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

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

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)

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

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.)

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.)

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.

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

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.)

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)

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)

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

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

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)

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

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)

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)

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.)

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

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

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

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

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.)

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

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

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.)

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).

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)

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)

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

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.

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

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.

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)

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.)

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.)

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.)

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.)

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

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)

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.

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

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.

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

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.

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

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

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

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)

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

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

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

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

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.)

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.

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)

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

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)

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

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

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)

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

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

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

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

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)

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

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.)

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

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)

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

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

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.)

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.

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)

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

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.

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

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

{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)

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)

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.

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)

Modeling and simulation performance of sucker rod beam pump

Energy Technology Data Exchange (ETDEWEB)

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

2015-09-30

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

Modeling and simulation performance of sucker rod beam pump

International Nuclear Information System (INIS)

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

2015-01-01

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

Control rod

International Nuclear Information System (INIS)

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

1987-01-01

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

Control rods

International Nuclear Information System (INIS)

Maruyama, Hiromi.

1984-01-01

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

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....

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.)

Automated nuclear fuel rod pattern loading system

International Nuclear Information System (INIS)

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

1991-01-01

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

Control rod 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

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

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

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)

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.)

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

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

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

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.)

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

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.)

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.)

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

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.)

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

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

Morphoelastic rods. Part I: A single growing elastic rod

KAUST Repository

Moulton, D.E.

2013-02-01

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

Morphoelastic rods. Part I: A single growing elastic rod

KAUST Repository

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

2013-01-01

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

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)

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

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

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

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

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)

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)

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

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.)

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.

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

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, ...

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)

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.)

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)

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

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)

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)

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

Science.gov (United States)

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

2006-01-28

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

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.

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

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

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)

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)

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.)

Nuclear fuel rod end plug weld inspection

International Nuclear Information System (INIS)

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

1985-01-01

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

Control rod drives

International Nuclear Information System (INIS)

Futatsugi, Masao.

1980-01-01

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

Control rod drives

International Nuclear Information System (INIS)

Oonuki, Koji.

1981-01-01

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

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.

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)

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

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

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.)

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

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

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)

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.

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.)

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

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)

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.

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.)

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.

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.

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

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.

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)

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

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)

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.)

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

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.)

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)

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

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

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

Une hanche douloureuse révélatrice d'une histiocytose osseuse multifocale

Science.gov (United States)

Lahrach, Kamal; Alaoui, Adil; el Kadi, Khalid Ibn; Marzouki, Amine; Boutayeb, Fawzi

2014-01-01

L'histiocytose langerhansienne est une maladie rare qui touche principalement l'enfant et l'adulte jeune. Elle peut prendre plusieurs aspects, L'atteinte osseuse peut être uni- ou multifocale. Nous rapportons une observation d'histiocytose langerhansienne osseuse multifocale, révélée chez un patient âgé de 23 ans et ayant touché le col fémoral droit. La scintigraphie osseuse a permis de retrouver plusieurs localisations: l'os temporal, humérale et scapulaire droite. En raison du risque fracturaire, le patient a bénéficié d'une ostéosynthèse par vis-plaque DHS avec curetage biopsie de la lésion. L'examen histologique a révélé une histiocytose langerhansienne. L'évolution fut favorable après chimiothérapie par voie générale. PMID:25018840

Analysis of RSG-GAS Control Rod Worth Due to Perturbation Reactivity

International Nuclear Information System (INIS)

Taswanda Taryo

2004-01-01

The control rod interaction effect of RSG-GAS typical working core was studied using a method based on the exact perturbation theory with three simplifying assumptions which require only N+1 criticality calculations. The interaction effect between two interacting rods reached up to 19 % while the interaction effect of multiple interacting rods reached up to 32 % for all (8) control rods involved. The accuracy of the adopted method were extensively investigated to determine the error sources and the magnitude of the error. Through comparison of the present results with ones of the simple summation method, it was obvious that the adopted method was superior in that a significant improvement on the accuracy of the calculated reactivity worth can be achieved with a small number of criticality calculations. (author)

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)

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)

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.)

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.)

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.

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.)

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

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.

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)

Henri Lopès : d’Une Quête Incessante à une identité plurielle

Directory of Open Access Journals (Sweden)

Vincent Simédoh

2006-04-01

Full Text Available Dans plusieurs oeuvres d’Henri Lopès, les personnages sont constamment en quête d’une identité. Une quête qu’ils entament comme un parcours puisque désormais ils vivent au confluent de plusieurs cultures et ils se sentent comme en exil. Ce n’est pas forcément un exil extérieur parce qu’ils ont quitté leur pays d’origine mais c’est plutôt un exil intérieur. Ainsi les personnages se sentent incapables de se situer par rapport à eux-mêmes et à autrui. Dès lors surgit une tentation, celle de se réfugier dans le souvenir qui n’existe même plus. La mémoire s’estompe. L’histoire elle-même fuit et disparaît. Ce qui fait naître chez les personnages un vide qui conduit à un malaise identitaire, d’où déchirement. En face de cet écartèlement, surgit une multitude de pistes. Et c’est dans cette perspective qu’Henri Lopès propose une identité plurielle.

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.

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

International Nuclear Information System (INIS)

Lima, R. de C.F. de.

1984-01-01

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

Impact d'une modulation duale sur les performances d'une liaison ...

African Journals Online (AJOL)

Le présent document présente la technique de modulation duale Fréquence - Amplitude dans le cas d'une liaison optique du type IM-DD. Ce travail révèle que la modulation duale Fréquence - Amplitude permet de générer un signal à bande latérale unique. Les performances d'une liaison optique IM-DD basée sur cette ...

Nuclear fuel rod loading apparatus

International Nuclear Information System (INIS)

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

1980-01-01

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

Control rod shutdown system

International Nuclear Information System (INIS)

Miyamoto, Yoshiyuki; Higashigawa, Yuichi.

1996-01-01

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

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)

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

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.

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

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

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.)

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

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.)

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

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

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)

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

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

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

DEFF Research Database (Denmark)

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

2011-01-01

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

Chitin Fiber and Chitosan 3D Composite Rods

International Nuclear Information System (INIS)

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

2010-01-01

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

Commissioning of a passive rod scanner at INB

Energy Technology Data Exchange (ETDEWEB)

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

2015-07-01

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

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)

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

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

Science.gov (United States)

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

2014-02-10

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

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.

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

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)

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.

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

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.))

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)

Nuclear Fuel Test Rod Fabrication for Data Acquisition Test

International Nuclear Information System (INIS)

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

2014-01-01

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

Nuclear Fuel Test Rod Fabrication for Data Acquisition Test

Energy Technology Data Exchange (ETDEWEB)

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

2014-10-15

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

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.)

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

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.)

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)

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)

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

International Nuclear Information System (INIS)

Jeon, Yu Mi; Bae, Jun Ho; Park, Joo Hwan

2010-01-01

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

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,

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)

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

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.)

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)

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

International Nuclear Information System (INIS)

Jeon, Yu Mi; Park, Joo Hwan

2010-09-01

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

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.

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.

L'allocation universelle, une solution d'avenir?

OpenAIRE

Van Parijs, Philippe

2012-01-01

Le fait L’avenir de la sécurité sociale fait aussi débat en Suisse. Une initiative populaire vient d’être lancée “pour un revenu de base inconditionnel”. Si elle récolte 100000 signatures, cette proposition pourra être soumise à une votation pour devenir loi. Plus fort, une “initiative citoyenne européenne” sur cette question va être lancée en septembre. Objectif: un million de signatures. De quoi parle-t-on? Il s’agit d’une allocation universelle attribuée à toute personne individuellement, ...

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

International Nuclear Information System (INIS)

Baron, Clemens

2008-01-01

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

Integrated control rod monitoring device

International Nuclear Information System (INIS)

Saito, Katsuhiro

1997-01-01

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

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

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)

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)

Control rods

International Nuclear Information System (INIS)

Koga, Isao; Masuoka, Ryuzo.

1979-01-01

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

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

International Nuclear Information System (INIS)

Malang, S.

1975-11-01

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

Arthropathie destructrice des epaules au cours d�une acromegalie

Directory of Open Access Journals (Sweden)

Nessrine Akasbi

2011-10-01

Full Text Available L�acromegalie est une maladie endocrinienne rare, en rapport avec une hypersecretion d�hormone de croissance. Elle a des consequences rhumatologiques: l�arthropathie peripherique, l�atteinte rachidienne et les syndromes canalaires. L�atteinte articulaire accompagne une acromegalie active, sa survenue apres un traitement radical et une remission complete est rare. Nous presentons le cas d�une patiente de 70 ans ayant un antecedent d�acromegalie sur adenome hypophysaire il y a 25 ans, traitee chirurgicalement et declaree en remission complete, a developpe une arthropathie destructrice des deux epaules. Le but de notre observation est de mettre le point sur la possibilite d�une atteinte articulaire au cours de l�acromegalie et de son retentissement fonctionnelle.

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

International Nuclear Information System (INIS)

Figueredo, A.M. de.

1985-12-01

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

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

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)

Nonlinear tension-bending deformation of a shape memory alloy rod

International Nuclear Information System (INIS)

Shang, Zejin; Wang, Zhongmin

2012-01-01

Based on the measured shape memory alloy (SMA) stress–strain curve and the nonlinear large deformation theory of extensible beams (or rods), the first-order nonlinear governing equations of a SMA cantilever straight rod are established. They consist of a boundary-value problem of ordinary differential equations with a strong nonlinearity, in which seven unknown functions are contained and the arc length of the deformed axis is considered as one of the basic unknown functions. The shooting method combining with the Newton–Raphson iteration method is applied to solve the equations numerically. For a SMA cantilever rod subjected to a transverse uniformly distributed force, the deformation characteristics curves, the maximum strain and the maximum stress distribution curves along the longitudinal direction of rod, and the relation curves between deformation characteristic parameters and transverse uniformly force under different slenderness ratios are obtained. The effects of material nonlinearity, geometrical nonlinearity and slenderness ratio on the tension-bending deformation of the SMA cantilever rod are investigated. The numerical simulation results are in good agreement with the experimental data from the literature, verifying the soundness of the entire numerical simulation scheme. (paper)

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)

Analysis of Longitudinal Waves in Rod-Type Piezoelectric Phononic Crystals

Directory of Open Access Journals (Sweden)

Longfei Li

2016-04-01

Full Text Available Phononic crystals can be used to control elastic waves due to their frequency bands. This paper analyzes the passive and active control as well as the dispersion properties of longitudinal waves in rod-type piezoelectric phononic crystals over large frequency ranges. Based on the Love rod theory for modeling the longitudinal wave motions in the constituent rods and the method of reverberation-ray matrix (MRRM for deriving the member transfer matrices of the constituent rods, a modified transfer matrix method (MTMM is proposed for the analysis of dispersion curves by combining with the Floquet–Bloch principle and for the calculation of transmission spectra. Numerical examples are provided to validate the proposed MTMM for analyzing the band structures in both low and high frequency ranges. The passive control of longitudinal-wave band structures is studied by discussing the influences of the electrode’s thickness, the Poisson’s effect and the elastic rod inserts in the unit cell. The influences of electrical boundaries (including electric-open, applied electric capacity, electric-short and applied feedback control conditions on the band structures are investigated to illustrate the active control scheme. From the calculated comprehensive frequency spectra over a large frequency range, the dispersion properties of the characteristic longitudinal waves in rod-type piezoelectric phononic crystals are summarized.

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.)

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

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.

Le syndrome d’Usher: à propos d’une observation

OpenAIRE

Daoudi, Chama; boutimzine, Noureddine; Haouzi, Samia El; Lezrek, Omar; Tachfouti, Samira; Lezrek, Mounir; Laghmari, Mina; Daoudi, Rajae

2017-01-01

Résumé Le syndrome d'Usher est une maladie génétique comportant une double atteinte sensorielle (auditive et visuelle) appelée surdicécité. Nous rapportons l'observation d'un patient de 50 ans, issue d'un mariage consanguin présentant une surdité congénitale avec une fonction vestibulaire normale et une rétinopathie pigmentaire responsable d'une baisse bilatérale de l'acuité visuelle apparue vers l'âge de 16 ans. Cette association compose le type 2 du syndrome d'Usher, affection rare de trans...

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

The BWR Hybrid 4 control rod

International Nuclear Information System (INIS)

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

1988-01-01

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

Burnable poison rod

International Nuclear Information System (INIS)

Natsume, Tomohiro.

1988-01-01

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

Mise en place d'une base de données pour une modélisation ...

African Journals Online (AJOL)

Mise en place d'une base de données pour une modélisation hydrologique distribuée du bassin versant du Bandama (Côte d'Ivoire) : apport d'un modèle numérique d'altitude, de la télédétection et du SIG Physitel.

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

Science.gov (United States)

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

2013-04-15

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

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

International Nuclear Information System (INIS)

Becvar, J.; Saroch, V.

1977-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2008-07-01

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

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)

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

Directory of Open Access Journals (Sweden)

Engin Çetin

2015-01-01

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

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

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.

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

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.)

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.

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

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)

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)

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)

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 ...

Numerical simulation of bubble motion about a grid spacer in a rod bundle

International Nuclear Information System (INIS)

Zhang, Zheng; Hosokawa, Shigeo; Hayashi, Kosuke; Tomiyama, Akio

2009-01-01

Numerical simulations based on a three-dimensional two-way bubble tracking method are carried out to predict bubble motions in a square duct with an obstacle and in a two-by-three rod bundle with a grid spacer. Comparisons between measured and predicted bubble motions demonstrate that the two-way bubble tracking method gives good predictions for trajectories of small bubbles in the upstream side of the grid spacer in the rod bundle geometry. The predicted bubble trajectories clearly show that bubbles are apt to migrate toward the rod surface in the vicinity of the bottom of the grid spacer. Analysis of forces acting on the bubbles confirms that pressure gradient force induced by the presence of the spacer is the main cause of the bubble lateral migration toward the rod surface. Motions of steam bubbles at a nominal operating condition of a pressurized water reactor (PWR) are also predicted by using the bubble tracking method, which indicates that steam bubbles also migrate toward the rod surface at the upstream side of the spacer due to the spacer-induced pressure gradient force. (author)

Control rod velocity limiter

International Nuclear Information System (INIS)

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

1986-01-01

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

Control rod position control device

International Nuclear Information System (INIS)

Ubukata, Shinji.

1997-01-01

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

Investigation of axial power gradients near a control rod tip

Energy Technology Data Exchange (ETDEWEB)

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

2011-07-15

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

Investigation of axial power gradients near a control rod tip

International Nuclear Information System (INIS)

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

2011-01-01

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

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 ...

Migration d'une broche dans le canal médullaire cervical après une ...

African Journals Online (AJOL)

Migration d'une broche dans le canal médullaire cervical après une ostéosynthèse de l'articulation acromio-claviculaire. Bah Aliou, El Alaoui Adil. Abstract. We report the case of a 49 year-old right-handed chef with left shoulder trauma occurred during a football game. The diagnosis was fracture of the distal quarter of ...

Control rod selecting and driving device

International Nuclear Information System (INIS)

Isobe, Hideo.

1981-01-01

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

NEUTRONIC REACTOR CONTROL ROD DRIVE APPARATUS

Science.gov (United States)

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

1959-12-15

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

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

Inspecting a research reactor's control rod surface for pitting using a machine vision

International Nuclear Information System (INIS)

Tokuhiro, Akira T.; Vadakattu, Shreekanth

2005-01-01

Inspection for pits on the control rod is performed to study the degradation of the control rod material which helps estimating the service life of the control rod at UMR nuclear reactor (UMRR). This inspection task is visually inspected and recorded subjectively. The conventional visual inspection to identify pits on the control rod surface can be automated using machine vision technique. Since the in-service control rods were not available to capture images and measure number of pits and size of the pits, the applicability of machine vision method was applied on SAE 1018 steel coupons immersed in oxygen saturated de-ionized water at 30deg, 50deg and 70deg. Images were captured after each test cycle at different light intensity to reveal surface topography of the coupon surface and analyzed for number of pits and pit size using EPIX XCAP-Std software. The captured and analyzed images provided quantitative results for the steel coupons and demonstrated that the method can be applied for identifying pits on control rod surface in place of conventional visual inspection. (author)

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.

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.

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

International Nuclear Information System (INIS)

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

1997-01-01

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

ZnO Nano-Rod Devices for Intradermal Delivery and Immunization.

Science.gov (United States)

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

2017-06-15

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

Control rod drives

International Nuclear Information System (INIS)

Ikakura, Hiroaki.

1986-01-01

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

Method 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.)

Control rod drives

International Nuclear Information System (INIS)

Asano, Hiromitsu.

1979-01-01

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

Individual nuclear fuel rod weighing system

International Nuclear Information System (INIS)

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

1985-01-01

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

Measuring device for control rod driving time

International Nuclear Information System (INIS)

Tanaka, Kazuhiko; Hanabusa, Masatoshi.

1993-01-01

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

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.

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.)

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.

Une restauration « spectaculaire »

Directory of Open Access Journals (Sweden)

Muriel Verbeeck-Boutin

2010-04-01

Full Text Available Cet article replace dans le contexte d’une époque la restauration d’une Vierge à l’enfant par Joseph Van der Veken. Documentée par le commanditaire, l’intervention ne répondit pas à son attente et déboucha sur une dérestauration. Ce fait, loin d’être anecdotique, témoigne d’une mutation des mentalités et des sensibilités dans la première moitié du vingtième siècle, et de l’émergence d’une nouvelle conception de la restauration.This article replaces the restoration of a Madonna and Child by Joseph Van der Vecken within the context of a period.Documented by the patron, the intervention did not meet his expectations and resulted in a de-restoration. This fact, far from being anecdotic, is proof of the change in mentalities and sensitivities in the first half of the twentieth century and of the emergence of a new concept of restoration

ELECTROMAGNETIC APPARATUS FOR MOVING A ROD

Science.gov (United States)

Young, J.N.

1958-04-22

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

Individual nuclear fuel rod weighing system

International Nuclear Information System (INIS)

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

1985-01-01

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

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

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

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1964-08-15

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

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

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...

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

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.)

Maintaining a critical spectra within Monteburns for a gas-cooled reactor array by way of control rod manipulation

International Nuclear Information System (INIS)

Adigun, Babatunde J.; Fensin, Michael L.; Galloway, Jack D.; Trellue, Holly R.

2016-01-01

Highlights: • Tested here are 4 methods for estimating critical rod position, in Monteburns, of a reactor fuel array. • Inverse multiplication methods better predict critical rod position at the cost of more iterations. • A polynomial fit technique can predict most plutonium isotopics to within 5%. - Abstract: This burnup study examined the effect of a predicted critical control rod position on the nuclide predictability of several axial and radial locations within a 4 × 4 graphite moderated gas cooled reactor fuel cluster geometry. To achieve this, a control rod position estimator (CRPE) tool was developed within the framework of the linkage code Monteburns between the transport code MCNP and depletion code CINDER90, and four methodologies were proposed within the tool for maintaining criticality. Two of the proposed methods used an inverse multiplication approach – where the amount of fissile material in a set configuration is slowly altered until criticality is attained – in estimating the critical control rod position. Another method carried out several MCNP criticality calculations at different control rod positions, then used a linear fit to estimate the critical rod position. The final method used a second-order polynomial fit of several MCNP criticality calculations at different control rod positions to guess the critical rod position. The results showed that consistency in prediction of power densities as well as uranium and plutonium isotopics was mutual among methods within the CRPE tool that predicted critical position consistently well. While the CRPE tool is currently limited to manipulating a single control rod, future work could be geared toward implementing additional criticality search methodologies along with additional features.

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

International Nuclear Information System (INIS)

Minarcin, Miroslav; Elko, Marek

1998-01-01

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

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

International Nuclear Information System (INIS)

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

2015-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-05-15

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

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

Monitoring device for withdrawing control rods

International Nuclear Information System (INIS)

Higashigawa, Yuichi.

1985-01-01

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

Nondestructive assay of HTGR fuel rods

International Nuclear Information System (INIS)

Menlove, H.O.

1974-01-01

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

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)

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

Une étape primordiale dans l’évaluation d’une politique de développement durable

Directory of Open Access Journals (Sweden)

Karine Bourgeois

2003-11-01

Full Text Available L’évaluation des politiques publiques prend une place institutionnelle de plus en plus importante. La commande de l’évaluation donne lieu à l’élaboration d’un cahier des charges, qui précise notamment les questions évaluatives stratégiques à traiter. Pour parvenir à une définition judicieuse de ces questions, la réalisation, en amont, d’une démarche interdisciplinaire comprenant une phase d’analyse de politique complétée d’une étude des modalités de mise en oeuvre s’avère nécessaire. A travers l’exemple de l’évaluation intermédiaire du Contrat Territorial d’Exploitation (CTE en Limousin, nous tentons de montrer en quoi le graphe d’objectifs peut constituer un outil adapté pour révéler la théorie d’action de la politique publique. La conception du cahier des charges contribue également à éclairer la décision publique en favorisant l’implication et la concertation des divers partenaires. Celle-ci est d’autant plus précieuse qu’il s’agit d’une politique de développement durable, visant, entre autres, à mieux gérer les biens publics environnementaux.An important step in a sustainable policy evaluation: the conception of terms of reference, The experience of the Land Management Contract mid-term review in Limousin. Terms of reference and particularly evaluative questions definition is one of the most important stage in evaluation process. This last aims to analyse the context of the policy and its theory  of action. In order to define, in a pertinent way, priority evaluative questions, both political analysis and implementation analysis (an interdisciplinary method are required, moreover when the policy is defined in a regional and co-operation framework. In the case of the Land Management Contract (LMC mid-term review, an objective tree has been drawn up to identify which social and economical mechanisms have been implemented to achieve the main objectives of the Limousin program

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

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)

Seismic scrammability of HTTR control rods

International Nuclear Information System (INIS)

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

1990-01-01

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

Effect of absorber rods on the space-energy distribution of thermal neutrons in water

International Nuclear Information System (INIS)

Hussein, A.Z.; Eid, Y.; Hamouda, I.

1975-01-01

Thermal neutron spectra have been measured in a vectorial direction with respect to cadmium, boron-filled and copper rod elements. The rods are infinite cylinders, of 21 mm diameter, each separately immersed in an infinite water moderator fed with neutrons from the ET-RR-1 research reactor. Measurement of spectra has been carried out, in the vicinity of the rod elements, at several distances by the time-of-flight method using a chopper and also by intergral flux activation method. The measured spectra near the copper rod were compared with transport calculations of the position-dependent spectrum. The calculations, based on a realistic kernel for water, were found to yield reasonable agreement with experiment. (orig.) [de

Ostéo arthrite tuberculeuse de la cheville et spondylodiscite: une ...

African Journals Online (AJOL)

La TDM et l'IRM rachidienne (A, B, C, D) ont montré: une collection épidurale en regard de D3D4D5D6, une destruction vertébrale de D4 avec recul du mur postérieur responsable d'une déformation dorsale avec compression médullaire, une atteinte D8D9D10D11D12 avec présence d'une collection intra-osseuse, une ...

Control rod testing apparatus

International Nuclear Information System (INIS)

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

1987-01-01

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

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

International Nuclear Information System (INIS)

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

1999-10-01

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

Control rod drive shaft latch

International Nuclear Information System (INIS)

Thorp, A.G. II.

1976-01-01

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

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

International Nuclear Information System (INIS)

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

1964-08-01

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

Express diagnostics of WWER fuel rods at nuclear power plants

International Nuclear Information System (INIS)

Pavlov, S.; Amosov, S.; Sagalov, S.; Kostyuchenko, A.

2009-01-01

Higher safety and economical efficiency of nuclear power plants (NPP) call for a continuous design modification and technological development of fuel assemblies and fuel rods as well as optimization of their operating conditions. In doing so the efficiency of new fuel introduction depends on the completeness of irradiated fuel data in many respects as well as on the rapidity and cost of such data obtaining. Standard examination techniques of fuel assemblies (FA) and fuel rods (FR) intended for their use in hot cell conditions do not satisfy these requirements in full extent because fuel assemblies require preliminary cooling at NPP to provide their shipment to the research center. Expenditures for FA transportation, capacity of hot cells and expenditures for the examined fuel handling do not make it possible to obtain important information about the condition of fuel assemblies and fuel rods after their operation. In order to increase the comprehensiveness of primary data on fuel assemblies and fuel rods immediately after their removal from the reactor, inspection test facilities are widely used for these purposes. The inspection test facilities make it possible to perform nondestructive inspection of fuel in the NPP cooling pools. Moreover these test facilities can be used to repair failed fuel assemblies. The ultrasonic testing of failed fuel rods inside the fuel assembly was developed for stands of inspection and repair of TVSA WWER-1000 for the Kalinin NPP and Temelin NPP. This method was tested for eight leaking fuel assemblies WWER-440 and WWER-1000 with a burnup of ∼14 up to 38 MW·day/kgU. The ultrasonic testing proved its high degree of reliability and efficiency. The defectoscopy by means of the pulsed eddy-current method was adapted for the stand of inspection and repair of TVSA WWER-1000 for the Kalinin NPP. This method has been used at RIAR as an express testing method of FR claddings during the post-irradiation examinations of fuel assemblies WWER

Restauration et traduction : une question de philosophie

Directory of Open Access Journals (Sweden)

Pierre Leveau

2011-05-01

Full Text Available On peut faire de la restauration une mission de service public, ou une entreprise privée soumise à la loi du marché. Quel sens donner à cette activité ? Philosophiquement, la méthodologie est à réinventer. L’auteur propose pour cela de définir la restauration comme une traduction, c’est-à-dire une transmission en même temps qu’une transaction. Il formule cette hypothèse en engageant un dialogue entre le réalisme et le constructionnalisme.The conservation can be viewed as a public service and also as an economical activity. How to understand such a pratice ? From a philosophical standpoint, this question is methologically challenging. The author claims that conservation can be defined as a translation, i.e. as a transmission and a transaction. He tests this hypothesis through a confrontation between realism and constructionism.

Quand la biopsie cutanee peut etiqueter une epilepsie

Directory of Open Access Journals (Sweden)

Taoufiq Harmouch

2011-10-01

Full Text Available La maladie de Lafora (ML represente une forme rare et grave d�epilepsie myoclonique progressive. C�est une affection a transmission autosomique recessive, heterogene sur le plan genetique. Nous rapportons le cas d�une adolescente de 16 ans, issue de parents consanguins de premier degre, qui presente depuis l�age de 14 ans des crises d�epilepsie et des myoclonies. L'examen neurologique a montre un syndrome cerebelleux et une deterioration intellectuelle. La biopsie cutanee etait indispensable pour orienter le diagnostic. La ML a un pronostic constamment fatal. L�etude histologique confirme le diagnostic et l�etude moleculaire peut aider a etablir un conseil genetique.

Influence of implant rod curvature on sagittal correction of scoliosis deformity

DEFF Research Database (Denmark)

Salmingo, Remel A.; Tadano, Shigeru; Abe, Yuichiro

2014-01-01

of the implant rod’s angle of curvature during surgery and establish its influence on sagittal correction of scoliosis deformity. STUDY DESIGN: A retrospective analysis of the preoperative and postoperative implant rod geometry and angle of curvature was conducted. PATIENT SAMPLE: Twenty adolescent idiopathic......BACKGROUND CONTEXT: Deformation of in vivo–implanted rods could alter the scoliosis sagittal correction. To our knowledge, no previous authors have investigated the influence of implanted-rod deformation on the sagittal deformity correction during scoliosis surgery. PURPOSE: To analyze the changes...... scoliosis patients underwent surgery. Average age at the time of operation was 14 years. OUTCOME MEASURES: The preoperative and postoperative implant rod angle of curvature expressed in degrees was obtained for each patient. METHODS: Two implant rods were attached to the concave and convex side...

Study of pellet clad interaction defects in Dresden-3 fuel rods

International Nuclear Information System (INIS)

Pasupathi, V.; Perrin, J.S.

1979-01-01

During Cycle-3 operation of Dresden-3, fuel rod failures occurred following a transient power increase. Ten fuel rods from five of the leaking fuel assemblies were examined at Battelle's Columbus Laboratory and General Electric-Vallecitos Nuclear Center. Examinations consisted of nondestructive and destructive methods including metallography and scanning electron microscopy (SEM). Results showed the cause of fuel rod failure to be pellet clad interaction involving stress corrosion cracking. Results of SEM studies of the cladding crack surfaces and deposits on clad inner surfaces were in agreement with those reported by other investigators

ZnO Nano-Rod Devices for Intradermal Delivery and Immunization

Directory of Open Access Journals (Sweden)

Tapas R. Nayak

2017-06-01

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

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.

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 ...

Reconstitutable control rod spider assembly

International Nuclear Information System (INIS)

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

1990-01-01

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

Estimation of irradiated control rod worth

International Nuclear Information System (INIS)

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

2009-01-01

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

Rod cluster having improved vane configuration

International Nuclear Information System (INIS)

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

1989-01-01

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

Sucker rod motor

Energy Technology Data Exchange (ETDEWEB)

Radzalov, N N; Radzhabov, N A

1983-01-01

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

Duke Power Company's control rod wear program

International Nuclear Information System (INIS)

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

1990-01-01

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

Fuel rod pellet loading head

International Nuclear Information System (INIS)

Howell, T.E.

1975-01-01

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

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

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.

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.

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)

Le soja local en Allemagne, une demande soutenue, une offre qui s’étoffe

Directory of Open Access Journals (Sweden)

Recknagel Jürgen

2015-09-01

Full Text Available Depuis 2009, la sole de soja ne cesse d’augmenter en Allemagne, surtout dans le sud du pays mais aussi à l’est et un peu partout dans les régions au climat propice. Un programme national a permis de créer les références nécessaires pour un conseil qualifié apporté aux agriculteurs via un réseau de développement. Ainsi toute une filière soja est déjà bien avancée dans le sud et en cours de mise en place dans le reste du pays, pour organiser la collecte auprès des producteurs et obtenir une offre groupée pour les transformateurs pour l’alimentation animale et humaine. Le soja allemand ne veut pas rivaliser avec le soja OGM ni pour instant avec le soja non-OGM des Amériques mais vise plutôt les filières courtes régionales et le marché bio qui est encore plus déficitaire en auto-approvisionnement que le conventionnel. Avec 10 000 ha en 2014, dont 25 % en bio, le soja en Allemagne est encore loin de la surface du maïs grain (2008–2014 autour de 500 000 ha. En pratique, 20 % de ces surfaces de maïs au climat a priori propice pourraient représenter le potentiel réel pour le soja en Allemagne, sous condition de prix garantissant une rentabilité suffisante par rapport aux autres grandes cultures. Pour les programmes non-OGM des grands distributeurs, il faut de toute façon une base de production élargie comme celle proposée par l’initiative Danube-Soja, qui dans une version Europe-Soja engloberait aussi la France. Avec le verdissement de la PAC et la montée du cours du Dollar par rapport à l’Euro, les perspectives pour une augmentation de la surface de production de soja en Allemagne et en Europe paraissent bonnes. Maintenant il faut travailler sur les débouchés pour garantir l’absorption de la production grandissante par les marchés et les consommateurs à tous les niveaux : régional, national et européen.

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

International Nuclear Information System (INIS)

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

2005-06-01

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

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)

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.

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.

Le microdosage des engrais, une innovation qui profite aux femmes

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

non seulement une diminution de leurs revenus, mais doivent aussi composer avec une pénurie chronique de céréales. Au Niger, par exemple, 2,5 millions de personnes ont eu besoin d'une aide alimentaire en 2005, par suite d'une grave sécheresse (PAM, 2010). Cette situation s'est reproduite en 2010, la mauvaise.

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

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

International Nuclear Information System (INIS)

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

1982-01-01

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

KfK-analysis of the SUPER-PHENIX-1 control rod experiments. Pt. 2

International Nuclear Information System (INIS)

Giese, H.

1992-10-01

Using standard data and codes, this first analysis campaign led to results that drastically overestimated measured control rod worths with C/E ratios ranging from 1.17 to 1.28. These findings were in sharp contrast to the results of the majority of earlier control rod experiments in zero-power facilities, where C/E ratios were usually comprised between 1.0 and 1.1. Investigations were then launched to identify the origin of this discrepancy. By comparison with the calculation methods employed by the other working groups engaging in the analysis of the SPX-1 experiments it was found that the principal problem arose from the KfK/BN-procedure used for the homogenization of control rod absorber cross sections. The specific failure of this standard procedure in the case of the SPX-1 analysis is ascribed to the extremely heterogeneous structure of the SPX-1 control rods. In an attempt to improve on this point, a new method for the production of homogenized absorber cross sections was developed and applied to the SPX-1 analysis. The report concludes with a description of this method and a survey of the results obtained for SPX-1. It is found that this revised analysis leads to a significantly improved agreement of measured and calculated control rod worths and to a better consistency with the results of earlier control rod experiments in zero-power facilities. (orig.)

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

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.

Development of the spent fuel rod cutting device using the blade cutters

International Nuclear Information System (INIS)

Jung, Jae Hoo; Yoon, Ji Sup; Hong, Dong Hee; Kim, Young Hwan; Park, Gee Yong; Kim, Do Woo

2000-11-01

A spent fuel rod cutting device is to cut a spent nuclear fuel rod to optimal size for consequent decladding operation. In this paper, various properties of fuel rod, such as a dimension and material of zircaloy tubes and fuel pellets, are investigated. Also, commercially available cutting method and tools is investigated in terms of its performance. As a result, the blade cutter is selected for the design. In order to fabricate the durable blade cutter, various materials are analyzed in terms of material properties, cutter shape, and heat treatment method, etc. Also, the durability of this tool is tested by cutting the SUS tubes and zircaloy tubes. In the device design, the remote maintainability is considered so that the modularized design is accomplished. Also, the other factors considered in the design are the round shape sustainability at the cut surface, the amount of debris generation, and the fire risk, etc. Considering these design consideration, the spent fuel rod cutting device is fabricated and tested

Preliminary nuclear design for test MOX Fuel rods

Energy Technology Data Exchange (ETDEWEB)

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

1997-10-01

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

Single-phase convective heat transfer in rod bundles

International Nuclear Information System (INIS)

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

2008-01-01

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

Single-phase convective heat transfer in rod bundles

Energy Technology Data Exchange (ETDEWEB)

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

2008-04-15

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

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

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

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�

Vortex Noise from Rotating Cylindrical Rods

Science.gov (United States)

Stowell, E Z; Deming, A F

1935-01-01

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

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

International Nuclear Information System (INIS)

Gaspar Junior, Joao Carlos Aguiar

2010-01-01

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

On-line detection of key radionuclides for fuel-rod failure in a pressurized water reactor.

Science.gov (United States)

Qin, Guoxiu; Chen, Xilin; Guo, Xiaoqing; Ni, Ning

2016-08-01

For early on-line detection of fuel rod failure, the key radionuclides useful in monitoring must leak easily from failing rods. Yield, half-life, and mass share of fission products that enter the primary coolant also need to be considered in on-line analyses. From all the nuclides that enter the primary coolant during fuel-rod failure, (135)Xe and (88)Kr were ultimately chosen as crucial for on-line monitoring of fuel-rod failure. A monitoring system for fuel-rod failure detection for pressurized water reactor (PWR) based on the LaBr3(Ce) detector was assembled and tested. The samples of coolant from the PWR were measured using the system as well as a HPGe γ-ray spectrometer. A comparison showed the method was feasible. Finally, the γ-ray spectra of primary coolant were measured under normal operations and during fuel-rod failure. The two peaks of (135)Xe (249.8keV) and (88)Kr (2392.1keV) were visible, confirming that the method is capable of monitoring fuel-rod failure on-line. Copyright © 2016 Elsevier Ltd. All rights reserved.

Calculation model of non-linear dynamic deformation of composite multiphase rods

Directory of Open Access Journals (Sweden)

Mishchenko Andrey Viktorovich

2014-05-01

Full Text Available The method of formulating non-linear physical equations for multiphase rods is suggested in the article. Composite multiphase rods possess various structures, include shear, polar, radial and axial inhomogeneity. The Timoshenko’s hypothesis with the large rotation angles is used. The method is based on the approximation of longitudinal normal stress low by basic functions expansions regarding the linear viscosity low. The shear stresses are calculated with the equilibrium equation using the subsidiary function of the longitudinal shift force. The system of differential equations connecting the internal forces and temperature with abstract deformations are offered by the basic functions. The application of power functions with arbitrary index allows presenting the compact form equations. The functional coefficients in this system are the highest order rigidity characteristics. The whole multiphase cross-section rigidity characteristics are offered the sums of the rigidity characteristics of the same phases individually. The obtained system allows formulating the well-known particular cases. Among them: hard plasticity and linear elastic deformation, different module deformation and quadratic Gerstner’s low elastic deformation. The reform of differential equations system to the quasilinear is suggested. This system contains the secant variable rigidity characteristics depending on abstract deformations. This system includes the sum of the same uniform blocks of different order. The rods phases defined the various set of uniform blocks phase materials. The integration of dynamic, kinematic and physical equations taking into account initial and edge condition defines the full dynamical multiphase rods problem. The quasilinear physical equations allow getting the variable flexibility matrix of multiphase rod and rods system.

Vibrational characteristics and wear of fuel rods

International Nuclear Information System (INIS)

Schmugar, K.L.

1977-01-01

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

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

International Nuclear Information System (INIS)

Flessner, H.; Paeserack, U.

1974-01-01

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

Control rod excess withdrawal prevention device

International Nuclear Information System (INIS)

Takayama, Yoshihito.

1992-01-01

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

Hollow rods for the oil producing industry

Energy Technology Data Exchange (ETDEWEB)

Khalimova, L M; Elyasheva, M A

1970-01-01

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

Heating of leads casks. An analytical solution to the heat equation made up of a series of Laguerre functions; Echauffement des chateaux de plomb. Une solution analytique a l'equation de la chaleur constituee par une serie de fonctions de Laguerre

Energy Technology Data Exchange (ETDEWEB)

Formery, Ph; Gilles, A [Commissariat a l' Energie Atomique, Dir. des Productions, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

1968-07-01

The packing used for the transport of highly radioactive materials such as in-pile irradiated rods; have to comply to fairly strict safety standards. They should in particular resist to fire without the radioactive protection being seriously affected. The heating of a transport cask placed in a fire has been calculated by normal automatic computation methods assuming that only thermal radiation is responsible for the heating and that this obeys STEFAN'S law. Simultaneously, a purely analytical treatment has been attempted as follows. The existence of a simple solution, of the Laguerre function type, to the heat equation has been demonstrated. By superposing an infinite number of simple solutions, it is possible to produce a fairly general solution, depending on parameters, which satisfies the initial state and the limiting conditions. The parameters can be adjusted so that the temperature and the flux produced on the shell by this solution satisfy approximately STEFAN'S relationship. (authors) [French] Les emballages qui servent au transport de produits fortement radioactifs, tels que des barreaux irradies dans les piles, doivent satisfaire a des normes de securite assez strictes. Ils doivent, en particulier, resister au feu sans que la protection contre le rayonnement soit sensiblement entamee. L'echauffement, par seul rayonnement thermique suppose obeir a la loi de STEFAN, d'un chateau de transport plonge dans un feu a ete calcule par les methodes habituelles du calcul automatique. Parallelement a ete tentee l'approche purement analytique que voici: Une solution simple, du type fonction de LAGUERRE, a l'equation de la chaleur est mise en evidence. La superposition, en nombre infini, de solutions simples, permet de fabriquer une solution assez generale dependant de parametres, satisfaisant a l'etat initial et aux conditions aux limites. Les parametres peuvent etre ajustes de facon que la temperature et le flux engendres sur la coque par cette solution

Autostéréogramme d'une montagne gaussienne -utilisant une structure paradoxale périodique comme texture de camouflage-

OpenAIRE

Colonna , Jean-François

2011-01-01

Autostereogram -using a periodical paradoxal structure as a desguise texture- with an hidden gaussian mountain (Autostéréogramme d'une montagne gaussienne -utilisant une structure paradoxale périodique comme texture de camouflage-)

Heat flux and temperature determination on the control rod outer surface

International Nuclear Information System (INIS)

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

2011-01-01

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

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

International Nuclear Information System (INIS)

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

1998-01-01

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

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)

Auto adaptation incluant une double mobilité logicielle et physique ...

African Journals Online (AJOL)

Dans ce papier, nous étudions les possibilités d'auto adaptation que pourrait offrir un calcul réparti incluant une double mobilité : d'une part une mobilité physique des sites supports d'exécution et, d'autre part, une mobilité des composants logiciels métiers ou clients. Cette mobilité de niveau logiciel est envisagée dans le ...

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 ...

Testing of LWR fuel rods to support criticality safety analysis of transport accident conditions

Energy Technology Data Exchange (ETDEWEB)

Purcell, P.C. [BNFL International Transport, Spent Fuel Services (United Kingdom); Dallongeville, M. [COGEMA Logistics (AREVA Group) (France)

2004-07-01

For the transport of low enriched materials, criticality safety may be demonstrated by applying pessimistic modelling assumptions that bound any realistic case. Where Light Water Reactor (LWR) fuel is being transported, enrichment levels are usually too high to permit this approach and more realistic data is needed. This requires a method by which the response of LWR fuel under impact accident conditions can be approximated or bounded. In 2000, BNFL and COGEMA LOGISTICS jointly commenced the Fuel Integrity Project (FIP) whose objective was to develop such methods. COGEMA LOGISTICS were well advanced with a method for determining the impact response of unirradiated fuel, but required further test data before acceptance by the Transport Regulators. The joint project team extensively discussed the required inputs to the FIP, from which it was agreed that BNFL would organise new tests on both unirradiated and irradiated fuel samples and COGEMA LOGISTICS would take major responsibility for evaluating the test results. Tests on unirradiated fuel rod samples involved both dynamic and quasi-static loading on fuel samples. PWR fuel rods loaded with uranium pellets were dropped vertically from 9m onto a rigid target and this was repeated on BWR fuel rods, similar tests on empty fuel rods were also conducted. Quasi-static tests were conducted on 530 mm long PWR and BWR fuel specimens under axial loading. Tests on irradiated fuel samples were conducted on high burn-up fuel rods of both PWR and BWR types. These were believed original to the FIP project and involved applying bending loads to simply supported pressurised rod specimens. In one test the fuel rod was heated to nearly 500oC during loading, all specimens were subject to axial impact before testing. Considerable experience of fuel rod testing and new data was gained from this test programme.

Testing of LWR fuel rods to support criticality safety analysis of transport accident conditions

International Nuclear Information System (INIS)

Purcell, P.C.; Dallongeville, M.

2004-01-01

For the transport of low enriched materials, criticality safety may be demonstrated by applying pessimistic modelling assumptions that bound any realistic case. Where Light Water Reactor (LWR) fuel is being transported, enrichment levels are usually too high to permit this approach and more realistic data is needed. This requires a method by which the response of LWR fuel under impact accident conditions can be approximated or bounded. In 2000, BNFL and COGEMA LOGISTICS jointly commenced the Fuel Integrity Project (FIP) whose objective was to develop such methods. COGEMA LOGISTICS were well advanced with a method for determining the impact response of unirradiated fuel, but required further test data before acceptance by the Transport Regulators. The joint project team extensively discussed the required inputs to the FIP, from which it was agreed that BNFL would organise new tests on both unirradiated and irradiated fuel samples and COGEMA LOGISTICS would take major responsibility for evaluating the test results. Tests on unirradiated fuel rod samples involved both dynamic and quasi-static loading on fuel samples. PWR fuel rods loaded with uranium pellets were dropped vertically from 9m onto a rigid target and this was repeated on BWR fuel rods, similar tests on empty fuel rods were also conducted. Quasi-static tests were conducted on 530 mm long PWR and BWR fuel specimens under axial loading. Tests on irradiated fuel samples were conducted on high burn-up fuel rods of both PWR and BWR types. These were believed original to the FIP project and involved applying bending loads to simply supported pressurised rod specimens. In one test the fuel rod was heated to nearly 500oC during loading, all specimens were subject to axial impact before testing. Considerable experience of fuel rod testing and new data was gained from this test programme

Control rod

International Nuclear Information System (INIS)

Takeda, Toshikazu; Inoue, Kotaro.

1979-01-01

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

Control rod

International Nuclear Information System (INIS)

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

1994-01-01

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

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

International Nuclear Information System (INIS)

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

1977-01-01

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

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

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.)

Photonic crystal rod fibers: Understanding a new class of active optical waveguides

DEFF Research Database (Denmark)

Laurila, Marko

core modes and the cladding band as the underlying mechanics to ensure SM operation of the new rod fiber design (85 μm core diameter), which was developed during this thesis work. The second focus of this work is the study of the new ytterbiumdoped rod fiber design under active operation. Performance...... of the rod fiber is evaluated in high power laser and laser amplifier configurations. The high power rod amplifier setup including the seed source is developed and characterized. Results obtained from the rod fiber showed simultaneously SM, near diffraction limited output beam quality with high average power...... and pulse energy generation using both laser and laser amplifier configurations. Modal instabilities (MIs) in high power fiber amplifiers are discussed, and a memory effect of the MI threshold level together with a recovery method and evidence of improved performance while suppressing MIs are reported...

Le mélanome malin: une tumeur rare des fosses nasales - à propos d'une série de 10 cas

Science.gov (United States)

Errachdi, Amal; Epala, Brice Nkoua; Asabbane, Amal; Kabbali, Naoual; Hemmich, Mariem; Kebdani, Tayeb; Benjaafar, Noureddine

2014-01-01

Le mélanome malin des fosses nasales est une tumeur rare mais très agressive, de traitement complexe et de pronostic défavorable. Son traitement relève en principe d'une prise en charge essentiellement chirurgicale complétée par une radiothérapie. L'objectif de ce travail est de rapporter les caractéristiques cliniques, thérapeutiques et évolutives des mélanomes des fosses nasales. Nous avons analysé rétrospectivement 10 cas de mélanomes des fosses nasales suivis à l'institut national d'oncologie de Rabat. La rhinoscopie avec biopsie a permis la confirmation histologique du diagnostic de mélanome. Le bilan d'extension comprenait une tomodensitométrie ou imagerie par résonnance magnétique du massif facial, une radiographie thoracique et une échographie abdominale. Dans notre série, l’âge médian était de 67.5 ans, avec une prédominance féminine (7femmes et 3hommes). Le délai médian de découverte était de 6 mois. Deux patients étaient métastatiques d'emblée, et toutes les tumeurs étaient localement avancées au moment du diagnostic. Sept patients ont été opérés avec des limites chirurgicales envahies dans 2 cas et 3 patients étaient inopérables. 2 patients ont été irradiés après la chirurgie et 2 patients ont reçu une chimiothérapie arrêtée au moment de la progression. Deux patients ont récidivé après traitement, et un patient était en mauvais état général et a bénéficié uniquement de soins palliatifs. Tous les patients sont décédés avec un délai médian de survie de 12 mois. Le mélanome malin muqueux des fosses nasales, bien que rare, demeure une pathologie de pronostic défavorable et pose des problèmes de prise en charge. PMID:25404963

Compact, low-loss and broadband photonic crystal circulator based on a star-type ferrite rod

Directory of Open Access Journals (Sweden)

Xiang Xi

Full Text Available We propose and investigate a compact, low-loss and broadband circulator based on a star-type ferrite rod in two-dimensional square-lattice photonic crystals. Only one ferrite rod is required to be inserted in our structure. Firstly, the performances of circulator based on the star-type, circle, and square ferrite rod are compared, showing that the circulator with the star-type ferrite rod performs better than the other two ones. And then, based on the star-type ferrite rod circulator, four cases of improvement, in which the background rods around the center ferrite rod are replaced respectively by the backward-triangle, forward-triangle, backward-semicircle, and forward-semicircle rods, are investigated to modulate the coupling between the center magneto-optical micro-cavity and the corresponding waveguides. The results show that, with proper parameters, all the four cases can greatly improve the output properties of the circulator, and different cases have its own advantages. The mechanism behind these improvements is also discussed. Finite-element method is used to calculate the characteristics of the circulator and Nelder-Mead optimization method is employed to obtain the optimized parameters. The ideas presented here are useful for designing broadband, low insertion loss, and high-isolation circulators which have potential application in integrated photonic crystal devices. Keywords: Photonic crystals, Circulator, Magneto-optical material, Photonic crystal waveguides

The Third ATLAS ROD Workshop

CERN Multimedia

Poggioli, L.

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

Temperature actuated automatic safety rod release

Science.gov (United States)

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

1984-03-13

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

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

International Nuclear Information System (INIS)

Nasimi, E.; Gabbar, H.A.

2011-01-01

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

Shielding device for control rod in nuclear reactor

International Nuclear Information System (INIS)

Sakamaki, Kazuo; Tomatsu, Tsutomu.

1995-01-01

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

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.)

Urea controlled hydrothermal synthesis of ammonium aluminum carbonate hydroxide rods

Science.gov (United States)

Wang, Fang; Zhu, Jianfeng; Liu, Hui

2018-03-01

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

Control rod for HTGR type reactor

International Nuclear Information System (INIS)

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

1990-01-01

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

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.

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.

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

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

International Nuclear Information System (INIS)

Tuerk, W.

1980-01-01

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

Spacers for fuel rod clusters

International Nuclear Information System (INIS)

Jabsen, F.S.