gov (United States)

    Argudo, David; Purohit, Prashant K.

    2014-01-01

    We discuss the statistical mechanics of a heterogeneous elastic rod with bending, twisting and stretching. Our model goes beyond earlier works where only homogeneous rods were considered in the limit of high forces and long lengths. Our methods allow us to consider shorter fluctuating rods for which boundary conditions can play an important role. We use our theory to study structural transitions in torsionally constrained DNA where there is coexistence of states with different effective properties. In particular, we examine whether a newly discovered left-handed DNA conformation called L-DNA is a mixture of two known states. We also use our model to investigate the mechanical effects of the binding of small molecules to DNA. For both these applications we make experimentally falsifiable predictions.

  5. Tensile Characterization of FRP Rods for Reinforced Concrete Structures

    Science.gov (United States)

    Micelli, F.; Nanni, A.

    2003-07-01

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

  6. Development of a unified sizing method for gas radiation heating facilities used in large-volume buildings; Developpement d'une methode de dimensionnement unifiee d'installations de chauffage radiant gaz pour les batiments de grand volume

    Energy Technology Data Exchange (ETDEWEB)

    Jimenez, M.; Bego, L.; Douls, Y.; Le Dean, P.; Paradowski, V. [Gaz de France, GDF, Dir. de la Recherche, 75 - Paris (France)

    2000-07-01

    Builders now have perfect command of the natural gas heating technique used for large-volume buildings. However, the sizing of heating facilities still leaves grounds for discussion, whatever the energies actually used. Accordingly, between 1997 and 1999, the ATG (technical association of the Gas industry in France), seven French manufacturers of 'large volume' heating equipment, the Chaleur Et Rayonnement (CER) association and Gaz de France decided to collaborate and develop a 'unified sizing method' for heating facilities using radiating emitters. During the first year of the study, the above partners worked on the said method (theoretical thermal study of the radiative phenomena, and then adaptation to the methods currently used by the various manufacturers). In 1998, with the support of the ADEME (the French environment and energy control agency), the partners tested the method on five industrial buildings (studying the thermal behavior and making air renewal measurements with search gases). This work made it possible to either confirm or adapt the theoretical evaluations which had been made originally. In 1999, a software program was produced to make the developed method more user friendly. The program can be used to determine the power to be installed, but also to assess the quality of the chosen configuration of the emitters (unit power, inclination, orientation) for optimum customer comfort. (authors)

  7. Quantum dynamics through a wave packet method to study electron-hydrogen and atom-dihydrogen collisions; Dynamique quantique par une methode de paquets d'ondes. Etude des collisions electron-hydrogene et atome-dihydrogene

    Energy Technology Data Exchange (ETDEWEB)

    Mouret, L

    2002-11-01

    The thesis concerns the development and implementation of numerical methods for solving the time-dependent Schroedinger equation. We first considered the case of electron-hydrogen scattering. The originality of our method is the use of a non-uniform radial grid defined by a Schwarz interpolation based on a Coulomb reference function. This grid allows many hydrogen bound states and associated matrix elements of various operators to be reproduced to machine accuracy. The wave function is propagated in time using a Split-Operator method. The efficiency of our method allows the wave function to be propagated out to large distances for all partial waves. We obtain excitation and ionization cross sections in excellent agreement with the best experimental and theoretical data. We subsequently adapted the method and the program package to study reactive atom-dihydrogen scattering. The wave packet is described using product Jacobi coordinates on a regular grid of radial coordinates combined with a basis of Legendre polynomials for the angular part (partial wave S). The wave function is analysed using a time-to-energy Fourier transform, which provides results over the energy range covered by the initial wave packet in one calculation. The method was first tested on the quasi-direct (F,H2) reaction and then applied to the indirect (C(1D),H2)reaction. The state-to-state reaction probabilities are in good agreement with those obtained by a time-independent approach. In particular, the strongly resonant structure of the (C(1D),H2) reaction probabilities is well reproduced. (author)

  8. Quantum dynamics through a wave packet method to study electron-hydrogen and atom-dihydrogen collisions; Dynamique quantique par une methode de paquets d'ondes. Etude des collisions electron-hydrogene et atome-dihydrogene

    Energy Technology Data Exchange (ETDEWEB)

    Mouret, L

    2002-11-01

    The thesis concerns the development and implementation of numerical methods for solving the time-dependent Schroedinger equation. We first considered the case of electron-hydrogen scattering. The originality of our method is the use of a non-uniform radial grid defined by a Schwarz interpolation based on a Coulomb reference function. This grid allows many hydrogen bound states and associated matrix elements of various operators to be reproduced to machine accuracy. The wave function is propagated in time using a Split-Operator method. The efficiency of our method allows the wave function to be propagated out to large distances for all partial waves. We obtain excitation and ionization cross sections in excellent agreement with the best experimental and theoretical data. We subsequently adapted the method and the program package to study reactive atom-dihydrogen scattering. The wave packet is described using product Jacobi coordinates on a regular grid of radial coordinates combined with a basis of Legendre polynomials for the angular part (partial wave S). The wave function is analysed using a time-to-energy Fourier transform, which provides results over the energy range covered by the initial wave packet in one calculation. The method was first tested on the quasi-direct (F,H2) reaction and then applied to the indirect (C(1D),H2)reaction. The state-to-state reaction probabilities are in good agreement with those obtained by a time-independent approach. In particular, the strongly resonant structure of the (C(1D),H2) reaction probabilities is well reproduced. (author)

  9. A New Method of Estimating Wind Tunnel Wall Interference in the Unsteady Two-Dimensional Flow (Nouvelle Methode D’Estimation de la Perturbation des Ecoulements Instationnaires par les Parois d’une Soufflerie).

    Science.gov (United States)

    1983-01-01

    disturbance theory . The main feature of the method is the use of measured pressure along lines in the flow direction near the tunnel walls. This method...disturbance theory , then $can be written ( , = qo( , ) .@ (:. S-in(.t + 0.( or s CO (8) Defining cw as co S . ^(9) gives Sin= C, f(4,.) + OCr,z)co.s(0t...AUTHOR (S)/ AUTEUR (S) H. Sawada, visiting scientist 2nd Aerodynamics Division, National Aerospace Laboratory, Japan SERIES/SERIE Aeronautical Note 6

  10. Evaluation of droplet deposition in rod bundle

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  11. Model investigation of fuel rod behaviour

    International Nuclear Information System (INIS)

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

    1985-06-01

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

  12. Vibrational characteristics and wear of fuel rods

    International Nuclear Information System (INIS)

    Schmugar, K.L.

    1977-01-01

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

  13. Inverse method for effects characterization from ultrasonic b-scan images; Caracterisation des defauts par une methode d'inversion lors d'un controle ultrasonore. Application au controle des defauts en paroi externe

    Energy Technology Data Exchange (ETDEWEB)

    Faur, M. [Paris-11 Univ., 91 - Orsay (France)

    1999-02-01

    In service inspections of French nuclear pressure water reactor vessels are carried out automatically in complete immersion from the inside by means of ultrasonic focused probes working in the pulse echo mode. Concern has been expressed about the capabilities of performing non destructive evaluation of the Outer Surface Defects (OSD), i.e. defects located in the vicinity of the outer surface of the inspected components. OSD are insonified by both a direct field that passes through the inner surface (water/steel) of the component containing the defect and a secondary field reflected from the outer surface. Consequently, the Bscan images, containing the signatures of such defects, are complicated and their interpretation is a difficult task. This work deals with extraction of the maximum available information for characterizing OSD from ultrasonic Bscan images. Our main objectives are to obtain the type of OSD and their geometric parameters by means of two specific inverse methods. The first method is used for the identification of the geometrical parameters of the equivalent planar OSD from segmented Bscan images. Ultrasonic equivalent defect sizing model-based methods may be used to size a defect in a material by obtaining a best-fit simple equivalent shape that matches the ultrasonic observed data. We illustrate the application of such an equivalent sizing OSD method that is based on a simplified direct model. The major drawback of this identification method, as used to date, is that only a part of the useful information contained into original Bscan image, i.e. segmented Bscan image, is used for defect characterization. Moreover, it requires the availability of defect classification information (i.e. if the defect is volumetric or planer, e. g. a crack or a lack of fusion), which, generally, may be as difficult to obtain as the defect parameters themselves. Therefore, we propose a parameter estimation method for extracting complementary information on the defect

  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. Pellet clad interaction analysis of AFA 3G fuel rod

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  16. 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

  17. Rod-drop analysis in fast and thermal spectra

    International Nuclear Information System (INIS)

    Broccoli, U.

    1988-01-01

    The application of Carpenter's method to power profiles resulting from simulated or real rod-drop events has been tested. The conditions which allow the errors to be reduced to a minimum are highlighted. The results obtained show a good agreement with simulated and experimental data. (author). 1 ref., 21 figs, 6 tabs

  18. 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

  19. 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

  20. Dynamics of Longitudinal Impact in the Variable Cross-Section Rods

    Science.gov (United States)

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

    2018-03-01

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

  1. 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.)

  2. Set up of a method for the adjustment of resonance parameters on integral experiments; Mise au point d`une methode d`ajustement des parametres de resonance sur des experiences integrales

    Energy Technology Data Exchange (ETDEWEB)

    Blaise, P.

    1996-12-18

    Resonance parameters for actinides play a significant role in the neutronic characteristics of all reactor types. All the major integral parameters strongly depend on the nuclear data of the isotopes in the resonance-energy regions.The author sets up a method for the adjustment of resonance parameters taking into account the self-shielding effects and restricting the cross section deconvolution problem to a limited energy region. (N.T.).

  3. Transient methods to characterize flows and mass transfer in a packed column by tracers; Methodes transitoires de caracterisation des ecoulements et du transfert de masse dans une colonne a garnissage a l'aide de traceurs

    Energy Technology Data Exchange (ETDEWEB)

    Perrin, S.

    1998-06-11

    The aim of this study is to propose a packed column characterization method in the form of phases flows and mass transfer model, in which the parameters are estimated by transient technique. After a bibliographic study a model is performed and validated. It allows efficiency and precision in the parameters choice. Two tracer techniques have been implemented: they show interesting possibilities of flow diagnosis. (A.L.B.)

  4. Computational fluid dynamics (CFD) round robin benchmark for a pressurized water reactor (PWR) rod bundle

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Shin K., E-mail: paengki1@tamu.edu; Hassan, Yassin A.

    2016-05-15

    Highlights: • The capabilities of steady RANS models were directly assessed for full axial scale experiment. • The importance of mesh and conjugate heat transfer was reaffirmed. • The rod inner-surface temperature was directly compared. • The steady RANS calculations showed a limitation in the prediction of circumferential distribution of the rod surface temperature. - Abstract: This study examined the capabilities and limitations of steady Reynolds-Averaged Navier–Stokes (RANS) approach for pressurized water reactor (PWR) rod bundle problems, based on the round robin benchmark of computational fluid dynamics (CFD) codes against the NESTOR experiment for a 5 × 5 rod bundle with typical split-type mixing vane grids (MVGs). The round robin exercise against the high-fidelity, broad-range (covering multi-spans and entire lateral domain) NESTOR experimental data for both the flow field and the rod temperatures enabled us to obtain important insights into CFD prediction and validation for the split-type MVG PWR rod bundle problem. It was found that the steady RANS turbulence models with wall function could reasonably predict two key variables for a rod bundle problem – grid span pressure loss and the rod surface temperature – once mesh (type, resolution, and configuration) was suitable and conjugate heat transfer was properly considered. However, they over-predicted the magnitude of the circumferential variation of the rod surface temperature and could not capture its peak azimuthal locations for a central rod in the wake of the MVG. These discrepancies in the rod surface temperature were probably because the steady RANS approach could not capture unsteady, large-scale cross-flow fluctuations and qualitative cross-flow pattern change due to the laterally confined test section. Based on this benchmarking study, lessons and recommendations about experimental methods as well as CFD methods were also provided for the future research.

  5. 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

  6. 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.)

  7. 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

  8. 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

  9. La culture, une arme de constructions massives.

    Directory of Open Access Journals (Sweden)

    Patrick Poncet

    2008-06-01

    Full Text Available Le recueil de texte que nous propose Jean-Michel Tobelem autour des questions de « diplomatie culturelle » peut être considéré comme un objet à double sens : une source et une exploration. Comme recueil de documents, le directeur d’ouvrage s’est attaché à réunir dans un volume somme toute restreint des contributions de spécialistes de divers sujets « culturels » à qui il était demandé de livrer leur savoir sous un angle valorisant le lien entre diplomatie et culture. ...

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

    Directory of Open Access Journals (Sweden)

    V. L. Danilov

    2015-01-01

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

  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. A system automatic study for the spent fuel rod cutting and simulated fuel pellet extraction device

    International Nuclear Information System (INIS)

    Jeong, J. H.; Yun, J. S.; Hong, D. H.; Kim, Y. H.; Park, K. Y.

    2001-01-01

    A fuel pellet extraction device of the spent fuel rods is described. The device consists of a cutting device of the spent fuel rods and the decladding device of the fuel pellets. The cutting device is to cut a spent fuel rod to n optimal size for fast decladding operation. To design the device, the fuel rod properties are investigated including the dimension and material of fuel rod tubes and pellets. Also, various methods of existing cutting method are investigated. The design concepts accommodate remote operability for the Hot-Cell(radioactive ) area operation. Also, the modularization of the device structure is considered for the easy maintenance. The decladding device is to extract the fuel pellet from the rod cut. To design this device, the existing method is investigated including the chemical and mechanical decladding methods. From the view point of fuel recovery and feasibility of implementation. it is concluded that the chemical decladding method is not appropriate due to the mass production of radioactive liquid wastes, in spite of its high fuel recovery characteristics. Hence, in this paper, the mechanical decladding method is adopted and the device is designed so as to be applicable to various lengths of rod-cuts. As like the cutting device,the concepts of remote operability and maintainability is considered. Both devices are fabricated and the performance is investigated through a series of experiments. From the experimental result, the optimal operational condition of the devices is established

  13. Expandable device for a nuclear fuel rod

    International Nuclear Information System (INIS)

    Gesinski, L.T.

    1978-01-01

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

  14. 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.)

  15. Temperature actuated automatic safety rod release

    Science.gov (United States)

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

    1984-03-13

    A temperature-actuated apparatus is disclosed for releasably supporting a safety rod in a nuclear reactor, comprising a safety rod upper adapter having a retention means, a drive shaft which houses the upper adapter, and a bimetallic means supported within the drive shaft and having at least one ledge which engages a retention means of the safety rod upper adapter. A pre-determined increase in temperature causes the bimetallic means to deform so that the ledge disengages from the retention means, whereby the bimetallic means releases the safety rod into the core of the reactor.

  16. Absorber rod drive for nuclear reactors

    International Nuclear Information System (INIS)

    Acher, H.

    1985-01-01

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

  17. Acoustic loading effects on oscillating rod bundles

    International Nuclear Information System (INIS)

    Lin, W.H.

    1980-01-01

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

  18. Growth and Morphology of Rod Eutectics

    Energy Technology Data Exchange (ETDEWEB)

    Jing Teng; Shan Liu; R. Trivedi

    2008-03-17

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

  19. Nuclear reactor fuel rod attachment system

    International Nuclear Information System (INIS)

    Christiansen, D.W.

    1983-01-01

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

  20. Dry rod consolidation technology development

    International Nuclear Information System (INIS)

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

    1986-01-01

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