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Sample records for agincd control rod

  1. Behavior of AgInCd absorber material in Zry/UO{sub 2} fuel rod simulator bundles tested at high temperatures in the CORA facility

    Energy Technology Data Exchange (ETDEWEB)

    Sepold, L.; Hagen, S.; Hofmann, P.; Schanz, G.

    2009-01-15

    The CORA experiments carried out in an out-of-pile facility at the Kernforschungszentrum Karlsruhe (KfK), Federal Republic of Germany, are part of the ''Severe Fuel Damage'' (SFD) program. The experimental program is to provide information on the failure mechanisms of Light Water Reactor (LWR) fuel elements in a temperature range from 1200 C to 2000 C and in a few cases up to 2400 C. In the CORA experiments two different bundle configurations are tested: PWR (Pressurized Water Reactor) and BWR (Boiling Water Reactor) bundles. The PWR-type assemblies usually consist of 25 rods with 16 electrically heated fuel rod simulators and nine unheated rods (full-pellet and absorber rods). Bundle CORA-5 contained one Ag/In/Cd - steel absorber rod whereas two absorber rods were used in CORA-12, CORA-15, and CORA-9. The larger bundle CORA-7 contained 5 absorber rods. CORA-12 was terminated by quenching with water from the bottom. In CORA-15 the heated and unheated rods were pressurized to achieve pronounced clad ballooning. Bundle CORA-9 was tested with a system pressure of 1.0 MPa instead of 0.22 MPa. The test bundles were subjected to temperature transients of a slow heatup rate in a steam environment. Thus, an accident sequence is simulated, which may develop from a small-break loss-of-coolant accident of a LWR. The transient phases of the tests were initiated with a temperature ramp rate of 1 K/s. The temperature escalation due to the exothermal zircaloy (Zry)-steam reaction started at about 1100 C, leading the bundles to maximum temperatures of approximately 2000 C. Rod destruction started with the failure of the absorber rod cladding at about 1200 C, i.e. about 250 K below the melting regime of steel. Penetration of the steel cladding was presumably caused by a eutectic interaction between steel and the zircaloy guide tube. The test bundles resulted in severe oxidation and partial melting of the cladding, fuel dissolution by Zry/UO{sub 2} interaction

  2. CONTROL ROD

    Science.gov (United States)

    Zinn, W.H.; Ross, H.V.

    1958-11-18

    A control rod is described for a nuclear reactor. In certaln reactor designs it becomes desirable to use a control rod having great width but relatively llttle thickness. This patent is addressed to such a need. The neutron absorbing material is inserted in a triangular tube, leaving volds between the circular insert and the corners of the triangular tube. The material is positioned within the tube by the use of dummy spacers to achleve the desired absorption pattern, then the ends of the tubes are sealed with suitable plugs. The tubes may be welded or soldered together to form two flat surfaces of any desired width, and covered with sheetmetal to protect the tubes from damage. This design provides a control member that will not distort under the action of outside forces or be ruptured by gases generated within the jacketed control member.

  3. Radiological characterization of spent control rod assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Lepel, E.A.; Robertson, D.E.; Thomas, C.W.; Pratt, S.L.; Haggard, D.L. [Pacific Northwest Lab., Richland, WA (United States)

    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), {sup 60}Co and {sup 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 {sup 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.

  4. Control rod

    International Nuclear Information System (INIS)

    Purpose: To enable semi-permanent and safety use of a control rod in a water cooled type reactor operated under high temperature and high pressure conditions by using a blade in which hafnium materials at a nuclear reactor quality are covered with stainless steels or zirconium alloys. Constitution: A plate-like hafnium material is surrounded with a thin plate of stainless steels or zirconium alloys under vacuum and the joint portions of the thin plate is subjected to seam welding. Then a blade is prepared by welding the remaining joining portions at both ends in a conventional manner. The welding method usable herein includes electron beam welding, laser welding and the like. If it is required to increase the close bondability between the halfnium plate and the thin plate, the blade thus obtained is subjected as it is to extrusion fabrication thereby obtain a desired increased bondability. (Kawakami, Y.)

  5. Reactor control rod

    International Nuclear Information System (INIS)

    Object: To enable quick descent of a control rod body even when some relative phase deviation between upper drive means and wrapper tube is produced, while permitting a coolant to effectively flow into a protective tube irrespective of the position of the control rod body. Structure: In a control rod used for a nuclear reactor such as a fast breeder, an orifice which dispenses with a cylindrical guide tube and has a greater inner diameter than the outer diameter of the protective tube of the control rod body is provided on the inner side of a wrapper tube, thus permitting smooth operation of the control rod body and also permitting the coolant to effectively flow into the protective tube irrespective of the control rod body. (Horiuchi, T.)

  6. Control rod drives

    International Nuclear Information System (INIS)

    Purpose: To rapidly detect the position to which a control rod has been rapidly inserted into the reactor core upon scram in the control rod drives for use in LMFBR type reactors. Constitution: In control rod drives comprising an acceleration spring disposed to the outside of an extension pipe and an acceleration pipe for conducting the spring force to a control rod for rapidly dropping the rod into the reactor core, a magnet having a repulsive force is disposed to each acceleration pipe and guide pipe as decelerating and buffering means for the acceleration pipe. The position of the control rod is detected by the interaction between the magnet and the coils attached to the inside of the guide pipe or reactor lead switch. According to this invention, 85 % scram signal which has hitherto been difficult to be processed electrically can be obtained with a sufficient intensity and with no delay to thereby improve the entire safety of the reactor system. Then, the inserted position and the insertion time can accurately and rapidly be detected. (Horiuchi, T.)

  7. Control rod drive system

    International Nuclear Information System (INIS)

    The present invention concerns an electromotive driving-type control rod driving system of a BWR type reactor, for which sliding resistance (friction) test can be performed of a movable portion of the control rod driving mechanisms. Namely, a hydraulic pressure control unit has following constitutions in addition to a conventional constitution as a sliding resistance test performing function. (1) A restricting valve is disposed downstream of the scram valve of scram pipelines to control flow rate and pressure of pressurized water flown in the pipelines. (2) A pressure gauge detects a pressure between the scram valve and the restricting valve. (3) A flow meter detects the flow rate of pipelines controlled by the restricting valve. (4) A recording pressure detector detects the pressure at the downstream of the restricting valve. (5) The recording device is attached when the sliding resistant test is performed for tracing the pressure measured by the pressure detection device. Further, the scram valve sends electric signals to a central operation chamber when it is fully closed. The central operation chamber has a function of fully opening the restricting valve by way of the electric signals. (I.S.)

  8. Microstructure Change of Ag-In-Cd Alloy after Irradiation in Reactor%Ag-In-Cd合金辐照后的微观组织变化

    Institute of Scientific and Technical Information of China (English)

    龙冲生; 肖红星; 高雯; 陈洪生

    2015-01-01

    Ag-In-Cd control rods are widely used in PWR nuclear power plants.The irradiation swelling behavior of Ag-In-Cd alloy is very important to the safety assess-ment of control rod during its operation.In this work,to simulate the change of micro-structure and density of Ag-In-Cd alloy after irradiation in reactor,a series of simulation alloys were prepared,and the effect of composition on the microstructure and density was investigated.A formula to calculate the alloy density with different compositions was obtained by fitting the experimental values.It is found that simulation alloy will consist of two phases,an fcc phase and an hcp phase,when the content of Ag was 77.5% (mass fraction).In the fcc phase,Ag content will be higher than its average content,and there is a little amount of Sn.In the hcp phase,Ag content will be below its average content,and Sn content will be relatively high.After irradiation,Ag-In-Cd alloy will be single hcp phase when Ag content is between 5 5% and 6 1%.%Ag-In-Cd合金在核电站压水堆控制棒中广泛使用,其辐照肿胀行为是评价Ag-In-Cd控制棒使用寿命的关键因素。本文通过制备不同成分的模拟合金,来模拟Ag-In-Cd合金在堆内辐照后的成分变化,分析合金的密度及微观组织特点。结果发现,当Ag含量低至77.5%(质量分数)时,合金会分解为fcc和hcp两相,fcc相中贫Sn高Ag,hcp相中富Sn低Ag。当Ag含量在55%~61%之间时,合金以hcp单相存在。由实测的密度拟合出了合金密度随成分变化的关系式。此结果对于理解和掌握Ag-In-Cd合金的辐照肿胀行为有重要意义。

  9. Control rod drives

    International Nuclear Information System (INIS)

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

  10. Blackness coefficients, effective diffusion parameters, and control rod worths for thermal reactors

    International Nuclear Information System (INIS)

    Simple diffusion theory cannot be used to evaluate control rod worths in thermal reactors because of the strongly absorbing character of the control material. However, good results can be obtained from a diffusion calculation by representing the absorber slab by means of a suitable pair of internal boundary conditions, α and β, which are ratios of neutron flux to neutron current. Methods for calculating α and β in the P1, P3, and P5 approximations, with and without scattering, are presented. By appropriately weighting the fine-group blackness coefficients, broad group values, and , are obtained. The technique is applied to the calculation of control rod worths of Cd, Ag-In-Cd, and Hf control elements. Results are found to compare very favorably with detailed Monte Carlo calculations. 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 is briefly discussed and applied to the calculation of control rod worths in the Ford Nuclear Reactor at the University of Michigan. Calculated and measured worths are found to be in good agreement

  11. Blackness coefficients, effective diffusion parameters, and control rod worths for thermal reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bretscher, M.M.

    1984-09-01

    Simple diffusion theory cannot be used to evaluate control rod worths in thermal reactors because of the strongly absorbing character of the control material. However, good results can be obtained from a diffusion calculation by representing the absorber slab by means of a suitable pair of internal boundary conditions, ..cap alpha.. and ..beta.., which are ratios of neutron flux to neutron current. Methods for calculating ..cap alpha.. and ..beta.. in the P/sub 1/, P/sub 3/, and P/sub 5/ approximations, with and without scattering, are presented. By appropriately weighting the fine-group blackness coefficients, broad group values, <..cap alpha..> and <..beta..>, are obtained. The technique is applied to the calculation of control rod worths of Cd, Ag-In-Cd, and Hf control elements. Results are found to compare very favorably with detailed Monte Carlo calculations. 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 is briefly discussed and applied to the calculation of control rod worths in the Ford Nuclear Reactor at the University of Michigan. Calculated and measured worths are found to be in good agreement.

  12. Control rod drive

    International Nuclear Information System (INIS)

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

  13. Process and apparatus for controlling control rods

    International Nuclear Information System (INIS)

    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

  14. Advanced gray rod control assembly

    Science.gov (United States)

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

    2013-09-17

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

  15. Reactor control rod timing system. [LMFBR

    Science.gov (United States)

    Wu, P.T.K.

    1980-03-18

    A fluid driven jet-edge whistle timing system is described for control rods of a nuclear reactor for producing real-time detection of the timing of each control rod in its scram operation. An important parameter in reactor safety, particularly for liquid metal fast breeder reactors (LMFBR), is the time deviation between the time the control rod is released and the time the rod actually reaches the down position. The whistle has a nearly pure tone signal with center frequency (above 100 kHz) far above the frequency band in which the energy of the background noise is concentrated. Each control rod can be fitted with a whistle with a different frequency so that there is no ambiguity in differentiating the signal from each control rod.

  16. Control Rod Malfunction at the NRAD Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Thomas L. Maddock

    2010-05-01

    The neutron Radiography Reactor (NRAD) is a training, research, and isotope (TRIGA) reactor located at the INL. The reactor is normally shut down by the insertion of three control rods that drop into the core when power is removed from electromagnets. During a routine shutdown, indicator lights on the console showed that one of the control rods was not inserted. It was initially thought that the indicator lights were in error because of a limit switch that was out of adjustment. Through further testing, it was determined that the control rod did not drop when the scram switch was initially pressed. The control rod anomaly led to a six month shutdown of the reactor and an in depth investigation of the reactor protective system. The investigation looked into: scram switch operation, console modifications, and control rod drive mechanisms. A number of latent issues were discovered and corrected during the investigation. The cause of the control rod malfunction was found to be a buildup of corrosion in the control rod drive mechanism. The investigation resulted in modifications to equipment, changes to both operation and maintenance procedures, and additional training. No reoccurrences of the problem have been observed since corrective actions were implemented.

  17. Control rod housing alignment and repair method

    International Nuclear Information System (INIS)

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

  18. Apparatus for handling control rod drives

    International Nuclear Information System (INIS)

    An apparatus for handling control rod drives (CRD's) attached by detachable fixing means to housings mounted in a reactor pressure vessel and each coupled to one of control rods inserted in the reactor pressure vessel is described. The apparatus for handling the CRD's comprise cylindrical housing means, uncoupling means mounted in the housing means for uncoupling each of the control rods from the respective CRD, means mounted on the housing means for effecting attaching and detaching of the fixing means, means for supporting the housing means, and means for moving the support means longitudinally of the CRD

  19. Control rod for a nuclear reactor

    International Nuclear Information System (INIS)

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

  20. Thermal and stress analysis of control rod

    International Nuclear Information System (INIS)

    In order to survey the mechanical integrity of a control rod in the high temperature core of the VHTR, thermal analysis and thermal stress analysis were carried out by means of calculus of finite differentials and finite element methods for the plant under the normal operating condition as well as under several abnormal conditions. The results of the analyses have been applied to refine the mechanical design of the control rod

  1. Magnetic switch for reactor control rod. [LMFBR

    Science.gov (United States)

    Germer, J.H.

    1982-09-30

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

  2. Regulatory perspective on incomplete control rod insertions

    Energy Technology Data Exchange (ETDEWEB)

    Chatterton, M.

    1997-01-01

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

  3. Study of Fluidized-Bed Control Rods

    International Nuclear Information System (INIS)

    Control of nuclear reactors with fluidized-bed control rods (FBCR) has been previously proposed; but, despite some apparent advantages over electromechanical systems, such rods have not received widespread attention. With the FBCR concept, the reactor control system becomes a flow-regulating system using either variable-speed pumps or motor-driven control valves in the main coolant. Alternatively, in-core by-pass piping similar to control systems now being developed for fluidized-bed reactors may be utilized. Some of the possible advantages of the FBCR concept are as follows: (1) Most pressure-vessel head penetrations are eliminated, and refueling is simplified; (2) Automatic scram results from a loss-of-flow accident; (3) Axial power can be shaped by the use of contoured channels or variable-sized particles; (4). Water-gap flux peaking can be reduced for the partially withdrawn control rod; (5) The temperature reactivity allowance may be reduced if the fluidized control rods have a negative temperature coefficient; and (6) Fabrication costs are much lower than for electromechanical systems. An evaluation of the FBCR concept, including construction of prototype models and testing of the hydraulic and nuclear characteristics, has been performed. Two types of rods were studied: transmission rods (thickness ≦ 2 mean-free-paths) and reflection rods (thickness ≦ 4 mean-free-paths). Acceptable hydraulic and nuclear characteristics are possible with both types. The feasibility of controlling low-power reactors by either transmission- or reflection-type fluidized.-bed control rods has been established. Furthermore, it was shown that the FBCR concept has good control properties which may be calculated by standard theoretical methods. For high-power, high-temperature applications, additional information on particle material characteristics is needed. A great advantage offered by the FBCR is the possibility of shaping the axial flux either by the use of particles of

  4. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

    Magnusson, Per

    2008-01-15

    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 {sup 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% {sup 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 {sup 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.

  6. Method for controlling xenone control rod in nuclear reactors

    International Nuclear Information System (INIS)

    Purpose: To adequately conduct entire withdrawal or entire insertion of xenone control rods by comparing the control rod worths of the xenone control rods and the xenone amount. Method: A turbine output signal is inputted to a load variation detection circuit and, if the variation coefficient is with a predetermined setting value, it is judged that the load variation has been completed and the load is settled constant, the result of which is inputted to the control rod selection device. The reactor power signal is inputted to a control rod selection device and the number density of iodine and the number density of xenone are determined based on the neutron flux and the maximum or minimum value of the xenone at a constant load is calculated based on both of the data. Then, a function representing the variation amount of the xenone reactivity having the maximum or minimum value as the variant is determined. By comparing the function and the constant load signal, the operation for the xenone control rods is judged or selected. According to this invention, the conventional method of compensating the xenone amount variation with the adjustment of the boron concentration can be substituted with the xenone control rods. (Kawakami, Y.)

  7. Analysis of buffering process of control rod hydraulic absorber

    International Nuclear Information System (INIS)

    Control Rod Hydraulic Drive Mechanism(CRHDM) is a newly invented build-in control rod drive mechanism. Hydraulic absorber is the key part of this mechanism, and is used to cushion the control rod when the rod scrams. Thus, it prevents the control rod from being deformed and damaged. In this paper dynamics program ANSYS CFX is used to calculate all kinds of flow conditions in hydraulic absorber to obtain its hydraulic characteristics. Based on the flow resistance coefficients obtained from the simulation results, fluid mass and momentum equations were developed to get the trend of pressure change in the hydraulic cylinder and the displacement of the piston rod during the buffering process of the control rod. The results obtained in this paper indicate that the hydraulic absorber meets the design requirement. The work in this paper will be helpful for the design and optimization of the control rod hydraulic absorber. (author)

  8. Control rod for PWR type reactor

    International Nuclear Information System (INIS)

    Since a silver-indium-cadmium alloy has been used as the absorber for control rods, swelling due to neutron absorption has been caused. On the other hand, a stainless steel cladding tube for the absorber gradually reduces its outer diameter by the pressure of reactor coolants and neutron irradiation and causes contact during working life to often bring about cracking in the cladding tube. Then, the control rod is divided into two independent portions and joined by an intermediate end plug into a single rod, in which the upper portion is made free from pressure and the lower portion is pressurized. Further, a large gap is formed between the lower absorber and the lower cladding tube. Further, chromium or chromium carbide is coated to the outer surface of the upper cladding tube for improving the abrasion resistance. Thus, the cladding tube is made abrasion resistant and it is possible to prevent cracking in the cladding tube due to interaction between the tube and the absorber, inner presurization at the lower portion, reduced diameter for the absorber and the gap of the tube. (N.H.)

  9. Scram release for a control rod

    International Nuclear Information System (INIS)

    A double-sided rack and pinion control element drive mechanism for a nuclear reactor is disclosed wherein scram release is accomplished by the dual action of withdrawing the pinion from its engagement with the rack while at the same time rotating the pinion in a direction consistent with the movement of the rack and the control rod in their downward travel. The pinion is withdrawn from engagement with the rack while remaining in engagement with its stationary driving means. The continuing engagement with the stationary driving means causes the pinion to rotate when the pinion is caused to move away from the rack during disengagement. 10 claims, 7 drawing figures

  10. Rolls-Royce digital Rod Control System

    International Nuclear Information System (INIS)

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

  11. Rolls-Royce digital Rod Control System

    Energy Technology Data Exchange (ETDEWEB)

    Pouillot, M. [Rolls-Royce Civil Nuclear SAS (France)

    2010-07-01

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

  12. Minimization of PWR reactor control rods wear

    International Nuclear Information System (INIS)

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

  13. Dependence of control rod worth on fuel burnup

    Energy Technology Data Exchange (ETDEWEB)

    Savva, P., E-mail: savvapan@ipta.demokritos.g [NCSR ' DEMOKRITOS' , PoB 60228, 15310 Aghia Paraskevi (Greece); Varvayanni, M., E-mail: melina@ipta.demokritos.g [NCSR ' DEMOKRITOS' , PoB 60228, 15310 Aghia Paraskevi (Greece); Catsaros, N., E-mail: nicos@ipta.demokritos.g [NCSR ' DEMOKRITOS' , PoB 60228, 15310 Aghia Paraskevi (Greece)

    2011-02-15

    Research highlights: Diffusion and MC calculations for rod worth dependence on burnup and Xe in reactors. One-step rod withdrawal/insertion are used for rod worth estimation. The study showed that when Xe is present the rods worth is significantly reduced. Rod worth variation with burnup depends on rod position in core. Rod worth obtained with MC code is higher than that obtained from deterministic. - Abstract: One important parameter in the design and the analysis of a nuclear reactor core is the reactivity worth of the control rods, i.e. their efficiency to absorb excess reactivity. The control rod worth is affected by parameters such as the fuel burnup in the rod vicinity, the Xe concentration in the core, the operational time of the rod and its position in the core. In the present work, two different computational approaches, a deterministic and a stochastic one, were used for the determination of the rods worth dependence on the fuel burnup level and the Xe concentration level in a conceptual, symmetric reactor core, based on the MTR fuel assemblies used in the Greek Research Reactor (GRR-1). For the deterministic approach the neutronics code system composed by the SCALE modules NITAWL and XSDRN and the diffusion code CITATION was used, while for the stochastic one the Monte Carlo code TRIPOLI was applied. The study showed that when Xe is present in the core, the rods worth is significantly reduced, while the rod worth variation with increasing burnup depends on the rods position in the core grid. The rod worth obtained with the use of the Monte Carlo code is higher than the one obtained from the deterministic code.

  14. Controlling a nuclear reactor with dropped control rods

    International Nuclear Information System (INIS)

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

  15. The analytic method for calculating the control rod worth

    International Nuclear Information System (INIS)

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

  16. Optimized Control Rods of the BR2 Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kalcheva, Silva; Koonen, E.

    2007-09-15

    At the present time the BR-2 reactor uses control elements with cadmium as neutron absorbing part. The lower section of the control element is a beryllium assembly cooled by light water. Due to the burn up of the lower end of the cadmium section during the reactor operation, the presently used rods for reactivity control of the BR-2 reactor have to be replaced by new ones. Considered are various types Control Rods with full active part of the following materials: cadmium (Cd), hafnium (Hf), europium oxide (Eu2O3) and gadolinium (Gd2O3). Options to decrease the burn up of the control rod material in the hot spot, such as use of stainless steel in the lower active part of the Control Rod are discussed. Comparison with the characteristics of the presently used Control Rods types is performed. The changing of the characteristics of different types Control Rods and the perturbation effects on the reactor neutronics during the BR-2 fuel cycle are investigated. The burn up of the Control Rod absorbing material, total and differential control rods worth, macroscopic and effective microscopic absorption cross sections, fuel and reactivity evolution are evaluated during approximately 30 operating cycles.

  17. Reactivity control rod for controlling reactor power distribution

    International Nuclear Information System (INIS)

    Since a cladding tube is situated at the outer side, it undergoes neutron irradiation in a reactor core and also undergoes compression force due to high pressure of reactor coolants to cause a creep phenomenon, and the diameter is reduced as it is used. Then, neutron absorbing rods as reactivity control rods for controlling the power distribution are constituted with a cladding tube, a spacer tube disposed at the central portion of the cladding tube and a borosilicate glass tube disposed between the cladding tube and the spacer tube. The gap between the borosilicate glass tube and the spacer tube is gradually changed so that the inner diameter of the borosilicate glass is increased as it comes closer to the lower end plug. The time of contact between the cladding tube and the spacer tube in the inside is delayed by the constitution of the borosilicate glass tube disposed in the cladding tube of the neutron absorbing rod as the reactivity control rod thereby capable of extending the integral working life time with no rupture of the cladding tube. (N.H.)

  18. Material operating behaviour of ABB BWR control rods

    International Nuclear Information System (INIS)

    The BWR control rods made by ABB use boron carbide (B4C and hafnium as absorber material within a cladding of stainless steel. The general behaviour under operation has proven to be very good. ABB and many of their control rod customers have performed extensive inspection programs of control rod behaviour. However, due to changes in the material properties under fast and thermal neutron irradiation defects may occur in the control rods at high neutron fluences. Examinations of irradiated control rod materials have been performed in hot cell laboratories. The examinations have revealed the defect mechanism Irradiation Assisted Stress Corrosion Cracking (IASCC) to appear in the stainless steel cladding. For IASCC to occur three factors have to act simultaneously. Stress, material sensitization and an oxidising environment. Stress may be obtained from boron carbide swelling due to irradiation. Stainless steel may be sensitized to intergranular stress corrosion cracking under irradiation. Normally the reactor environment in a BWR is oxidising. The presentation focuses on findings from hot cell laboratory work on irradiated ABB BWR control rods and studies of irradiated control rod materials in the hot cells at PSI. Apart from physical, mechanical and microstructural examinations, isotope analyses were performed to describe the local isotopic burnup of boron. Consequences (such as possible B4C washout) of a under operation in a ABB BWR, after the occurrence of a crack is discussed based on neutron radiographic examinations of control rods operated with cracks. (author)

  19. Control rod drive for high temperature gas cooled reactor

    Institute of Scientific and Technical Information of China (English)

    DengJun-Xian; XuJi-Ming; 等

    1998-01-01

    This control rod drive is developed for HTR-10 high temperature gas cooled test reactor.The stepmotor is prefered to improve positioning of the control rod and the scram behavior.The preliminary test in 1600170 ambient temperature shows that the selected stepmotor and transmission system can meet the main operation function requirements of HTR-10.

  20. End plug welding method for control rod

    International Nuclear Information System (INIS)

    A cladding tube of a control rod has a coating layer of plated hard chromium or the like on the outer surface of the tube main body. The outer edge of an end plug to be attached to the end portion of the cladding tube has a tapered face opposing to the end portion of the cladding tube. The end plug is inserted under pressure to the end portion of the cladding tube in a state where neutron absorbers are contained and a coil spring is inserted in the cladding tube. Electric current is applied between the tube main body and the end plug in this state. The tube main body and the end plug are heated by their intrinsic resistance and contact resistance up to a weldable temperature. The tube main body and the end plug are joined by an urging pressure applied between the tube main body and the end plug. Since the end plug is welded to the end portion of the cladding tube at the circumference thereof by resistance welding, there is no worry of intruding the coating material to the welding portion, thereby enabling to attain satisfactory welding. (I.N.)

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

    International Nuclear Information System (INIS)

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

  2. Fatigue Life Improving of Drill Rod by Inclusion Control

    Science.gov (United States)

    Wang, Linzhu; Yang, Shufeng; Li, Jingshe; Liu, Wei; Zhou, Yinghao

    2016-08-01

    Large and hard inclusions often deteriorate the service performance and reduce the fatigue lifetime of drill rods. In this paper, the main reasons of the rupture of drill rods were analyzed by the examination of their fracture and it is found that the large inclusions were the main reason of breakage of rod drill. The inclusions were high of Ca content or Al2O3 rich. Smaller and better deformability inclusions were obtained by the optimization of refining slag, calcium treatment process and the flow control devices of tundish. Results of industrial experiment after optimization show that total oxygen content of drill rods decreased by more than 50%, macro-inclusions weight fraction decreased from about 4 mg/10 kg to about 0.3 mg/10 kg and the micro-inclusions average size decreased from 6 to 3.6 μm. The average using times of drill rods after optimization were increased by about 60%.

  3. Design requirement on KALIMER control rod assembly duct

    International Nuclear Information System (INIS)

    This document establishes the design guidelines which are needs for designing the control rod assembly duct of the KALIMER as design requirements. it describes control rod assembly duct of the KALIMER and its requirements that includes functional requirements, performance requirements, interfacing systems, design limits and strength requirements, seismic requirements, structural requirements, environmental requirements, reliability and safety requirements, standard and codes, QA programs, and other requirements. The control rod system consists of three parts, which are drive mechanism, drive-line, and absorber bundle. This report deals with the absorber bundle and its outer duct only because the others are beyond the scope of fuel system design. The guidelines for design requirements intend to be used for an improved design of the control rod assembly duct of the KALIMER. (author). 19 refs

  4. Design requirement on KALIMER control rod assembly duct

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, W.; Kang, H. Y.; Nam, C.; Kim, J. O.; Kim, Y. J

    1998-03-01

    This document establishes the design guidelines which are needs for designing the control rod assembly duct of the KALIMER as design requirements. it describes control rod assembly duct of the KALIMER and its requirements that includes functional requirements, performance requirements, interfacing systems, design limits and strength requirements, seismic requirements, structural requirements, environmental requirements, reliability and safety requirements, standard and codes, QA programs, and other requirements. The control rod system consists of three parts, which are drive mechanism, drive-line, and absorber bundle. This report deals with the absorber bundle and its outer duct only because the others are beyond the scope of fuel system design. The guidelines for design requirements intend to be used for an improved design of the control rod assembly duct of the KALIMER. (author). 19 refs.

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

    International Nuclear Information System (INIS)

    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

  6. Development of aluminum (Al5083)-clad ternary Ag-In-Cd alloy for JSNS decoupled moderator

    Energy Technology Data Exchange (ETDEWEB)

    Teshigawara, M. [Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki 319-1195 (Japan)]. E-mail: teshigawara.makoto@jaea.go.jp; Harada, M. [Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki 319-1195 (Japan); Saito, S. [Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki 319-1195 (Japan); Oikawa, K. [Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki 319-1195 (Japan); Maekawa, F. [Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki 319-1195 (Japan); Futakawa, M. [Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki 319-1195 (Japan); Kikuchi, K. [Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki 319-1195 (Japan); Kato, T. [Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki 319-1195 (Japan); Ikeda, Y. [Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki 319-1195 (Japan); Naoe, T. [Ibaraki University, 4-12-1 Nakanarusawa-cho, Hitachi, Ibaraki 316-8511 (Japan); Koyama, T. [Ibaraki University, 4-12-1 Nakanarusawa-cho, Hitachi, Ibaraki 316-8511 (Japan); Ooi, T. [Ibaraki University, 4-12-1 Nakanarusawa-cho, Hitachi, Ibaraki 316-8511 (Japan); Zherebtsov, S. [Ibaraki University, 4-12-1 Nakanarusawa-cho, Hitachi, Ibaraki 316-8511 (Japan); Kawai, M. [High Energy Accelerator Research Organization, 1-1, Oho, Tsukuba-shi, Ibaraki 305-0801 (Japan); Kurishita, H. [International Research Center for Nuclear Materials Science, Institute for Materials Research (IMR), Tohoku University, Narita-machi, Oarai-machi, Higashi ibaraki-gun, Ibaraki 311-1313 (Japan); Konashi, K. [International Research Center for Nuclear Materials Science, Institute for Materials Research (IMR), Tohoku University, Narita-machi, Oarai-machi, Higashi ibaraki-gun, Ibaraki 311-1313 (Japan)

    2006-09-15

    To develop Ag (silver)-In (indium)-Cd (cadmium) alloy decoupler, a method is needed to bond the decoupler between Al alloy (Al5083) and the ternary Ag-In-Cd alloy. We found that a better HIP condition was temperature, pressure and holding time at 803 K, 100 MPa and 10 min. for small test pieces ({phi}22 mm in dia. x 6 mm in height). Hardened layer due to the formation of AlAg{sub 2} was found in the bonding layer, however, the rupture strength of the bonding layer is more than 30 MPa, the calculated design stress. Bonding tests of a large size piece (200 x 200 x 30 mm{sup 3}), which simulated the real scale, were also performed according to the results of small size tests. The result also gave good bonding and enough required-mechanical-strength.

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

    International Nuclear Information System (INIS)

    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

  8. Dysprosium titanate as an absorber material for control rods

    Energy Technology Data Exchange (ETDEWEB)

    Risovany, V.D. E-mail: fae@niiar.ru; Varlashova, E.E.; Suslov, D.N

    2000-09-02

    Disprosium titanate is an attractive control rod material for the thermal neutron reactors. Its main advantages are: insignificant swelling, no out-gassing under neutron irradiation, rather high neutron efficiency, a high melting point ({approx}1870 deg. C), non-interaction with the cladding at temperatures above 1000 deg. C, simple fabrication and easily reprocessed non-radioactive waste. It can be used in control rods as pellets and powder. The disprosium titanate control rods have worked off in the MIR reactor for 17 years, in VVER-1000 - for 4 years without any operating problems. After post-irradiation examinations this type of control rod having high lifetime was recommended for the VVER and RBMK. The paper presents the examination results of absorber element dummies containing dysprosium titanate, irradiated in the SM reactor to the neutron fluence of 3.4x10{sup 22} cm{sup -2} (E>0.1 MeV) and, also, the data on structure, thermal-physical properties of dysprosium titanate, efficiency of dysprosium titanate control rods.

  9. Dysprosium titanate as an absorber material for control rods

    Science.gov (United States)

    Risovany, V. D.; Varlashova, E. E.; Suslov, D. N.

    2000-09-01

    Disprosium titanate is an attractive control rod material for the thermal neutron reactors. Its main advantages are: insignificant swelling, no out-gassing under neutron irradiation, rather high neutron efficiency, a high melting point (˜1870°C), non-interaction with the cladding at temperatures above 1000°C, simple fabrication and easily reprocessed non-radioactive waste. It can be used in control rods as pellets and powder. The disprosium titanate control rods have worked off in the MIR reactor for 17 years, in VVER-1000 - for 4 years without any operating problems. After post-irradiation examinations this type of control rod having high lifetime was recommended for the VVER and RBMK. The paper presents the examination results of absorber element dummies containing dysprosium titanate, irradiated in the SM reactor to the neutron fluence of 3.4×10 22 cm -2 ( E>0.1 MeV) and, also, the data on structure, thermal-physical properties of dysprosium titanate, efficiency of dysprosium titanate control rods.

  10. Control rod worth calculations using deterministic and stochastic methods

    Energy Technology Data Exchange (ETDEWEB)

    Varvayanni, M. [NCSR ' DEMOKRITOS' , PO Box 60228, 15310 Aghia Paraskevi (Greece); Savva, P., E-mail: melina@ipta.demokritos.g [NCSR ' DEMOKRITOS' , PO Box 60228, 15310 Aghia Paraskevi (Greece); Catsaros, N. [NCSR ' DEMOKRITOS' , PO Box 60228, 15310 Aghia Paraskevi (Greece)

    2009-11-15

    Knowledge of the efficiency of a control rod to absorb excess reactivity in a nuclear reactor, i.e. knowledge of its reactivity worth, is very important from many points of view. These include the analysis and the assessment of the shutdown margin of new core configurations (upgrade, conversion, refuelling, etc.) as well as several operational needs, such as calibration of the control rods, e.g. in case that reactivity insertion experiments are planned. The control rod worth can be assessed either experimentally or theoretically, mainly through the utilization of neutronic codes. In the present work two different theoretical approaches, i.e. a deterministic and a stochastic one are used for the estimation of the integral and the differential worth of two control rods utilized in the Greek Research Reactor (GRR-1). For the deterministic approach the neutronics code system SCALE (modules NITAWL/XSDRNPM) and CITATION is used, while the stochastic one is made using the Monte Carlo code TRIPOLI. Both approaches follow the procedure of reactivity insertion steps and their results are tested against measurements conducted in the reactor. The goal of this work is to examine the capability of a deterministic code system to reliably simulate the worth of a control rod, based also on comparisons with the detailed Monte Carlo simulation, while various options are tested with respect to the deterministic results' reliability.

  11. Characterization of control rod worths and fuel rod power peaking factors in the university of Utah TRIGA Mark I reactor

    OpenAIRE

    Alroumi Fawaz; Kim Donghoon; Schow Ryan; Jevremovic Tatjana

    2016-01-01

    Control rod reactivity (worths) for the three control rods and fuel rod power peaking factors in the University of Utah research reactor (100 kW TRIGA Mark I) are characterized using the AGENT code system and the results described in this paper. These values are compared to the MCNP6 and existing experimental measurements. In addition, the eigenvalue, neutron spatial flux distributions and reaction rates are analyzed and discussed. The AGENT code system is ...

  12. Simulation on the HTTR Control Rod Withdrawal Test

    International Nuclear Information System (INIS)

    This paper describes the GAMMA+ code simulation of HTTR control rod withdrawal test. The simulation is done to examine the effect of GAMMA+ code's single-zone and multi-zone point kinetics models on the prediction of the reactor power response during HTTR control rod withdrawal test. In addition, it has an objective to examine how the reactor power response is affected by the application of the fuel temperature coefficients on TRISO kernel or compact rod. The calculation results of reactivity response and reactor power response are compared with the test results which were obtained at the initial power of 15.2 MW with the amount of reactivity insertion by control rod withdrawal to 3.4e-04 (dk/k) in 6.59 seconds. All GAMMA+ simulation results on a HTTR CRW test showed good predictions with the measured data. In particular, TRISO Kernel Model where the fuel temperature coefficients applied on the TRISO particle produced a better prediction within a 1.5% measured data and made no difference between the single-zone model and the multi-zone point kinetics model. During the control rod withdrawal event which is a fast transient, the total reactivity is mainly affected by the inserted reactivity and the reactivity response due to the change of the fuel temperature and the graphite moderator temperature

  13. Simulation on the HTTR Control Rod Withdrawal Test

    Energy Technology Data Exchange (ETDEWEB)

    Jun, Ji Su; Tak, Nam-il; Lim, Hong Sik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    This paper describes the GAMMA+ code simulation of HTTR control rod withdrawal test. The simulation is done to examine the effect of GAMMA+ code's single-zone and multi-zone point kinetics models on the prediction of the reactor power response during HTTR control rod withdrawal test. In addition, it has an objective to examine how the reactor power response is affected by the application of the fuel temperature coefficients on TRISO kernel or compact rod. The calculation results of reactivity response and reactor power response are compared with the test results which were obtained at the initial power of 15.2 MW with the amount of reactivity insertion by control rod withdrawal to 3.4e-04 (dk/k) in 6.59 seconds. All GAMMA+ simulation results on a HTTR CRW test showed good predictions with the measured data. In particular, TRISO Kernel Model where the fuel temperature coefficients applied on the TRISO particle produced a better prediction within a 1.5% measured data and made no difference between the single-zone model and the multi-zone point kinetics model. During the control rod withdrawal event which is a fast transient, the total reactivity is mainly affected by the inserted reactivity and the reactivity response due to the change of the fuel temperature and the graphite moderator temperature.

  14. Estimation and control in HTGR fuel rod fabrication

    Energy Technology Data Exchange (ETDEWEB)

    Downing, D J; Bailey, M J

    1980-01-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Sease, J.D.

    1998-03-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-06-01

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

  19. Device for rearranging control rods of experimental reactors

    International Nuclear Information System (INIS)

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

  20. Improvement Research of Control Rod Drive Mechanism in CARR

    Institute of Scientific and Technical Information of China (English)

    ZHU; Xue-wei; ZHEN; Jian-xiao; LUO; Zhong; YANG; Kun; WANG; Yi-shi; JIA; Yue-guang

    2013-01-01

    We take an improvement research of synchronization in process of control rod drive mechanism(CRDM)inversion.An experimental prototype is designed based on the structure and function of the CRDM,we take some experiments on this experimental prototype,such as maximum loading force experiment,coil temperature rise experiment and stiffness experiment,achieve important magnetic

  1. Dysprosium and hafnium base absorbers for advanced WWER control rods

    International Nuclear Information System (INIS)

    Dysprosium titanate is an attractive control rod material for thermal neutron nuclear reactors such as WWER and RBMK. Its main advantages are almost non-swelling, no out-gassing under neutron irradiation, quit high neutron efficiency, a high melting point (∼ 1870 deg. C), non-interaction with the cladding at temperatures above 1000 deg. C, simple fabrication. nonradioactive waste and easy to reprocess. The dysprosium titanate control rods have worked without operating problems in the reactor MIR during 17 years and in WWER-1000 4 years. After post-irradiation examinations, this long-life control rod type was recommended for using in the nuclear reactors. Dysprosium hafnate is a promising absorber ceramic material. The research results confirmed that it has a large radiation damage resistance. The examination results of hafnium dummies (GFE-1) irradiated in BOR-60 are presented. The maximum accumulated neutron fluence was 3.4 x 1022cm-2 (E>0.1 MeV) and the temperature range was 340 to 360 deg. C. Due to high radiation growth (3-4 %) and the absence of an axial gap between the dummy and the upper capsule tip the dummies were bent. The irradiated dummies have high mechanical properties. Other aspects of the expected hafnium irradiation behaviour and the use of hafnium in control rods are discussed. This report presents some experimental data on Dy2O3·TiO2, Hf, Dy2O3·HfO2 and possibilities of their use in WWER control rods. (author)

  2. Measurements of Control Rod Worths in Critical and Exponential Assemblies

    International Nuclear Information System (INIS)

    Control rods of cadmium, stainless steel and a Cd -In -A g alloy have been investigated by means of three different methods: (a) Measurements of buckling differences and migration areas in a zero-power reactor (R0); (b ) Pulsed subcritical measurements in RO with a compact neutron pulse generator; and (c ) Buckling measurements in an exponential assembly (ZEBRA). The measurements were made in different lattices of natural uranium metal rods and heavy water. Radial and axial statistical weights for one and two control rods were measured by means of method (a) with an accuracy in ΔB2 of 0.005 m-2. The upper limit in the value of ΔB2 is about -1.5 m-2, equivalent to -4% of reactivity. The accuracy in the pulsed measurements is of the order of 5%, but in this case it is possible to measure very large negative reactivities. In the exponential assembly we have tried to separate the thermal and epithermal absorption effects by measuring the control-rod worths with only moderator in the tank as well as in different reactor lattices. The accuracy was of the same order as in the critical measurements. (author)

  3. Anti-ejection system for control rod drives

    International Nuclear Information System (INIS)

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

  4. 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.

  5. Analysis and test verification of control rod buffer in HTR

    International Nuclear Information System (INIS)

    The thin-walled shell buffer in high temperature gas-cooled reactor (HTR) was designed for absorbing the kinetic energy of the control rod drop in the drive line fracture accident. The thin-walled cylinder structure satisfying the requirements of actual working condition was design by using the energy absorption model of the classical cylinder shell under axial pressure. By using ABAQUS/explicit with J-C constitutive model, the finite element models of both the real reactor condition and the test condition were built to simulate the collision. Based on the analysis results, the control rod fall- down test was designed and implemented. The test results demonstrate that stable pro gressive buckling occurs when the full size buffer is impacted by equiponderance test bar, and the buffer can reduce the crush force effectively and protect the graphite from being destroyed. The analysis results show that the test model can represent and envelope the real condition in reactor. (authors)

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

    OpenAIRE

    Markov P.; Valtr O.

    2007-01-01

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

  7. A Complete Analysis for Pump Controlled Single Rod Actuators

    OpenAIRE

    Çalışkan,Hakan; Balkan, Tuna; Platin, Bülent E.

    2016-01-01

    In the current study a variable speed pump controlled hydrostatic circuit where an underlapped shuttle valve is utilized to compensate the unequal flow rate of a single rod actuator is analyzed. Parameters of the shuttle valve are included in the system analysis, rather than treating it as an ideal switching element as handled in literature. A linearized model of the system is obtained. An inverse kinematic model, which calculates the required pump drive speed for a desired actuator speed and...

  8. Characterization of control rod worths and fuel rod power peaking factors in the university of Utah TRIGA Mark I reactor

    Directory of Open Access Journals (Sweden)

    Alroumi Fawaz

    2016-01-01

    Full Text Available Control rod reactivity (worths for the three control rods and fuel rod power peaking factors in the University of Utah research reactor (100 kW TRIGA Mark I are characterized using the AGENT code system and the results described in this paper. These values are compared to the MCNP6 and existing experimental measurements. In addition, the eigenvalue, neutron spatial flux distributions and reaction rates are analyzed and discussed. The AGENT code system is widely benchmarked for various reactor types and complexities in their geometric arrangements of the assemblies and reactor core material distributions. Thus, it is used as a base methodology to evaluate neutronics variables of the research reactor at the University of Utah. With its much shorter computation time than MCNP6, AGENT provides agreement with the MCNP6 within a 0.5 % difference for the eigenvalue and a maximum difference of 10% in the power peaking factor values. Differential and integral control rod worths obtained by AGENT show well agreement with MCNP6 and the theoretical model. However, regulating the control rod worth is somewhat overestimated by both MCNP6 and AGENT models when compared to the experimental/theoretical values. In comparison to MCNP6, the total control rod worths and shutdown margin obtained with AGENT show better agreement to the experimental values.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-06-01

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-10-15

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

  13. Control rod reactivity worth determination of a typical MTR research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Rizwan, M.; Raza, S.S.; Khan, R. [Pakistan Institute of Engineering and Applied Sciences (PIEAS), Islamabad (Pakistan). Dept. of Nuclear Engineering

    2015-10-15

    The safe and reliable utilization of research reactor demands the possible accurate information of control rod (CR) worths. The criticality positions of the control rods changes with time due to build up fission products. It is therefore important to determine the reactivity worth of control rods. The aim of this article is to estimate the reactivity worth of controls rods in the equilibrium core of a Materials Testing Reactor (MTR). A deterministic model of the reactor core was developed and confirmed against the reference results of excess reactivity, shutdown margin and combined control rod reactivity worth using the combination of WIMS/D4 and CITATION computer codes.

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

    Directory of Open Access Journals (Sweden)

    Markov P.

    2007-10-01

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

  15. Managing the aging of BWR control rod drive systems

    International Nuclear Information System (INIS)

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

  16. Control assembly behaviour, inspection techniques and remedies against wear at pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Knaab, H.; Fuchs, H.P. (Siemens AG Unternehmensbereich KWU, Erlangen (Germany, F.R.))

    1989-08-01

    Rod cluster control assemblies as fabricated by Siemens for pressurized water reactors have been in use for up to 20 plant operating cycles. All of them have demonstrated excellent operational behaviour. The design features austenitic steel spiders formed by electrical discharge machining from a forging, lock-welded screw connections for the individual rods, and swelling-resistant Ag-In-Cd absorbers. Additionally, several optional features are available and in use for improved resistance against wear and irradiation-assisted stress corrosion cracking. Specialized non-destructive testing techniques were developed and are in use for routine in-pool inspections of the assemblies for fretting, clad thinning and irradiation-assisted stress corrosion cracking. (orig.).

  17. Decontamination of control rod housing from Palisades Nuclear Power Station.

    Energy Technology Data Exchange (ETDEWEB)

    Kaminski, M.D.; Nunez, L.; Purohit, A.

    1999-05-03

    Argonne National Laboratory has developed a novel decontamination solvent for removing oxide scales formed on ferrous metals typical of nuclear reactor piping. The decontamination process is based on the properties of the diphosphonic acids (specifically 1-hydroxyethane-1,1-diphosphonic acid or HEDPA) coupled with strong reducing-agents (e.g., sodium formaldehyde sulfoxylate, SFS, and hydroxylamine nitrate, HAN). To study this solvent further, ANL has solicited actual stainless steel piping material that has been recently removed from an operating nuclear reactor. On March 3, 1999 ANL received segments of control rod housing from Consumers Energy's Palisades Nuclear Plant (Covert, MI) containing radioactive contamination from both neutron activation and surface scale deposits. Palisades Power plant is a PWR type nuclear generating plant. A total of eight segments were received. These segments were from control rod housing that was in service for about 6.5 years. Of the eight pieces that were received two were chosen for our experimentation--small pieces labeled Piece A and Piece B. The wetted surfaces (with the reactor's pressurized water coolant/moderator) of the pieces were covered with as a scale that is best characterized visually as a smooth, shiny, adherent, and black/brown in color type oxide covering. This tenacious oxide could not be scratched or removed except by aggressive mechanical means (e.g., filing, cutting).

  18. Application of homogeneity procedure for partly immerse control rods in axially reflected system

    International Nuclear Information System (INIS)

    Homogenization procedure is applied for calculating the reactivity of heavy water reactor and depth of control rods immersion in order to maintain the criticality dependent on the fuel burnup. Since a real reactor is axially reflected a practical formula is derived for obtaining homogenized L2 values for reflector containing control rod lattice. Computer codes for standard calculation of control rod parameters in power thermal reactor

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

    International Nuclear Information System (INIS)

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

  20. Methods of Control-Rod Calibration in the Windscale Advanced Gas-Cooled Reactor

    International Nuclear Information System (INIS)

    Different techniques were used to calibrate control rods and to measure individual rod worths during the commissioning of the WAGR. These methods are described and the results are presented. The methods described are: (a ) Air poisoning - Changes in air pressure allow axial movement of rods at the critical condition. Thus rod movement can be related to pressure variation which is equivalent to a reactivity change. Also, rod slope measurements can be made for different rod insertions. (b) Rod slopes - The rods are withdrawn to make the reactor supercritical; then later they are inserted to make the reactor subcritical. From a measurement of the doubling and halving times the change in reactivity between the super- and subcritical states can be determined.- (c ) Absorber addition and withdrawal - The number of fixed localized absorbers is varied to give a method similar to the use of uniform air poisoning. (d) Pulsatron - This technique is used to give subcritical measurements of rod worth. (e) Rod run-in - The reactor is initially critical, and then the rods are run into the core. Analysis of the flux response relates reactivity to rod movement. (author)

  1. Dysprosium hafnate as absorbing material for control rods

    International Nuclear Information System (INIS)

    Dysprosium hafnate is proposed as a promising absorbing material for control rods of thermal nuclear reactors. The properties of dysprosium hafnate pellets with different Dy and Hf contents are presented in this article. The fluorite phase is characterized by the density range 6.8-7.8 g/cm3 and; the thermal diffusivity achieves 0.58-0.83 mm2/s at 20 deg. C, thermal conductivity of 1.5-2.0 W/(K m) and TLEC of (8.4-8.6) x 10-6 K-1 at 20 deg. C. The temperature dependence of the thermophysical properties of dysprosium hafnate are presented. The neutron absorption efficiency of dysprosium hafnate was estimated in comparison with boron carbide. The radiation resistance of pellets after irradiation in the BOR-60 reactor is presented as well

  2. Radial brake assembly for a control rod drive

    International Nuclear Information System (INIS)

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

  3. The Design of Control-Rod Drives for Large Graphite-Moderated Reactors

    International Nuclear Information System (INIS)

    Because graphite-moderated tube-type power or desalinisation reactors are more economical in the larger ratings, control-rod drives may require strokes in the 20 to 60 ft range. Speed-of-insertion requirements may vary by a factor of 300 to 1 between the low-speed normal control requirements and the high-speed emergency shutdown requirements. Internal rod cooling is often required in addition to the prevention of reactor atmosphere leakage where the control rod penetrates the .reactor envelope. These requirements in addition to those of rod deceleration, shielding, space limitations, stored or emergency energy sources, maintenance provisions and overall drive-system cost increase the design problems associated with control rods for this type of reactor. Several unique control and/or shutdown rod drives have been designed for horizontal and vertical operation in large graphite-moderated power and study reactors. These designs include (1) air-operated shutdown rods with high insertion speeds, (2) hydraulic motor-driven, chain-type shutdown control rods with short storage sections and a compact drive; and (3) hydraulic cylinder-operated, force-multiplication shutdown control rods. Each of these drives compromises the requirements listed above to some extent; however, operable drives have been designed and tested. (author)

  4. Analytical estimation of control rod shadowing effect for excess reactivity measurement of High Temperature Engineering Test Reactor (HTTR)

    Energy Technology Data Exchange (ETDEWEB)

    Nakano, Masaaki; Yamashita, Kiyonobu; Fujimoto, Nozomu; Nojiri, Naoki; Takeuchi, Mitsuo; Fujisaki, Shingo [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Tokuhara, Kazumi; Nakata, Tetsuo

    1998-05-01

    The control rod shadowing effect has been estimated analytically in application of the fuel addition method to excess reactivity measurement of High Temperature Engineering Test Reactor (HTTR). The movements of control rods in the procedure of the fuel addition method have been simulated in the analysis. The calculated excess reactivity obtained by the simulation depends on the combinations of measuring control rods and compensating control rods and varies from -10% to +50% in comparison with the excess reactivity calculated from the effective multiplication factor of the core where all control rods are fully withdrawn. The control rod shadowing effect is reduced by the use of plural number of measuring and compensation control rods because of the reduction in neutron flux deformation in the measuring procedure. As a result, following combinations of control rods are recommended; 1) Thirteen control rods of the center, first, and second rings will be used for the reactivity measurement. The reactivity of each control rod is measured by the use of the other twelve control rods for reactivity compensation. 2) Six control rods of the first ring will be used for the reactivity measurement. The reactivity of each control rod is measured by the use of the other five control rods for reactivity compensation. (author)

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

    International Nuclear Information System (INIS)

    An investigation of control rod shadowing and anti-shadowing (interaction) effects has been carried out in the context of a design study of the control rod pattern for the large 2400 MWth Generation IV Gas-cooled Fast Reactor (GFR). For the calculations, the deterministic code system ERANOS-2.0 has been used, in association with a full core model including a European Fast Reactor (EFR)-type pattern for the control rods. More specifically, the core contains a total of 33 control (CSD) and safety (DSD) rods implemented in three banks: -1) a first bank of 6 CSD rods, placed at 64 cm from core centre in the inner fuel zone (Pu content 16.3 % vol.), -2) a safety bank consisting of 9 DSD rods, at an average distance of 118 cm, and -3) a third bank with 18 CSD rods, placed at 171 cm, i.e. at the interface between the inner and outer (Pu content 19.2 % vol.) core regions. Each control rod has been modelled as a homogeneous material containing 90%-enriched B4C, steel and helium. Considerable shadowing effects have been observed between the first bank and the safety bank, as also between individual rods within the first bank. Large anti-shadowing effects take place in an even greater number of the studied rod configurations. The largest interaction is between the two CSD banks, the anti-shadowing value being 46% in this case, implying that the total rod worth is increased by a factor of almost 2 when compared to the sum of the individual bank values. Additional investigations have been performed, in particular the computation of the first order eigenvalue and the eigenvalue separation. The main finding is that the interactions are lower when one of the control rod banks is located at a radial position corresponding to half the core radius. (authors)

  6. Problems Related to the Nuclear and Mechanical Design of the Programma Reattore Organico Reactor Control Rods

    International Nuclear Information System (INIS)

    The paper illustrates the methods used for calculating the nuclear design of the control rods in the preliminary and operational phases of the PRO project. Comparisons are made with experimental data and a summary is given of the programming studies carried out. Finally, consideration is given to certain problems connected with the mechanical design of the control rods. (author)

  7. Analysis of the fluid and structure interaction in the control rod drop process

    International Nuclear Information System (INIS)

    Background: The drop time of control rod assembly is one of the most important parameters to ensure the safe operation of nuclear power plants. Due to the fluid-structure interaction (FSI), the elastic structures, such as control rods and guide tubes, will vibrate in the dropping of control rod assembly. The impact and friction between the control rod and guide tubes caused by large transverse vibration will influence the drop time calculation. Purpose: To study in detail the flow-induced vibration and the friction, this paper focus on the fluid-structure interaction in the control rod drop process. Methods: Firstly, the vibration equations of control rod and guide tubes considering the fluid-structure interaction are established. Then the various fluid forces are analyzed in accordance with their qualities, and the influences of different guide tubes in a guide tubes array are also considered. Results: The friction between control rod and guide tube is not zero, and the friction under seismic condition is larger. Conclusions: The analysis on the fluid and structure interaction presented in this paper is reasonable and can improve current analytical models of control rod drop time calculation. (authors)

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

    International Nuclear Information System (INIS)

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

  9. Structure Optimization Design of the Electronically Controlled Fuel Control Rod System in a Diesel Engine

    OpenAIRE

    Hui Jin; Haosen Wang

    2015-01-01

    Poor ride comfort and shorter clutch life span are the key factors restricting the commercialization of automated manual transmission (AMT). For nonelectrically controlled engines or AMT where cooperative control between the engine and the transmission is not realizable, applying electronically controlled fuel control rod systems (ECFCRS) is an effective way to solve these problems. By applying design software such as CATIA, Matlab and Simulink, and MSC Adams, a suite of optimization design m...

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

    International Nuclear Information System (INIS)

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

  11. Control rod assembly drops to fully inserted position

    Energy Technology Data Exchange (ETDEWEB)

    Groudev, Pavlin P. [Institute for Nuclear Research and Nuclear Energy, Bulgarian Academy of Sciences, Tzarigradsko Shaussee 72, Sofia 1784 (Bulgaria)]. E-mail: pavlinpg@inrne.bas.bg; Stefanova, Antoaneta E. [Institute for Nuclear Research and Nuclear Energy, Bulgarian Academy of Sciences, Tzarigradsko Shaussee 72, Sofia 1784 (Bulgaria)]. E-mail: antoanet@inrne.bas.bg

    2005-11-15

    This paper describes validation of a computer model that has been developed for VVER 440 Nuclear Power Plant (NPP) for use with RELAP5/MOD 3.2 computer code in the analysis of the following transient: 'Control rod assembly drops to fully inserted position'. This validation is a process that compares the analytical results obtained by RELAP5/MOD3.2 model of the VVER 440 with experimental transient data received from Kozloduy NPP, Unit no. 2. The model of VVER 440 was developed at the the Institute for Nuclear Research and Nuclear Energy for analyses of operational occurrences, abnormal events and design basis scenarios. It will provide a significant analytical capability for the Bulgarian technical specialists located at the Kozloduy NPP. The criteria used in selecting transient are: importance to safety, availability and suitability of data followed by suitability for RELAP5 code validation. The comparisons between the RELAP5 results and the test data indicate good agreement.

  12. Nonlinear Magnetic Circuit Analysis of SMART Control Rod Drive Actuator

    Energy Technology Data Exchange (ETDEWEB)

    Noh, Myounggyu; Gi, Myung Ju; Kim, Myounggon; Park, Youngwoo [Chungnam Nat' l Univ., Daejeon (Korea, Republic of); Lee, Jaeseon; Kim, Jongwook [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    In this paper, we derive a nonlinear magnetic circuit model of an electromagnetic control-rod actuator in the SMART. The results of the nonlinear model are compared with those by linear circuit model and finite-element analyses. gnetic circuit modeling is a useful tool when designing an electromagnetic actuator, as it allows fast calculations and enables parametric studies. It is particularly essential when the actuator is to be used in a very complex system such as a nuclear reactor. Important design parameters must be identified at the early stage of the design process. Once the design space is narrowed down, more accurate methods such finite-element analyses (FEA) can be employed for detailed design. Magnetic circuit modeling is based on the assumption that a flux path consists of sections in each of which field quantities are constant with linear constitutive relations. This assumption fails to hold when portions of the flux path become saturated. The magnetic circuit must be modified in order to accurately describe the nonlinear behavior of saturation.

  13. Fuel element reshuffling and fuel follower control rods (FFCR) replacement for PUSPATI TRIGA reactor

    International Nuclear Information System (INIS)

    The PUSPATI TRIGA Reactor has been utilized for more than 25 years using the same fuel elements and control rods. Generally, there are four control rods being used to control the neutron production inside the reactor core. A maintenance program has been developed to ensure its integrity, capability and safety of the reactor and it has been maintained twice a year since the first operation in 1982. The activities involve during the maintenance period including fuel elements and control rods inspections, electronics and mechanical systems, and others related works. During the maintenance in August 2008, there are some irregularities found on the fuel follower control rods and needed to be replaced. Even though the irregularities was not contributed into any unwanted incident, it were decided to replace with new control rods to avoid any potential hazards and unsafe condition occurred during operation later. Replacing any of the control rods would involved in imbalance of neutron flux and power distribution inside the core. Therefore, a number of fuel elements need to be reshuffled in order to compensate the neutron flux and power distribution as well as to balance the fuel elements burn-up in the core. This paper will described the fuel elements reshuffling and fuel follower control rods (FFCR) replacement for PUSPATI TRIGA Reactor. (Author)

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

    International Nuclear Information System (INIS)

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

  15. A simple approach to eliminate background signals in dynamic control rod reactivity measurements for LWRs

    Energy Technology Data Exchange (ETDEWEB)

    Lee, E. K.; Woo, I. T.; Shin, H. C.; Ryu, S. J.; Bae, S. M. [Korea Electric Power Research Institute, Taejon (Korea, Republic of)

    2003-07-01

    Dynamic rod worth measurement (DRWM{sup TM}) methodology commercialized by Westinghouse was successfully applied to many nuclear power plants in USA to measure the control rod worth at Low Power Physics Tests. But in Korea, to increase the capacity of nuclear power plant, KEPRI has developed Dynamic Control rod Reactivity Measurement (DCRM) system using more rapid and sophisticated reactivity measurement methodology without the change of boron concentration. The object of this paper is to consider the practical method to eliminate background signals from measured ex-core detector signals. Because of relatively low rod insertion speed (40 {approx} 48 steps/min), the background signals affect the final results severely. Therefore a simple and practical method based on the behavior of integral rod worth curve was developed and applied. A total of 26 experimental results show that the proposed approach works to figure out the background signals.

  16. Failure of latch mechanism for motion control of safety rods

    International Nuclear Information System (INIS)

    During safety rod tests in K-reactor prior to startup, one safety rod could not be lifted because the ''button'' broke off and became lodged in the mechanism. Examination of the failed latch assembly along with other assemblies from both K-Area and L-Area revealed several missing buttons as well as severely deformed ''jaw hanger extensions.'' We participated in the investigation of the damage by request of the Reactor Restart Section. Based on our study of the latch mechanism, the modifications to the ''safety rod extension,'' and the operating history of the machine, this memorandum describes the causes of the observed damage with experimental evidence and calculations to support the findings. 3 refs

  17. Development of hydride absorber for fast reactor. Application of hafnium hydride to control rod of large fast reactor

    International Nuclear Information System (INIS)

    The application of hafnium hydride (Hf-hydride) to a control rod for a large fast reactor where the B4C control rod is originally employed is studied. Three types of Hf-hydride control rods are designed. The control rod worth and its change during the burnup are evaluated for different hydrogen-to-hafnium ratios and are compared with those of the original B4C control rod. The result indicates that the worths of the Hf-hydride and the 10B-enriched B4C control rods are approximately the same, and the lifetime of the Hf-hydride control rod is almost four times longer than that of the 10B-enriched B4C control rod. The core performances of the shutdown margin, sodium void reactivity, Doppler reactivity coefficient, and breeding ratio are analyzed. It is indicated that those for the Hf-hydride control rod are almost the same as those for the original B4C control rod. The behavior of neutrons moderated by the Hf-hydride control rod is analyzed. It is confirmed that the Hf-hydride control rod does not cause any thermal spike problems in the fast reactor core. (author)

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    This report documents the Contractor Readiness Assessment (CRA) for TREAT Fuel Movement and Control Rod Drives Isolation. The review followed the approved Plan of Action (POA) and Implementation Plan (IP) using the identified core requirements. The activity was limited scope focusing on the control rod drives functional isolation and fuel element movement. The purpose of this review is to ensure the facility's readiness to move fuel elements thus supporting inspection and functionally isolate the control rod drives to maintain the required shutdown margin.

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

    Energy Technology Data Exchange (ETDEWEB)

    Rowsell, David Leon [Idaho National Laboratory (INL), Idaho Falls, ID (United States)

    2015-06-01

    This report documents the Contractor Readiness Assessment (CRA) for TREAT Fuel Movement and Control Rod Drives Isolation. The review followed the approved Plan of Action (POA) and Implementation Plan (IP) using the identified core requirements. The activity was limited scope focusing on the control rod drives functional isolation and fuel element movement. The purpose of this review is to ensure the facility's readiness to move fuel elements thus supporting inspection and functionally isolate the control rod drives to maintain the required shutdown margin.

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

    International Nuclear Information System (INIS)

    The properties of rare earth elements as the material for control rods were studied. The rare earth elements, especially europium oxide, has the nuclear property corresponding to boron carbide, and its neutron absorption process does not emit alpha particles. The elements produced as a result of neutron capture also have large capture cross sections. This paper presents survey report on the properties and nuclear properties of rare earth elements, and comparison with other materials. Preliminary experiment was performed to make the pellets of europium oxide, and is described in this paper. Because of large density, the crystal form to be made was monoclinic system. Europium hydroxide was decomposed at 10000C and 10-5 torr. The obtained powder was dipped into benzene, and dryed in the air at 4500C. This powder was pressed and sintered in the air for one hour at 15000C. The density of the obtained pellets was 97.0% of the theoretical density. The cross section of europium for fast neutron absorption is not yet accurately obtained, and is in the range between 4.65 and 8.5 barn for 151Eu(n,γ) reaction. Since chain absorption reaction is caused in Eu, the overall capability of neutron absorption is not much changed by the loss of original material due to absorption. The pellets of europium oxide may be handled in air, but must be kept in dry atmosphere. The reactions of europium oxide with various metals were also investigated. The characteristic behavior in case of irradiation depends on the amount of silicon contained, and it was very good if the amount was less than 0.03%. (Kato, T.)

  2. Physical resuspension and revaporisation phenomena in control rod aerosols

    International Nuclear Information System (INIS)

    Physical resuspension and revaporisation processes could play a significant role in the transport of fission products in a severe reactor accident. The processes involved in physical resuspension and revaporisation of control rod alloy aerosol particles from a stainless steel substrate have been studied at room temperature under laminar and turbulent flow conditions (Reynolds numbers of between 70 and 7000), and at temperatures in the range from 370 K to 870 K under laminar and intermediate flow conditions (Reynolds numbers of between 7 and 1400) in the absence and presence of steam. The phenomena were investigated using bulk analyses to determine the quantity of material remaining on a coupon after each experiment, and standard surface analysis techniques were used to examine the composition and morphology of the particles. The main conclusions of this work are that: (i) physical resuspension is only significant in turbulent flow, (ii) two processes are involved in physical resuspension: the removal of surface layers which are only loosely bound to the substrate, and the removal of a more tightly-bound layer, (iii) the amount of material resuspended decreases exponentially with time, and the data have been correlated with a reverse isotherm model, (iv) the weight loss from the revaporisation experiments can be interpreted in terms of the effective vapour pressure of the deposit, and an equation has been derived to express this vapour pressure as a function of temperature. These studies have demonstrated the importance of a number of resuspension processes in generating a source of radioactive material that could be released after failure of the containment. Efforts are in hand to include these phenomena in the relevant modelling studies. (author)

  3. Estimate of control rods effectiveness of the RP-0 reactor 7A2 core by the rod-drop method using a compensated ionization chamber

    International Nuclear Information System (INIS)

    Value estimate results of the four control rods by the rod-drop method are presented using the 'point reactor model' for the RP-0 reactor 7A2 core employing the inverse kinetics neutronic noise equipment and a compensated ionization chamber located in the E2 core. At every moment, the reactor power was known and it was calibrated with the same equipment

  4. Preliminary Investigation of an Optimally Scramming Control Rod for Gas-Cooled Reactors

    International Nuclear Information System (INIS)

    A passively safe control rod for gas-cooled reactors is proposed. This Optimally Scramming Control Rod (OSCR) is lifted out of the core region by the core coolant and descends back into the core when the coolant flow is not sufficient for core cooling purposes or in the event of depressurization. It is shown that for the current design of the OSCR, the reactor can be operated under normal lower power conditions down to about 80% of total power. It is also shown that cold shutdown can be achieved with rods of sufficiently low mass to allow naturally passive operation of the concept. (authors)

  5. Validation of neutron flux redistribution factors in JSI TRIGA reactor due to control rod movements.

    Science.gov (United States)

    Kaiba, Tanja; Žerovnik, Gašper; Jazbec, Anže; Štancar, Žiga; Barbot, Loïc; Fourmentel, Damien; Snoj, Luka

    2015-10-01

    For efficient utilization of research reactors, such as TRIGA Mark II reactor in Ljubljana, it is important to know neutron flux distribution in the reactor as accurately as possible. The focus of this study is on the neutron flux redistributions due to control rod movements. For analyzing neutron flux redistributions, Monte Carlo calculations of fission rate distributions with the JSI TRIGA reactor model at different control rod configurations have been performed. Sensitivity of the detector response due to control rod movement have been studied. Optimal radial and axial positions of the detector have been determined. Measurements of the axial neutron flux distribution using the CEA manufactured fission chambers have been performed. The experiments at different control rod positions were conducted and compared with the MCNP calculations for a fixed detector axial position. In the future, simultaneous on-line measurements with multiple fission chambers will be performed inside the reactor core for a more accurate on-line power monitoring system. PMID:26141293

  6. BWR control rod patterns and fuel loading optimization using heuristic methods

    International Nuclear Information System (INIS)

    We show the results obtained with the OCOTH system to optimize the Fuel Reloads Design and Control Rod Patterns Design in a Boiling Water Reactor. Our system solves both problems in a coupled way. We used the 3-dimensional CM-PRESTO code to evaluate the solutions quality. The process has three stages. In the first step we obtain a Fuel Reload Design 'seed' using the Haling's principle. The followings steps are an iterative process between the Control Rod Patterns Designs and Fuel Reloads Design. Control Rod Patterns Design is proposed for the Fuel Reload Design 'seed' and then Control Rod Patterns Design is used to find a new Fuel Reload Design. Both processes are coupled in an iterative loop until a criterion stop is fulfilled. In the whole process, the genetic algorithms, neural networks and ant colony system optimization techniques were used. (authors)

  7. Coolability of a control rod which has melted and foamed in its septifoil channel

    Energy Technology Data Exchange (ETDEWEB)

    Walkowiak, D.A.

    1991-10-01

    During a Loss of Control Rod Cooling (LCRC) event, the control rods which are in the affected septifoil can be postulated to melt. Melting of a control rod which has been irradiated creates a special concern since the entrapped gases expand rapidly and cause the melt to manifest itself initially in a foamed state. The foamed material then contacts the septifoil outer housing and the inner septifoil web material, where heat is conducted out of the foamed material. A second concern relating to the foamed melt is that its thermal conductivity is greatly reduced from that of the solid material, and also that of the non-foamed liquid. The purpose of this report is to address how, even in the presence of decreased thermal conductivity, the foamed melt may aid in cooling the control rod material.

  8. Operation and maintenance experience with control rod and their drive mechanisms of fast breeder test reactor

    International Nuclear Information System (INIS)

    This paper explains the functional and construction features of Control Rod Drive Mechanism (CRDM) and control rod used in Fast Breeder Test Reactor (FBTR) which is a 40 MWt loop type sodium cooled fast reactor. It discusses all safety related incidents and failures encountered during its service in reactor, the solutions evolved and modifications carried out to prevent recurrence. It also details the maintenance activities and periodical surveillance carried out. The results of a reliability analysis done are also discussed. (author)

  9. Validation of neutron flux redistribution factors in JSI TRIGA reactor due to control rod movements

    International Nuclear Information System (INIS)

    For efficient utilization of research reactors, such as TRIGA Mark II reactor in Ljubljana, it is important to know neutron flux distribution in the reactor as accurately as possible. The focus of this study is on the neutron flux redistributions due to control rod movements. For analyzing neutron flux redistributions, Monte Carlo calculations of fission rate distributions with the JSI TRIGA reactor model at different control rod configurations have been performed. Sensitivity of the detector response due to control rod movement have been studied. Optimal radial and axial positions of the detector have been determined. Measurements of the axial neutron flux distribution using the CEA manufactured fission chambers have been performed. The experiments at different control rod positions were conducted and compared with the MCNP calculations for a fixed detector axial position. In the future, simultaneous on-line measurements with multiple fission chambers will be performed inside the reactor core for a more accurate on-line power monitoring system. - Highlights: • Neutron flux redistribution due to control rod movement in JSI TRIGA has been studied. • Detector response sensitivity to the control rod position has been minimized. • Optimal radial and axial detector positions have been determined

  10. Test results of dynamic control rod reactivity measurements method for LWRs

    Energy Technology Data Exchange (ETDEWEB)

    Lee, E. K.; Woo, I. T.; Shin, H. C.; Ryu, S. J.; Bae, S. M.; Lee, Y. G. [Korea Electric Power Research Institute, Taejon (Korea, Republic of)

    2003-10-01

    Recently, KEPRI has developed the Dynamic Control rod Reactivity Measurement (DCRMTM) methodology to measure the worths of control rod bank and safety rod bank which should be verified during the Low Power Physics Test (LPPT). DCRM has been applied to measure the worths of total 27 banks of six different nuclear power plants, including 2-and 3-Loop WH reactors and Korea Standard Nuclear Plants. The most sensitive part in the method is how to extract the background signals from the original data. To solve it, a simple approach reflecting the characteristic of dynamic reactivity was developed. Final results of 27 cases show that the average and standard difference between measurements and the estimations of core design code is 3.6%, 2.5% respectively, while the current rod worth measurement method 4.3% and 3.2%. Maximum error also decreases from 12.8% to 9%. It takes about 15 minutes to measure one rod bank. From the all observations, one knows definitely that DCRM can be an appropriate method to substitute the current boron dilution and rod swap method for measuring the rod worth.

  11. Physics and Material Problems of Reactor Control Rods. Proceedings of the Symposium on Physics and Material Problems of Reactor Control Rods

    International Nuclear Information System (INIS)

    The development of nuclear reactors is closely associated with the progress made in the solution of control problems. To survey the present state of the subject the International Atomic Energy Agency convened a symposium devoted to ''Physics and Material Problems in Reactor Control Rods''. The Symposium was held in Vienna from 11 to 15 November 1963 and was attended by more than 100 participants representing 21 of the Agency's Member States and two international organizations. Problems discussed in the 34 papers presented at 8 sessions covered many special aspects of theoretical and experimental physics, engineering, metallurgy, etc. The first session of the Symposium was devoted to different theoretical methods used for the determination of control rod effectiveness in a multi- regioned reactor, and in natural-uranium heavy-water moderated cores. Homogeneous and heterogeneous approaches were discussed and applicability of proposed methods for various forms of control elements considered. During the two following sessions a number of theoretical problems and mathematical models were examined together with various control rod experiments and measurements in exponential and critical assemblies and at commercial nuclear power stations. The next session dealt with the connection between physics and technology of control rods, the latter being the subject of the remainder of the Symposium. Testing and actual operating experience of control rods were also treated in some of the presented papers. The session on engineering aspects of control rod systems included presentation of research results in a marine control station, the design of large graphite reactor control drives and the description of different mechanisms for rapid insertion of control absorbers. Finally, the methods of fast reactor control were discussed, followed by the presentation of various ''unconventional'' methods of reactivity control, such as hydraulic ball, fluidized bed, gas pressure and soluble

  12. Development of dynamic control rod reactivity measurement methodology and computer code system for PWR

    Energy Technology Data Exchange (ETDEWEB)

    Zee, Sung Quun; Lee, Chung Chan; Song, Jae Seung [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2002-09-01

    In order to apply dynamic control rod reactivity measurement (DCRM) method to domestic nuclear power reactor, the methodology of EPRC, 'Dynamic Reactivity Measurement of Rod Worth', was reviewed. It was also reviewed that items should be improve in three-dimensional kinetics code MASTER, which was developed by Korea Atomic Energy Research Institute, for use in DCRM. The validity of DORT two-dimensional synthesis method to calculate excore detector weighting factor were benchmarked via Yonggwang Unit 3 three-dimensional TORT calculation. The consistency of MASTER static core calculation results using neutron cross sections generated by commercial design tools PHENIX/ANC and DIT/ROCS were also verified via rodded and unrodded radial power distributions and control rod worth comparisons. 14 refs., 28 figs., 3 tabs. (Author)

  13. Full Scale Component Test Facility KOPRA - Qualification Test of EPR Control Rod Drive Mechanism

    Energy Technology Data Exchange (ETDEWEB)

    Sykora, Alexander; Herr, Wolfgang [AREVA NP GmbH, P.O. Box 3220, 91050 Erlangen (Germany); Champomier, Francois [AREVA NP SAS, Tour AREVA - Cedex 16, 92084 Paris-La Defense (France)

    2008-07-01

    The test facility KOPRA is designed for full scale-tests on nuclear components under operational conditions. One part of it is the component test loop for developing and qualifying nuclear core components respecting temperature, pressure and mass flow of pressurized water reactor conditions. The KOPRA test facility and its measuring equipment is presented through qualification tests for the control rod drive mechanism and the control rod drive line of the new European Pressurized Water Reactor (EPR). The control rod drive mechanism qualification test program is split into three different test phases. At first, performance tests are conducted to verify the adequate performance of the new equipment, e.g. measurement of rod cluster control assembly drop time under different thermal hydraulic conditions, impact velocity of drive rod on CRDM latch tips and drive rod acceleration during stepping operation by means of strain gauges or through direct measurement. After these functional tests follow the stability tests to ensure that proper functioning is reliably achieved over an appreciable amount of time and the endurance tests to quantify the amount of time and/or the number of steps during which no appreciable wear, that could possibly alter the correct behaviour, is to be expected. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

  15. On-line monitoring of control rod integrity in BWRs using a mass spectrometer

    Science.gov (United States)

    Larsson, I.; Loner, H.; Ammon, K.; Sihver, L.; Ledergerber, G.

    2013-01-01

    Surveillance of fuel and control rod integrity in the core of a boiling water reactor is essential for maintaining a safe and reliable operation. Control rods of a boiling water reactor are mainly filled with boron carbide as a neutron absorber. Due to the irradiation of boron with neutrons, a continuous production of lithium and helium will occur inside a control rod. Most of the created helium will be retained in the boron carbide lattice; however a small part will escape into the void volume of the control blade. Therefore the integrity of control rods during operation can efficiently be followed by on-line measurements of helium concentration in the reactor off-gas system using a mass spectrometer. Since helium is a fill gas in fuel rods, the same method is a useful early warning system for primary fuel failures. In this paper, we introduce an on-line helium detector system which is installed at the nuclear power plant in Leibstadt. Furthermore the measuring experiences of control rod failure detection at the plant are presented. Different causes of increased helium levels in the off-gas system have been distinguished. There are spontaneous helium releases as well as helium releases caused by changed conditions in the reactor (power reduction, control rod movement, etc.). Helium peaks can also be characterized according to the released amount of helium, the peak shape and the duration of the release, which leads to different interpretations of the release mechanisms. In addition, the measured amount of released helium from a 50 days period (280 l) is also compared to the calculated amount of produced helium from the washed out boron during the same time period (190 l).

  16. A rule-based expert system for control rod pattern of boiling water reactors by hovering around haling exposure shape

    International Nuclear Information System (INIS)

    Feasible strategies for automatic BWR control rod pattern generation have been implemented in a rule-based expert system. These strategies are majorly based on a concept for which exposure distributions are hovering around the Haling exposure distribution through a cycle while radial and axial power distributions are dominantly controlled by some abstracted factors indicating the desired distributions. The system can either automatically generate expert-level control rod patterns or search for criteria-satisfied patterns originated from user's input. It has successfully been demonstrated by generating control rod patterns for the the 1775 MWth Chinshan plant in Unit I Cycle 13 alternate loading pattern and Unit 2 Cycle 8 but with longer cycle length. All rod patterns for two cycles result in all-rod-out at EOC and no violation against the four criteria. The demonstrations show that the system is considerably good in choosing initial trial rod patterns and adjusting rod patterns to satisfy the design criteria. (author)

  17. 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)

    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.)

  18. Estimation of Control Rod Worth in a VVER-1000 Reactor using DRAGON4 and DONJON4

    Directory of Open Access Journals (Sweden)

    Saadatian-derakhshandeh Farahnaz

    2014-07-01

    Full Text Available One of the main issues in safety and control systems design of power and research reactors is to prevent accidents or reduce the imposed hazard. Control rod worth plays an important role in safety and control of reactors. In this paper, we developed a justifiable approach called D4D4 to estimate the control rod worth of a VVER-1000 reactor that enables to perform the best estimate analysis and reduce the conservatism that utilize DRAGON4 and DONJON4. The results are compared with WIMS-D4/CITATION to show the effectiveness and superiority of the developed package in predicting reactivity worth of the rod and also other reactor physics parameters of the VVER-1000 reactor. The results of this study are in good agreement with the plant's FSAR.

  19. Development and application of dynamic control rod reactivity measurements methodology for LWRs

    Energy Technology Data Exchange (ETDEWEB)

    Woo, I. T.; Lee, E. K.; Sin, H. C.; Ryu, S. J.; Bae, S. M.; Park, M.K.; Lee, C. S. [Korea Electric Power Research Institute, Taejon (Korea, Republic of)

    2002-10-01

    Dynamic rod worth measurement (DRWM{sup TM}) methodology commercialized by Westinghouse Co. was successfully applied to many nuclear power plants in USA to measure the control rod worth at Low Power Physics Tests. But in Korea, to increase the nuclear power plant economy using more quick and sophisticated reactivity measurement methodology without the change of boron concentration, KEPRI has developed Dynamic Control rod Reactivity Measurement (DCRM{sup TM}) methodology that was the results of a cooperative work with KAERI except the development of core analysis codes. And KAERI recently published the preliminary results for 4 control rod worths using their own inverse kinetics code and measured detector signals. The object of this paper is to show some DCRM results for the same measured data using KEPRI tools, RAST-K and INVERSE, and introduce DCRM system that could measure top and bottom detector signals fully digitally. As a result, background and noises signals at the region of low signal strength were very important to determine the rod worth. But for now, because there was no numerical model to describe the behavior of background signals, a method reflecting the characteristics of dynamic reactivity was suggested. And for noise, traditional data averaging technique was adopted. Each static worth of 8 control assemblies well agreed with those of NDR within 15%, the requirement of Tech. Spec.

  20. Control rod calibration and reactivity effects at the IPEN/MB-01 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Pinto, Letícia Negrão; Gonnelli, Eduardo; Santos, Adimir dos [Nuclear Engineering Center, Nuclear and Energy Research Institute- IPEN/CNEN-SP, Av. Lineu Prestes 2242 - Cidade Universitária - 05508-000 - São Paulo - SP (Brazil)

    2014-11-11

    Researches that aim to improve the performance of neutron transport codes and quality of nuclear cross section databases are very important to increase the accuracy of simulations and the quality of the analysis and prediction of phenomena in the nuclear field. In this context, relevant experimental data such as reactivity worth measurements are needed. Control rods may be made of several neutron absorbing materials that are used to adjust the reactivity of the core. For the reactor operation, these experimental data are also extremely important: with them it is possible to estimate the reactivity worth by the movement of the control rod, understand the reactor response at each rod position and to operate the reactor safely. This work presents a temperature correction approach for the control rod calibration problem. It is shown the control rod calibration data of the IPEN/MB-01 reactor, the integral and differential reactivity curves and a theoretical analysis, performed by the MCNP-5 reactor physics code, developed and maintained by Los Alamos National Laboratory, using the ENDF/B-VII.0 nuclear data library.

  1. Silver-indium-cadmium control rod behaviour during a severe reactor accident

    International Nuclear Information System (INIS)

    An alloy of silver, indium and cadmium is commonly used as control rod material in pressurised water reactors (PWRs). The behaviour of this alloy has been studied in a series of experiments using an induction furnace to achieve temperatures up to 1900K. The aerosols released from overheated clad and unclad control rod samples have been characterised in both steam and inert atmospheres. Mass balance experiments have been undertaken to determine the distribution of the control rod alloy constituents following rupture of the cladding, and this work has been supported by thermogravimetric studies of silver-indium mixtures. Metallographic studies were also undertaken to assess the failure mode of the stainless steel cladding and the interaction of the molten alloy with Zircaloy. The results of this work are discussed in terms of aerosol/vapour behaviour during severe reactor accidents. (author)

  2. Experience with incomplete control rod insertion in fuel with burnup exceeding approximately 40 GWD/MTU

    Energy Technology Data Exchange (ETDEWEB)

    Kee, E. [Houston Lighting & Power Co., Wadworth, TX (United States)

    1997-01-01

    Analysis and measurement experience with fuel assemblies having incomplete control rod insertion at burnups of approximately 40 GWD/MTU is presented. Control rod motion dynamics and simplified structural analyses are presented and compared to measurement data. Fuel assembly growth measurements taken with the plant Refueling Machine Z-Tape are described and presented. Bow measurements (including plug gauging) are described and potential improvements are suggested. The measurements described and analysis performed show that sufficient guide tube bow (either from creep or yield buckling) is present in some high burnup assemblies to stop the control rods before they reach their full limit of travel. Recommendations are made that, if implemented, could improve cost performance related to testing and analysis activities.

  3. An optically sensed control rod drive system for use at the Nuclear Science Center Reactor

    International Nuclear Information System (INIS)

    The optically sensed rod drive control system, installed and modified at the NSCR is described. It has operated very well and has exhibited improved reliability over the previous system. The system has proven to give stable control rod positions, and the daily reset of the position indication serves to reduce the error between indicated and true rod position. The removal of the microswitches used for carriage up and carriage down indication in the previous system, and especially the 120 VAC motor control portion, has reduced the difficulty, time and uncertainty involved in upkeep of the system and also has removed a potentially dangerous source of personnel injury. As more operational experience is gained with this design, it is felt that other minor adjustments and logic changes may come about, but the present design of the system appears to be a successful and sufficient one

  4. Digital determination of TR-I control rod worths and time behaviour of neutron flux

    International Nuclear Information System (INIS)

    In this work, the control rod reactivity worth of the swimming-pool type reactor (TR-I) in CNAEM, Cekmece Nuclear Research and Training Centre has been measured digitally and the mean neutron lifetime has been estimated by a special miniature fission chamber with 60 nanosecond resolving time. The time behaviour of thermal neutrons is also compared with reactor control console results. (orig.)

  5. Application of the dynamic control rod reactivity measurement method to Korea standard nuclear power plants

    International Nuclear Information System (INIS)

    To measure and validate the worth of control bank or shutdown bank, the dynamic control rod reactivity measurement (DCRM) technique has been developed and applied to six cases of Low Power Physics Tests of PWRs including Korea Standard Nuclear Power plant (KSNP) based on the CE System 80 NSSS. Through the DORT results for each two ex-ore detector response and the three dimensional core transient simulations for rod movements, the key parameters of DCRM method are determined to implement into the Direct Digital Reactivity Computer System (DDRCS). A total of 9 bank worths of two KSNP plants were measured to compare with the worths of the conventional rod worth measurement method. The results show that the average error of DCRM method is nearly the same as the conventional Rod Swap and Boron Dilution Method but lower standard deviation. It takes about twenty minutes from the beginning of rod movement to final estimation of the integral static worth of a control bank. (authors)

  6. Development of spent-control rod cutting equipment by abrasive water jet

    Energy Technology Data Exchange (ETDEWEB)

    Usui, Shinichi; Komiya, Toshihiro [Kawasaki Heavy Industries Ltd., Tokyo (Japan)

    2000-11-01

    Kawasaki Heavy Industries, Ltd. developed the cutting apparatus for spent-control rods and channel boxes, which utilized Abrasive Water Jet, and delivered them to Japan Atomic Power Company, Ltd. An abrasive water jet cutting is cutting method by abrasive ejecting with very high pressurized water (300 Mpa) and has merit not affecting to the objects thermally. The cutting operation carries out remotely in underwater and ejected abrasives are collected and reused in order to decrease secondary wastes. The spent-control rods and channel boxes are divided into two or three pieces and stored in the can in layers. (author)

  7. Detailed analysis for a control rod worth of the gas turbine high temperature reactor (GTHTR300)

    International Nuclear Information System (INIS)

    GTHTR300 is composed of a simplified and economical power plant based on an inherent safe 600 MWt reactor and a nearly 50% high efficiency gas turbine power conversion cycle. GTHTR300 core consist of annular fuel region, center and outer side reflectors because of cooling it effectively in depressurized accident conditions, and all control rods are located in both side reflectors of annular core. As a thermal neutron spectrum is strongly distorted in reflector regions, an accurate calculation is especially required for the control rod worth evaluation. In this study, we applied the detailed Monte Carlo calculations of a full core model, and confirmed that our design method has enough accuracy. (author)

  8. Control Rod Pattern Planning of a BWR using Enhanced Nelder-Mead Method

    International Nuclear Information System (INIS)

    We propose a new optimization algorithm for the short-term planning of control rod patterns in an operating BWR. This algorithm is based on the enhanced Nelder-Mead simplex method in which convergence ability is improved for constrained problems in several ways. The main characteristic of this approach is it uses continuous values for the axial positions of control rods. Through calculations in an actual BWR plant, we showed that the new algorithm is effective for automation of short-term planning and reduction of the engineer's workload. (authors)

  9. Detailed analysis for a control rod worth of the gas turbine high temperature reactor (GTHTR300)

    Energy Technology Data Exchange (ETDEWEB)

    Nakata, Tetsuo; Katanishi, Shoji; Takada, Shoji; Yan, Xing; Kunitomi, Kazuhiko [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    2002-11-01

    GTHTR300 is composed of a simplified and economical power plant based on an inherent safe 600 MWt reactor and a nearly 50% high efficiency gas turbine power conversion cycle. GTHTR300 core consist of annular fuel region, center and outer side reflectors because of cooling it effectively in depressurized accident conditions, and all control rods are located in both side reflectors of annular core. As a thermal neutron spectrum is strongly distorted in reflector regions, an accurate calculation is especially required for the control rod worth evaluation. In this study, we applied the detailed Monte Carlo calculations of a full core model, and confirmed that our design method has enough accuracy. (author)

  10. Degradation in steam of 60 cm-long B4C control rods

    Science.gov (United States)

    Dominguez, C.; Drouan, D.

    2014-08-01

    In the framework of nuclear reactor core meltdown accident studies, the degradation of boron carbide control rod segments exposed to argon/steam atmospheres was investigated up to about 2000 °C in IRSN laboratories. The sequence of the phenomena involved in the degradation has been found to take place as expected. Nevertheless, the ZrO2 oxide layer formed on the outer surface of the guide tube was very protective, significantly delaying and limiting the guide tube failure and therefore the boron carbide pellet oxidation. Contrary to what was expected, the presence of the control rod decreases the hydrogen release instead of increasing it by additional oxidation of boron compounds. Boron contents up to 20 wt.% were measured in metallic mixtures formed during degradation. It was observed that these metallic melts are able to attack the surrounding fuel rods, which could have consequences on fuel degradation and fission product release kinetics during severe accidents.

  11. Improving flux tilt control while adjuster control rods are removed from the Pickering NGS A reactor

    International Nuclear Information System (INIS)

    Removal of adjuster control rods from the Pickering NGS A reactor core results in flux peaking and higher fuel powers in the centre region of the core. The present flux tilt control algorithm increases the level of the light water neutron absorber in the centre liquid zone controllers in an attempt to nullify flux peaking. However, due to the limited depth of the neutron absorption capability of the liquid zone controllers, the pre-removal zone powers can not be achieved. This results in saturation of liquid zone controller levels and reduced flux tilt control. Recent operating experience as shown that in certain situations the reduced flux tilt control capability with adjusters removed results in uncorrected side to side azimuthal flux tilts. To increase tilt control in these situations an improved flux tilt control algorithm has been developed which switches the zone power flux tilt control targets to more realistic obtainable values as adjusters are removed. In this paper the computer simulations and analysis performed to develop and test the improved flux tilt algorithm is described. Also the improved performance of the new algorithm in one event will be demonstrated. 2 refs., 9 figs

  12. Effectiveness of a Large Number of Control Rods in the Second Charge of the Halden Boiling Water Reactor

    International Nuclear Information System (INIS)

    The reactivity worth of various control-rod configurations has been measured in the second fuel charge of the Halden Boiling Heavy Water Reactor (HBWR) under low power conditions. The second fuel charge of HBWR consists of 7-rod UO2 cluster elements with 1.5% enrichment. A total of 30 control rods is placed in the open positions of the hexagonal fuel-lattice structure. In older to facilitate theoretical comparisons, measurements have been made on symmetrical control-rod configurations only. The experiment consisted of measuring the critical water level for the clean core and with the different rod configurations inserted to various distances from the bottom of the reactor. The temperature dependence of the reactivity worth was investigated by performing measurements, using a ring of 6 control rods, at the three different temperatures 34°C, 150°C and 220°C. Comparisons of the experimentally-determined critical water levels and the calculated critical water levels are presented. The critical water levels are calculated both by a method in which the control rods are homogenized together with fuel and moderator to form a control-rod zone, and also by a heterogeneous method in which the fuel elements and control rods are regarded as line sinks to thermal neutrons and the fuel elements are regarded as line sources of fast neutrons. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Itagaki, Masafumi; Miyoshi, Yoshinori (Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment); Gakuhari, Kazuhiko; Okada, Noboru; Sakai, Tomohiro

    1993-05-01

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

  14. EB welding and quality control of nuclear reactor fuel rods at ASEA-ATOM

    International Nuclear Information System (INIS)

    Fourteen years ago ASEA-ATOM chose EB welding for fuel rod plug/tube welds. This choice was made on the basis of 7 years of experience of EB-welding of fuel rods in a pilot plant. The specific reasons were the high quality and the high process yield, which are made possible by the great degree of controlability and reproducibility of this process and because the welds are suitable for QC inspection by an inline ultrasonic method which we developed at the same time. To date ASEA-ATOM has manufactured approximately 600,000 fuel rods with 1,200,000 EB-welds. The results have met expections as regards quality, process yield and service in BWR and PWR reactors. Descriptions are given of the automatic Sciaky EB welding machines, of the ultrasonic inspection equipment and of their process qualification. Some comments are made on quality and process yield

  15. Control of vortex breakdown in a closed cylinder with a small rotating rod

    DEFF Research Database (Denmark)

    Lo Jacono, D.; Sørensen, Jens Nørkær; Thompson, M.C.;

    2008-01-01

    Effective control of vortex breakdown in a cylinder with a rotating lid was achieved with small rotating rods positioned on the stationary lid. After validation with accurate measurements using a novel stereoscopic particle image velocimetry (SPIV) technique, analysis of numerical simulations using...

  16. An Analytical Study of Fuzzy Control of a Flexible Rod Mechanism

    Science.gov (United States)

    Beale, D.; Lee, S. W.; Boghiu, D.

    1998-02-01

    The non-linear nature of very high speed, flexible rod mechanisms has been previously confirmed, both experimentally and analytically in reference [1]. Therefore, effective control system design for flexible mechanisms operating at very high speeds must consider the non-linearities when designing a controller for very high speeds. Active control via fuzzy logic is assessed as means to suppress the elastic transverse bending vibration of a flexible rod of a slider crank mechanism. Several pairs of piezoelectric elements are used to provide the control action. Sensor output of deflection is fed to the fuzzy controller, which determines the voltage input to the actuators. A three mode approximation is used in the simulation study. Computer simulation shows that fuzzy control can be used to suppress bending vibrations at high speeds, and even at speeds where the uncontrolled response would be unstable.

  17. A Novel Control-rod Drive Mechanism via Electromagnetic Levitation in MNSR

    OpenAIRE

    Divandari Mohammad; Hashemi-Tilehnoee Mehdi; Khaleghi Masoud; Hosseinkhah Mohammadreza

    2014-01-01

    In this paper, an electromagnetic levitation system was used with a synchronous motor to navigate the control rod of a small-type research reactor. The result from this prototype magnetic levitation system was in agreement with simulation results. The control system was programmed in MATLAB through open-loop system, closed-loop with state feedback and closed-loop with state feedback integral tracking. The final control system showed the highest performance with a low positioning error. Our re...

  18. Application of Feedforward and Feedback Control Strategy for Wire Rod Hot Rolling Line

    Institute of Scientific and Technical Information of China (English)

    ZHU Hong-xiang; HAO Xiao-hong; WEN Zhi; HU Ze-qiang; ZHANG Yao-gen; CHEN Hu-qiu

    2005-01-01

    The feedforward and feedback control strategy of water flowrate based on the analysis of thermal process in water cooling box was proposed, and the control strategy was applied to wire rod hot rolling at Baosteel Co.The operation has proved that the strategy can control water flowrate in the cooling water box reasonably to ensure the temperature requirement of the wire discharged from the cooling water box.

  19. Control Rod Drive Mechanism Installed in the Internal of Reactor Pressure Vessel

    Energy Technology Data Exchange (ETDEWEB)

    Choi, M. H.; Choi, S.; Park, J. S.; Lee, J. S.; Kim, D. O.; Hur, N. S.; Hur, H.; Yu, J. Y

    2008-09-15

    This report describes the review results and important technologies related to the in-vessel type control rod drive mechanism. Generally, most of the CRDMs used in the PWR are attached outside of the reactor pressure vessel, and the pernetration of the vessel head can not avoid. However, in-vessel type CRDMs, which are installed inside the reactor vessel, can eliminate the possibility of rod ejection accidents and the penetration of the vessel head, and provide a compact design of the reactor vessel and containment. There are two kinds of in-vessel type CRDM concerning the driving force-driven by a driving motor and by a hydraulic force. Motor driven CRDMs have been mainly investigated in Japan(MRX, IMR, DRX, next generation BWR etc.), and developed the key components such as a canned motor, an integrated rod position indicator, a separating ball-nut and a ball bearing that can operate under the water conditions of a high temperature and pressure. The concept of hydraulically driven CRDMs have been first reported by KWU and Siemens for KWU 200 reactor, and Argentina(CAREM) and China(NHR-5, NHR-200) have been developed the internal CRDM with the piston and cylinder of slightly different geometries. These systems are driven by the hydraulic force which is produced by pumps outside of the reactor vessel and transmitted through a pipe penetrating the reactor vessel, and needs complicated control and piping systems including pumps, valves and pipes etc.. IRIS has been recently decided the internal CRDMs as the reference design, and an analytical and experimental investigations of the hydraulic drive concept are performed by POLIMI in Italy. Also, a small French company, MP98 has been developed a new type of control rods, called 'liquid control rods', where reactivity is controlled by the movement of a liquid absorber in a manometer type device.

  20. Control Rod Drive Mechanism Installed in the Internal of Reactor Pressure Vessel

    International Nuclear Information System (INIS)

    This report describes the review results and important technologies related to the in-vessel type control rod drive mechanism. Generally, most of the CRDMs used in the PWR are attached outside of the reactor pressure vessel, and the pernetration of the vessel head can not avoid. However, in-vessel type CRDMs, which are installed inside the reactor vessel, can eliminate the possibility of rod ejection accidents and the penetration of the vessel head, and provide a compact design of the reactor vessel and containment. There are two kinds of in-vessel type CRDM concerning the driving force-driven by a driving motor and by a hydraulic force. Motor driven CRDMs have been mainly investigated in Japan(MRX, IMR, DRX, next generation BWR etc.), and developed the key components such as a canned motor, an integrated rod position indicator, a separating ball-nut and a ball bearing that can operate under the water conditions of a high temperature and pressure. The concept of hydraulically driven CRDMs have been first reported by KWU and Siemens for KWU 200 reactor, and Argentina(CAREM) and China(NHR-5, NHR-200) have been developed the internal CRDM with the piston and cylinder of slightly different geometries. These systems are driven by the hydraulic force which is produced by pumps outside of the reactor vessel and transmitted through a pipe penetrating the reactor vessel, and needs complicated control and piping systems including pumps, valves and pipes etc.. IRIS has been recently decided the internal CRDMs as the reference design, and an analytical and experimental investigations of the hydraulic drive concept are performed by POLIMI in Italy. Also, a small French company, MP98 has been developed a new type of control rods, called 'liquid control rods', where reactivity is controlled by the movement of a liquid absorber in a manometer type device

  1. Determination of power peak factor using control rods, ex-core detectors and neural networks

    International Nuclear Information System (INIS)

    This work presents a methodology based on the artificial neural network technique to predict in real time the power peak factor in a form that can be implemented in reactor protection systems. The neural network inputs were those available in the reactor protection systems, namely, the axial and quadrant power differences obtained from measured ex-core detector signals, and the position of control rods. The response of ex core detector signals was measured in experiments especially performed in the IPEN/MB-01 zero-power reactor. Several reactor states with different power density distribution were obtained by positioning the control rods in different configurations. The power distribution and its peak factor were calculated for each of these reactor states using the Citation code. The obtained results show that the power peak factor correlates well with the control rod position and the quadrant power difference, and with a lesser degree with the axial power differences. The data presented an inherent organisation and could be classified into different classes of power peak factor behaviour as a function of position of control rods, axial power difference and quadrant power difference. The RBF networks were able to identify classes and interpolate the power peak factor values. The relative error for the power peak factor estimation ranged from 0.19 % to 0.67 %, less than the one that was obtained performing a power density distribution map with in-core detectors. It was observed that the positions of control rods bear the detailed and localised information about the power density distribution, and that the axial and the quadrant power difference describe its global variations in the axial and radial directions. The results showed that the RBF and MLP networks produced similar results, and that a neural network correlation can be implemented in power reactor protection systems. (author)

  2. Blackness coefficients, effective diffusion parameters, and control rod worths for thermal reactors - methods

    Energy Technology Data Exchange (ETDEWEB)

    Bretscher, M.M.

    1984-01-01

    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 P/sub 1/, P/sub 3/, and P/sub 5/ 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.

  3. Development and control of the process for the manufacture of zircaloy-4 tubing for LWBR fuel rods

    International Nuclear Information System (INIS)

    The technical requirements for the Light Water Breeder Reactor (LWBR) fuel elements (fuel rods) imposed certain unique requirements for the low hafnium Zircaloy-4 tubing used as fuel rod cladding. This report describes, in detail, the tube manufacturing process, the product and process controls used, the inspections and tests performed, and the efforts involved in refining a commercial tube reducing process to produce tubes that would satisfy the requirements for LWBR fuel rod cladding

  4. High Temperature Electromechanical Components for Control Rod Drive Assemblies

    Science.gov (United States)

    Gleason, Thomas E.; Lazarus, Jonathan D.; Yaspo, Robert; Cole, Allan R.; Otwell, Robert L.; Schuster, Gary B.; Jaing, Thomas J.; Meyer, Raymond A.; Shukla, Jaikaran N.; Maldonado, Jerry

    1994-07-01

    The SP-100 power system converts heat generated within a compact fast spectrum nuclear reactor directly to electricity for spacecraft applications. The reactor control system contains the only moving mechanical and electromechanical components in the entire electrical generating system. The high temperature, vacuum environment presents unique challenges for these reactor control system components. This paper describes the environmental testing of these components that has been completed and that is in progress. The specific components and assemblies include electromagnetic (EM) coils, stepper motors, EM clutches, EM brakes, ball bearings, ball screw assemblies, constant torque spring motors, gear sets, position sensors, and very high temperature sliding bearings.

  5. Searching for full power control rod patterns in a boiling water reactor using genetic algorithms

    Energy Technology Data Exchange (ETDEWEB)

    Montes, Jose Luis [Departamento Sistemas Nucleares, ININ, Carr. Mexico-Toluca Km. 36.5, Ocoyoacac, Edo. de Mexico (Mexico)]. E-mail: jlmt@nuclear.inin.mx; Ortiz, Juan Jose [Departamento Sistemas Nucleares, ININ, Carr. Mexico-Toluca Km. 36.5, Ocoyoacac, Edo. de Mexico (Mexico)]. E-mail: jjortiz@nuclear.inin.mx; Requena, Ignacio [Departamento Ciencias Computacion e I.A. ETSII, Informatica, Universidad de Granada, C. Daniel Saucedo Aranda s/n. 18071 Granada (Spain)]. E-mail: requena@decsai.ugr.es; Perusquia, Raul [Departamento Sistemas Nucleares, ININ, Carr. Mexico-Toluca Km. 36.5, Ocoyoacac, Edo. de Mexico (Mexico)]. E-mail: rpc@nuclear.inin.mx

    2004-11-01

    One of the most important questions related to both safety and economic aspects in a nuclear power reactor operation, is without any doubt its reactivity control. During normal operation of a boiling water reactor, the reactivity control of its core is strongly determined by control rods patterns efficiency. In this paper, GACRP system is proposed based on the concepts of genetic algorithms for full power control rod patterns search. This system was carried out using LVNPP transition cycle characteristics, being applied too to an equilibrium cycle. Several operation scenarios, including core water flow variation throughout the cycle and different target axial power distributions, are considered. Genetic algorithm fitness function includes reactor security parameters, such as MLHGR, MCPR, reactor k{sub eff} and axial power density.

  6. Application of Heterogeneous and Homogeneous Methods in the Calculation of Control-Rod Effects in D2O Lattices

    International Nuclear Information System (INIS)

    The application of heterogeneous and homogeneous calculation methods in the determination of control-rod effects in natural-uranium and heavy-water-moderated cores is discussed with reference to experiments performed in the Swedish RO. reactor. The experiments, involving the determination of the reactivity effects of both fully.and partially inserted absorber rods in different lattices, are used for comparison of the results of calculations in which (a) the individual control and fuel rods are treated by source-sink theory, and (b) the medium surrounding the control rods is treated as homogeneous. The agreements between the results from these theoretical treatments and the accuracy with which they predict the control-rod reactivity effects in heavy-water lattices are discussed. (author)

  7. Simulation and operation of the EBR-2 automatic control rod drive system

    Science.gov (United States)

    Lehto, W. K.; Larson, H. A.; Dean, E. M.; Christensen, L. J.

    An automatic control rod drive system (ACRDS) installed at EBR-II produces shaped power transients from 40% to full reactor power at a linear ramp rate of 4 MWt/s. A digital computer and modified control rod drive provides this capability. Simulation and analysis of ACRDS experiments establish the safety envelope for reactor transient operation. Tailored transients are required as part of USDOE operational reliability testing program for prototypic fast reactor fuel cladding breach behavior studies. After initial EBR-II driver fuel testing and system checkout, test subassemblies were subjected to both slow and fast transients. In addition, the ACRDS is used for steady state operation and will be qualified to control power ascent from initial critical to full power.

  8. State of Art of the CAREM-25 Hydraulic Control Rod Drives Feasibility Analysis

    International Nuclear Information System (INIS)

    The proposed design adopted for the control rod drives for the CAREM reactor is based on a hydraulic system.As any innovative device, the design process requires to obtain experimental evidence to identify the most important control parameters and to set their relationship with other design parameters, in order to guarantee its feasibility as a previous step to the design qualification tests at the working conditions at the reactor.This paper features a global evaluation of the analysis performed and experimental results obtained in a low pressure loop, design improvements, limiting phenomena identified and corrective actions analyzed or proposed.The evaluation is based on a repetitivity, sensitivity and scalability study of the control parameters and test conditions, as well as the dynamic response between rod drive and the hydraulic system and features related with the mechanical design.Obtained results show that present system has an adequate response compatible with functional and manufacturing requirements

  9. Stereotypical cell division orientation controls neural rod midline formation in zebrafish.

    Science.gov (United States)

    Quesada-Hernández, Elena; Caneparo, Luca; Schneider, Sylvia; Winkler, Sylke; Liebling, Michael; Fraser, Scott E; Heisenberg, Carl-Philipp

    2010-11-01

    The development of multicellular organisms is dependent on the tight coordination between tissue growth and morphogenesis. The stereotypical orientation of cell divisions has been proposed to be a fundamental mechanism by which proliferating and growing tissues take shape. However, the actual contribution of stereotypical division orientation (SDO) to tissue morphogenesis is unclear. In zebrafish, cell divisions with stereotypical orientation have been implicated in both body-axis elongation and neural rod formation, although there is little direct evidence for a critical function of SDO in either of these processes. Here we show that SDO is required for formation of the neural rod midline during neurulation but dispensable for elongation of the body axis during gastrulation. Our data indicate that SDO during both gastrulation and neurulation is dependent on the noncanonical Wnt receptor Frizzled 7 (Fz7) and that interfering with cell division orientation leads to severe defects in neural rod midline formation but not body-axis elongation. These findings suggest a novel function for Fz7-controlled cell division orientation in neural rod midline formation during neurulation.

  10. Magnetic Actuation Connector Between Extension Shaft and Armature for Bottom Mounted Control Rod Drive Mechanism

    Energy Technology Data Exchange (ETDEWEB)

    Huh, Hyung; Cho, Yeong Garp; Kim, Jong In [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    The electromagnet and armature inside the guide tube interact and produce magnetism, thus making the armature, connecting extension shaft and control rod move up and down to control the power of reactor. During the overhaul, the control absorber rod (CAR), extension shaft, and armature of BMCRDM are lifted together for closing a seal valve. But total length of CAR assembly is so long that it cannot be lifted due to exposure above the water level of pool which is strictly controlled. In addition to this, it is difficult to calibrate a position indicator and lifting force of electromagnet without armature assembly as a seal valve is closed. For this reason, it is necessary to install a disconnecting system between armature and extension shaft. Therefore, KAERI has developed magnetic actuation connector using plunger between armature and extension shaft for the bottom mounted control rod drive mechanism in research reactor. The results of a FEM and the experiments in this work lead to the following conclusions: The FEM result for the design of the magnetic actuation connector is compared with the measured lifting force of prototype production. As a result, it is shown that the lifting force of the prototype connector has a good agreement with the result of the FEM. A newly developed technique of prototype magnetic actuation connector which is designed by FEM analysis result is proposed.

  11. Magnetic Actuation Connector Between Extension Shaft and Armature for Bottom Mounted Control Rod Drive Mechanism

    International Nuclear Information System (INIS)

    The electromagnet and armature inside the guide tube interact and produce magnetism, thus making the armature, connecting extension shaft and control rod move up and down to control the power of reactor. During the overhaul, the control absorber rod (CAR), extension shaft, and armature of BMCRDM are lifted together for closing a seal valve. But total length of CAR assembly is so long that it cannot be lifted due to exposure above the water level of pool which is strictly controlled. In addition to this, it is difficult to calibrate a position indicator and lifting force of electromagnet without armature assembly as a seal valve is closed. For this reason, it is necessary to install a disconnecting system between armature and extension shaft. Therefore, KAERI has developed magnetic actuation connector using plunger between armature and extension shaft for the bottom mounted control rod drive mechanism in research reactor. The results of a FEM and the experiments in this work lead to the following conclusions: The FEM result for the design of the magnetic actuation connector is compared with the measured lifting force of prototype production. As a result, it is shown that the lifting force of the prototype connector has a good agreement with the result of the FEM. A newly developed technique of prototype magnetic actuation connector which is designed by FEM analysis result is proposed

  12. Physics Analysis of a Prismatic VHTR with Asymmetric Control Rods by Using the HELIOS/MASTER Code Package

    International Nuclear Information System (INIS)

    A new physics analysis procedure is under development for prismatic VHTRs based on a conventional two-step procedure for a PWR physics analysis. The HELIOS and MASTER codes were employed to generate the coarse group cross sections through a transport lattice calculation, and to perform the 3-dimensional core physics analysis by a nodal diffusion calculation, respectively. Since prismatic VHTRs such as a GT-MHR include asymmetrically located large control rods, a control rod treatment is a challenging issue in a physics analysis. Previously, we performed a physics analysis for a prismatic VHTR in which symmetric control rods were assumed. Large spectrum shifts due to a control rod insertion on the surrounding blocks could be covered by optimizing the coarse energy group structure. However, it was noted that some improvements should be made in the prediction of the reaction rates and the control rod worths. In this study a new analysis procedure has been developed to deal with asymmetric control rods more accurately. Surface dependent discontinuity factors obtained from multi-block models were applied to the core calculations for a better prediction of the reaction rates and control rod worths. Benchmark calculations were performed for the GT-MHR cores, where the reference solutions were obtained from the MCNP calculations

  13. Measurement of the neutron flux distribution in graphite-moderated core SHE-8 inserted with experimental control rods

    International Nuclear Information System (INIS)

    A very-high temperature gas cooled reactor is so designed to attain the desired outlet coolant gas temperature under the limiting maximum temperature of fuel elements. Optimization in the spatial power distribution is thus necessary by suitable arrangement of the control rods and fuel exchange program. In this respect, high accuracy in calculation of the spatial power distribution is required. In this report, experiment and calculation are compared for a 20% enriched uranium loaded and graphite moderated core, SHE-8. Induced activity of the copper pins were measured for the following four core configurations: Case 1 : standard core without control rods Case 2 : a single control rod inserted along the core axis Case 3 : a single control rod inserted off the core axis Case 4 : two control rods inserted symmetrically along the core axis. The experimental control rods used are the pellets of a cold pressed homogeneous mixture of carbon and B4C powders contained in thin-walled aluminum tubes. The diameter of the experimental control rods and their B4C content are the same as in the preliminary core design of UHTGR by JAERI. Calculation of the neutron flux distribution was made by the three-dimensional two-group source-sink method. Agreement beween experiment and calculation is fairly good, so the axial peaking factor can be estimated within the error of 1--3%. Discrepancies in the radial peaking factor are large, however, about 5%. (auth.)

  14. Feasibility study of the University of Utah TRIGA reactor power upgrade in respect to control rod system

    Science.gov (United States)

    Cutic, Avdo

    The objectives of this thesis are twofold: to determine the highest achievable power levels of the current University of Utah TRIG Reactor (UUTR) core configuration with the existing three control rods, and to design the core for higher reactor power by optimizing the control rod worth. For the current core configuration, the maximum reactor power, eigenvalue keff, shutdown margin, and excess reactivity have been measured and calculated. These calculated estimates resulted from thermal power calibrations, and the control rod worth measurements at various power levels. The results were then used as a benchmark to verify the MCNP5 core simulations for the current core and then to design a core for higher reactor power. This study showed that the maximum achievable power with the current core configuration and control rod system is 150kW, which is 50kW higher than the licensed power of the UUTR. The maximum achievable UUTR core power with the existing fuel is determined by optimizing the core configuration and control rod worth, showing that a power upgrade of 500 kW is achievable. However, it requires a new control rod system consisting of a total of four control rods. The cost of such an upgrade is $115,000.

  15. Interaction between silver-indium-cadmium control-rod alloy and Zircaloy at high temperatures

    International Nuclear Information System (INIS)

    In order to investigate the reactivity between control-rod (silver-indium-cadmium) alloy and Zircaloy during a severe accident of a pressurized water reactor, reaction couples of control-rod alloy and Zircaloy-4 were isothermally heated in argon at the temperatures ranging from 1273 to 1473K. The reaction rate increased with the temperature increase. About 1mm decrease in Zircaloy thickness was measured in the sample heated at 1473K for 60s. The reaction roughly obeyed a parabolic law, thereby the reaction rate constants and the apparent activation energy for the reaction, about 323kJ/mol, were determined. The microstructure and elemental distribution in reacted zones of samples was examined with an optical microscopy and an EPMA. (author)

  16. Control rod absorber section fabrication by square tube configuration and dual laser welding process

    International Nuclear Information System (INIS)

    This patent describes a process for the assembly of a planar section of a cruciform shaped control rod from tubes. It comprises: providing tubes, the tubes having cylindrical interior volumes for the containment of neutron absorbing poisons and having square external sections for being joined by welding in side-by-side relation; filling the cylindrical interior volumes with neutron absorbing poisons; plugging the tubes to seal the neutron absorbing poisons within the tubes: providing a jig for maintaining the tubes in side-by-side relation to form a planar section of the control rod, the jig having a leading end for holding the ends of the tubes in side-by-side relation and having a trailing end for holding the tubes in side-by-side relation

  17. Analysis of Reactivity Induced Accident for Control Rods Ejection with Loss of Cooling

    CERN Document Server

    Saad, Hend Mohammed El Sayed; Wahab, Moustafa Aziz Abd El

    2013-01-01

    Understanding of the time-dependent behavior of the neutron population in nuclear reactor in response to either a planned or unplanned change in the reactor conditions, is a great importance to the safe and reliable operation of the reactor. In the present work, the point kinetics equations are solved numerically using stiffness confinement method (SCM). The solution is applied to the kinetics equations in the presence of different types of reactivities and is compared with different analytical solutions. This method is also used to analyze reactivity induced accidents in two reactors. The first reactor is fueled by uranium and the second is fueled by plutonium. This analysis presents the effect of negative temperature feedback with the addition positive reactivity of control rods to overcome the occurrence of control rod ejection accident and damaging of the reactor. Both power and temperature pulse following the reactivity- initiated accidents are calculated. The results are compared with previous works and...

  18. A Novel Control-rod Drive Mechanism via Electromagnetic Levitation in MNSR

    Directory of Open Access Journals (Sweden)

    Divandari Mohammad

    2014-07-01

    Full Text Available In this paper, an electromagnetic levitation system was used with a synchronous motor to navigate the control rod of a small-type research reactor. The result from this prototype magnetic levitation system was in agreement with simulation results. The control system was programmed in MATLAB through open-loop system, closed-loop with state feedback and closed-loop with state feedback integral tracking. The final control system showed the highest performance with a low positioning error. Our results showed that the developed control system has the potential to be used as a reliable actuator in nuclear reactors to satisfy higher performance and safety.

  19. Assessing the Effect of Fuel Burnup on Control Rod Worth for HEU and LEU Cores of Gharr-1

    Directory of Open Access Journals (Sweden)

    E.K. Boafo

    2013-02-01

    Full Text Available An important parameter in the design and analysis of a nuclear reactor is the reactivity worth of the control rod which is a measure of the efficiency of the control rod to absorb excess reactivity. During reactor operation, the control rod worth is affected by factors such as the fuel burnup, Xenon concentration, Samarium concentration and the position of the control rod in the core. This study investigates the effect of fuel burnup on the control rod worth by comparing results of a fresh and an irradiated core of Ghana's Miniature Neutron Source Reactor for both HEU and LEU cores. In this study, two codes have been utilized namely BURNPRO for fuel burnup calculation and MCNP5 which uses densities of actinides of the irradiated fuel obtained from BURNPRO. Results showed a decrease of the control rod worth with burnup for the LEU while rod worth increased with burnup for the HEU core. The average thermal flux in both inner and outer irradiation sites also decreased significantly with burnup for both cores.

  20. Girth welding of control rod by pulse tungsten-inert-gas welding process

    International Nuclear Information System (INIS)

    The author studies the features of pulse tungsten-inert-gas (TIG) welding process in girth welding of the control rod for Pakistan Chashma (PC) Nuclear Power Plant and the availability of 0Cr18Ni11Ti, the cold-finished austenitic stainless steel material. It analyzes the reasons that the crack and the incomplete-fusion occur in the girth weld, presents the actions to be taken

  1. Peculiarity by Modeling of the Control Rod Movement by the Kalinin-3 Benchmark

    International Nuclear Information System (INIS)

    The paper presents an important part of the results of the OECD/NEA benchmark transient 'Switching off one main circulation pump at nominal power' analyzed as a boundary condition problem by the coupled system code ATHLET-BIPR-VVER. Some observations and comparisons with measured data for integral reactor parameters are discussed. Special attention is paid on the modeling and comparisons performed for the control rod movement and the reactor power history. (Authors)

  2. On Line Measurement of Reactivity Worth of TRIGA Mark-II Research Reactor Control Rods

    OpenAIRE

    Nusrat Jahan; Mamunur M. Rashid; F. Ahmed; M. G. S. Islam; M. Aliuzzaman; Islam, S.M.A

    2011-01-01

    The reactivity worth measurement system for control rods of the TRIGA MARK-II research reactor of Bangladesh has been design and developed. The theory of the kinetic technique of measuring reactivity has been used by this measurement system. The system comprises of indigenous hardware and software for online acquisition of neutron flux signals from reactor console and then computes the reactivity worth accordingly. Here for the TRIGA MARK-II research reactor, the reactivity measurement system...

  3. Computer simulation on the controlled cooling of 82B high-speed rod

    Institute of Scientific and Technical Information of China (English)

    Jinqiao Xu; Yazheng Liu; Shumei Zhou

    2008-01-01

    A modified temperature-phase transformation field coupled nonlinear mathematical model was made and used in com-puter simulation on the controlled cooling of 82B high-speed rods. The surface temperature history and volume fraction of pearlite as well as the phase transformation history were simulated by using the finite element software Marc/Mentat. The simulated results were compared with the actual measurement and the agreement is good which can validate the presented computational models.

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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

  6. Development of eddy current testing technique of the rod cluster control assembly of pressurized water reactor

    International Nuclear Information System (INIS)

    Rod Control Cluster Assembly(RCCA) of pressurized water reactor(PWR) can be damaged by neutron irradiation and continuous vibration caused by pressurized water flowing with a high speed within the reactor. Typically, there are three different types of RCCA damage: (1) Fretting wear caused by interactions of the control rod with upper internal guide cards, (2) Sliding wear caused by the up-and-down sliding movement of the control rod during the operation, and (3) Intergranular cracking caused by the material embrittlement stemming from neutron irradiation. In the past, either ultrasonics or Eddy current testing(ECT) methods were used to inspect RCCAs. However, due to inconvenient and tedious operation of ultrasonic method, Eddy current testing method is being used more frequently. Nondestructive Evaluation(NDE) group of the Materials and Corrosion Research Laboratory at KEPRI has recently developed ECT method and the associated testing equipment, and applied successfully to Ulchin Unit 1 and Kori Unit 2 nuclear power plants(NPPs) during the overhaul period. This paper summarizes the results of the ECT of RCCAs.

  7. Analysis on Electromagnetic Characteristics of Research Reactor Control Rod Drive Mechanism for Thrust Force Improvement

    Energy Technology Data Exchange (ETDEWEB)

    Huh, Hyung; Choi, Myoung Hwan; Yu, Je Yong; Cho, Yeong Garp; Kim, Jong In [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-10-15

    The control rod drive mechanism (CRDM) is the part of reactor regulating system (RRS), which is located in the reactor pool top or the room below the reactor pool. The function of the CRDM is to insert, withdraw or maintain neutron absorbing material (control rod) at any required position within the reactor core, in order to the reactivity of the core. There are so many kinds of CRDM, such as magnetic-jack type, hydraulic type, rack and pinion type, chain type and linear or rotary step motor and so on. As a part of a new project, we are investigating the movable coil electromagnetic drive mechanism (MCEDM) which is new scheme for the reactor control rod adopted by China Advanced Research Reactor (CARR). To have a better knowledge of the electromagnetic and magnetic characteristics, numerical models of MCEDM are proposed. Especially in order to achieve improved thrust force, numerical magnetic field calculations for various kinds of magnetic and electromagnetic configuration have been performed. As a result, we present the improved design of MCEDM for research reactor

  8. Sensory systems for a control rod position using reed switches for the integral reactor

    International Nuclear Information System (INIS)

    The system-integrated modular advanced reactor (SMART) currently under development at the Korea Atomic Energy Research Institute is being designed with a soluble boron free operation and the use of nuclear heating for the reactor start-up. These design features require a Control Element Drive Mechanism (CEDM) for the SMART to have a fine-step movement capability as well as a high reliability for a fine reactivity control. Also the reliability and accuracy of the information for the control rod position is very important to the reactor safety as well as the design of the core protection system. The position indicator is classified as a Class 1E component because the rod position signal of the position indicator is used in the safety related systems. Therefore it will be separated from the control systems to the extent that a failure of any single control system component of a channel and shall have sufficient independence, redundancy, and testability to perform its safety functions assuming a single failure. The position indicator is composed of a permanent magnet, reed switches and a voltage divider. Four independent position indicators around the upper pressure housing provide an indication of the position of a control rod comprising of a permanent magnet with a magnetic field concentrator which moves with the extension shaft connected to the control rod. The zigzag arranged reed switches are positioned along a line parallel to the path of the movement of the permanent magnet and it is activated selectively when the permanent magnet passes by. A voltage divider electrically connected to the reed switches provides a signal commensurate with the position of the control rod. The signal may then be transmitted to a position indicating device. In order to monitor the operating condition of the rotary step motor of CEDM, the angular position detector was installed at the top of the rotary step motor by means of connecting between the planetary gear and the rotating

  9. Design of Seismic Test Rig for Control Rod Drive Mechanism of Jordan Research and Training Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sun, Jongoh; Kim, Gyeongho; Yoo, Yeonsik; Cho, Yeonggarp; Kim, Jong In [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    The reactor assembly is submerged in a reactor pool filled with water and its reactivity is controlled by locations of four control absorber rods(CARs) inside the reactor assembly. Each CAR is driven by a stepping motor installed at the top of the reactor pool and they are connected to each other by a tie rod and an electromagnet. The CARs scram the reactor by de-energizing the electromagnet in the event of a safe shutdown earthquake(SSE). Therefore, the safety function of the control rod drive mechanism(CRDM) which consists of a drive assembly, tie rod and CARs is to drop the CAR into the core within an appropriate time in case of the SSE. As well known, the operability for complex equipment such as the CRDM during an earthquake is very hard to be demonstrated by analysis and should be verified through tests. One of them simulates the reactor assembly and the guide tube of the CAR, and the other one does the pool wall where the drive assembly is installed. In this paper, design of the latter test rig and how the test is performed are presented. Initial design of the seismic test rig and excitation table had its first natural frequency at 16.3Hz and could not represent the environment where the CRDM was installed. Therefore, experimental modal analyses were performed and an FE model for the test rig and table was obtained and tuned based on the experimental results. Using the FE model, the design of the test rig and table was modified in order to have higher natural frequency than the cutoff frequency. The goal was achieved by changing its center of gravity and the stiffness of its sliding bearings.

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

    International Nuclear Information System (INIS)

    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 keff 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

  11. Rod cluster control assemblies and rod cluster control guide tubes: wear and drop time; Grappes de commande et guides de grappes: usure et tempes de chute

    Energy Technology Data Exchange (ETDEWEB)

    Zbinden, M. [Electricite de France (EDF), Direction des Etudes et Recherches, 92 - Clamart (France)

    1997-12-31

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

  12. Monte Carlo simulation of a research reactor with nominal power of 7 MW to design new control safety rods

    Science.gov (United States)

    Shoushtari, M. K.; Kakavand, T.; Sadat Kiai, S. M.; Ghaforian, H.

    2010-03-01

    The Monte Carlo simulation has been established for a research reactor with nominal power of 7 MW. A detailed model of the reactor core was employed including standard and control fuel elements, reflectors, irradiation channels, control rods, reactor pool and thermal column. The following physical parameters of reactor core were calculated for the present LEU core: core reactivity ( ρ), control rod (CR) worth, thermal and epithermal neutron flux distributions, shutdown margin and delayed neutron fraction. Reduction of unfavorable effects of blockage probability of control safety rod (CSR)s in their interiors because of not enough space in their sites, and lack of suitable capabilities to fabricate very thin plates for CSR cladding, is the main aim of the present study. Making the absorber rod thinner and CSR cladding thicker by introducing a better blackness absorbing material and a new stainless steel alloy, respectively, are two studied ways to reduce the effects of mentioned problems.

  13. Insertion exponential of reactivity. Application to the calibration of control rods; Insercion exponencial de reactividad. Aplicacion a la calibracion de barras de control

    Energy Technology Data Exchange (ETDEWEB)

    Blazquez Martinez, J. B.; Becares, V.

    2012-07-01

    The control rods are calibrated with several steps traditionally small reactivity. An innovation in this process is continuously withdrawing the control rod from the reactor core to the periphery. Innovation is done by 10% of the time for the usual procedure, which can result in cost savings in commercial reactors.

  14. Mechanism design for the control rods conduction of TRIGA Mark III reactor in the NINR

    International Nuclear Information System (INIS)

    This work presents in the first chapter a general studio about the reactor and the importance of control rods in the reactor , the mechaniucal design attending to requisitions that are imposed for conditions of operation of the reactor are present in the second chapter, the narrow relation that exists with the new control console and the mechanism is developed in the thired chapter, this relation from a point of view of an assembly of components is presents in fourth chapter, finally reaches and perspectives of mechanism forming part of project of the automation of reactor TRIGA MARK III, are present in the fifth chapter. (Author)

  15. Research and application of an intelligent recloser controller installed on outdoor rod

    Institute of Scientific and Technical Information of China (English)

    廖力清; 陈燕辉; 凌玉华; 杨欣荣

    2002-01-01

    A new type of intelligent recolser controller installed on the outdoor rod is developed, which is mainly composed of microcontroller of Intel 87C196KC-20 and CPLD devices. This controller integrates all the functions of measuring, controlling, protection, fault diagnosis, communication, remote-controlled operation and self-power devices with infra-red remote control devices as a unit. The controller applies the distributed structure, field concentration line and intelligent technology to seal up the synthetic servomechanisms such as the microcomputer-based protection and measuring devices in the second stage of the mini out-door transformer substation, which are distributed on the outdoor circuit switches on the spot and formed as a whole. Therefore, this technology can transform a large number of ordinary homemade SF6 circuit beaker and vacuum circuit breaker into intelligent circuit recloser, thus replacing the expensive imported automatic circuit recolser.

  16. Conceptual Design on Primary Control Rod Drive Mechanism of a Prototype Gen-IV SFR

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae Han; Koo, Gyeong Hoi [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    This paper describes the key concept of the drive mechanism, and suggests a required motor power and reducer gears to meet the functional design requirements, and a seismic response analysis of CRDM housing is performed to check its structural integrity. An AC servo motor is selected as a CRA driving power because it uses permanent magnets and is brushless type while DC motor needs a brush and a coil rotates. The control shim motor size is constrained by a housing diameter of 250mm. The driving system has several design requirements. To calculate the motor power, the drive shaft torque is needed. One part of the drive shaft has a lead screw, driving by a ball-nut. The ball screw driver torque (Tr) is calculated by some equations as follow; A servo motor with a nominal power of 100W, a nominal torque of 0.32 N-m (max. 0.48N-m) is selected considering a safety margin. Its diameter is about 50mm. The fast drive-in motor needs a strong power to insert enforcedly the stuck CRA into core within a required time. The motor sizes are calculated by the same procedure. The diameters are in the range of 80mm to 110mm by the insertion time (10 ∼ 24 seconds). The prototype Gen-IV SFR (sodium-cooled Fast Reactor) is of 150MWe capacity. The reactor has six primary control rod assemblies(CRAs). The primary control rod is used for power control, burn-up compensation and reactor shutdown in response to demands from the plant control or protection systems. The control rod drive mechanism (CRDM) consists of the drive motor assembly, the driveline, and its housing. The driveline consists of three concentric members of a drive shaft, a tension tube, and a position indicator rod, and it connects the drive motor assembly to the CRA. Main issue is that these many driving parts shall be enclosed within a limited housing diameter because the available pitch of CRDMs is limited by 300mm.

  17. Analysis of the burnup of the control rods with the COREMASTER-Presto code; Analisis del quemado de barras de control con el codigo COREMASTER-PRESTO

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez, J.L.; Alonso, G.; Perusquia, R.; Montes, J.L.; Hernandez, H. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)]. e-mail: jlhm@nuclear.inin-mx

    2003-07-01

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

  18. Implementation and Tests of FPGA-embedded PowerPC in the control system of the ATLAS IBL ROD card

    CERN Document Server

    Balbi, G; The ATLAS collaboration; Falchieri, D; Gabrielli, A; Furini, M; Kugel, A; Travaglini, R; Wensing, M

    2012-01-01

    The Insertable B-layer project is planned for the upgrade of the ATLAS experiment at LHC. A silicon layer will be inserted into the existing Pixel Detector together with new electronics. The readout off-detector system is implemented with a Back-Of-Crate module implementing I/O functionality and a Readout-Driver card (ROD) for data processing. The ROD hosts the electronics devoted to control operations implemented both with a back- compatible solution (via DSP) and with a PowerPC embedded into an FPGA. In this document major firmware and software achievements concerning the PowerPC implementation, tested on ROD prototypes, will be reported.

  19. TREAT [Transient Reactor Test Facility] reactor control rod scram system simulations and testing

    International Nuclear Information System (INIS)

    Air cylinders moving heavy components (100 to 300 lbs) at high speeds (above 300 in/sec) present a formidable end-cushion-shock problem. With no speed control, the moving components can reach over 600 in/sec if the air cylinder has a 5 ft stroke. This paper presents an overview of a successful upgrade modification to an existing reactor control rod drive design using a computer model to simulate the modified system performance for system design analysis. This design uses a high speed air cylinder to rapidly insert control rods (278 lb moved 5 ft in less than 300 msec) to scram an air-cooled test reactor. Included is information about the computer models developed to simulate high-speed air cylinder operation and a unique new speed control and end cushion design. A patent application is pending with the US Patent ampersand Trade Mark Office for this system (DOE case number S-68,622). The evolution of the design, from computer simulations thru operational testing in a test stand (simulating in-reactor operating conditions) to installation and use in the reactor, is also described. 6 figs

  20. The effect of aging upon CE and B and W control rod drives

    International Nuclear Information System (INIS)

    The effect of aging upon the Babcock and Wilcox and Combustion Engineering control rod drive systems has been evaluated as part of the US Nuclear Regulatory Commission Nuclear Plant Aging Research program. Operating experience data for the 1980-1990 time period was reviewed to identify predominant failure modes, causes, and effects. These results, in conjunction with an assessment of component materials and operating environments, conclude that both systems are susceptible to age degradation. System failures have resulted in significant plant effects, including power reductions, plant shutdowns, scrams, and engineered safety feature actuation. Current industry inspection and maintenance practices were assessed. Some of these practices effectively address aging, while others do not

  1. The Domestication of a Interface Device for the HANARO Control Rod

    Energy Technology Data Exchange (ETDEWEB)

    Choe, Y. S.; Bae, S. H.; Kim, Y. K.; Jung, H. S.; Lee, J. H.; Kim, S. J.; Kang, K. D. [KAERI, Daejeon (Korea, Republic of)

    2010-12-15

    The signal processing unit for HANARO control rod supplied by a foreign company put difficulties on reactor operation due to discontinued production of the item and negative technical support. The development of the signal processing unit based on domestic technology has been carried out in order to solve the problems in issue and to ensure safe and reliable reactor operation. Considering the importance of its function, the project was proceeded by 3 separated steps as prototype, modification and practical application to HANARO. This paper describes the process test results of each developing stage

  2. R and D on Control Rod Magnetic Suspension Drive Mechanism of CARR

    Energy Technology Data Exchange (ETDEWEB)

    Xinxin, Wu; Huijie, Yan [Institute of Nuclear Energy Technology, Tsinghua University Neng Ke Lou (INET) Tsinghua University, 100084 Beijing (China)

    2011-07-01

    This paper deal with the research and develop (R and D) on Control Rod Magnetic Suspension Drive Mechanism (MSDM) of CARR. The MSDM is made up of tube, coil, armature, step motor, lead screw etc. The MSDM use electromagnetics as its main principle. The open solenoid electromagnet technique is employed to implement suspension function. It has advantages of high drive precision, high safety feature, good running reliability, easy maintenance and good economical property. The R and D process of MSDM has three phases including single coil electromagnet, principle prototype and engineering prototype. (author)

  3. A basic design of a double cladding fuel rod to control the irradiation temperature on nuclear fuels

    International Nuclear Information System (INIS)

    An instrumented capsule for a nuclear fuel irradiation test (hereinafter referred to as 'instrumented fuel capsule') has been developed to measure fuel characteristics, such as a fuel center and surface temperature, the internal pressure of a fuel rod, a fuel pellet elongation and neutron flux, during an irradiation test at HANARO. And six types of dual instrumented fuel rods, which allow for two characteristics to be measured simultaneously in one fuel rod, have been developed to enhance the efficiency of an irradiation test using an instrumented fuel capsule at HANARO. In the future, nuclear fuel irradiation tests under a high temperature condition are expected from users. To prepare for this request, we have continued developing the technologies for high temperature nuclear fuel irradiation tests at HANARO. The purpose of this paper is to control the temperature of nuclear fuels during an irradiation test at HANARO. Therefore we basically designed a double cladding fuel rod and an instrumented fuel capsule basically. The basic design of a double cladding rod was based on out-pile tests using mockups and the thermal analyses using some relevant codes. This paper presents the design and fabrication of the double cladding fuel rod mockups, the results of the out-pile tests, the results of the temperature calculation and the basic design of a double cladding fuel rod and an instrumented fuel capsule. (author)

  4. Development and design of control rod drive mechanisms for pressurized water reactors

    International Nuclear Information System (INIS)

    The Control Rod Drive Mechanisms (CRDM) for a Pressurized Water Reactor (PWR) are equipment, integrated to the reactor pressure vessel, incorporating mechanical and electrical components designed to move and position the control rods to guarantee the control of power and shutdown of the nuclear reactor, during normal operation, either in emergency or accidental situations. The type of CRDM used in PWR reactors, whose detailed individual description will be presented in this monograph are the Roller-Nut and Magnetic-Jack. The environment, where the CRDM performs its above presented operational functions, includes direct contact with the fluid used as coolant peculiar to the interior of the reactor, and its associated chemical characteristics, the radiation field next to the reactor core, and also the temperature and pressure in the reactor pressure vessel. So the importance of the CRDM design requirements related to its safety functions are emphasized. Finally, some aspects related to the mechanical and structural design of CRDM of a case study, considering the CRDM for a PWR from the experimental nuclear plant to be applied by CTMSP (Centro Tecnologico da Marinha em Sao Paulo), are pointed out. The design and development of these equipment (author)

  5. Aging and service wear of control rod drive mechanisms for BWR nuclear plants

    International Nuclear Information System (INIS)

    This Phase I Nuclear Plant Aging Research (NPAR) study examines the aging phenomena associated with BWR control rod drive mechanisms (CRDMs) and assesses the merits of various methods of ''managing'' this aging. Information for this study was acquired from (1) the results of a special CRDM aging questionnaire distributed to each US BWR utility, (2) a first-of-its-kind workshop held to discuss CRDM aging and maintenance concerns, (3) an analysis of the Nuclear Plant Reliability Data System (NPRDS) failure cases attributed to the control rod drive (CRD) system, and (4) personal information exchange with nuclear industry CRDM maintenance experts. Nearly 23% of the NPRDS CRD system component failure reports were attributed to the CRDM. The CRDM components most often requiring replacement due to normal wear and aging are the Graphiter seals. The predominant causes of aging for these seals are mechanical wear and thermally induced embrittlement More than 59% of the NPRDS CRD system failure reports were attributed to components that comprise the hydraulic control unit (HCU). The predominant HCU components experiencing the effects of service wear and aging are valve seals, discs, seats, stems, packing, and diaphragms. Since CRDM changeout and rebuilding is one of the highest dose, most physically challenging, and complicated maintenance activities routinely accomplished by BWR utilities, this report also highlights recent innovations in CRDM handling equipment and rebuilding tools that have resulted in significant dose reductions to the maintenance crews using them

  6. Parametric Study of Control Rod Exposure for PWR Burnup Credit Criticality Safety Analyses

    International Nuclear Information System (INIS)

    The Interim Staff Guidance on burnup credit (ISG-8) for pressurized water reactor (PWR) spent nuclear fuel (SNF), issued by the Nuclear Regulatory Commission's (NRC) Spent Fuel Project Office, recommends the use of analyses that provide an ''adequate representation of the physics'' and notes particular concern with the ''need to consider the more reactive actinide compositions of fuels burned with fixed absorbers or with control rods fully or partly inserted.'' In the absence of readily available information on the extent of control rod (CR) usage in U.S. PWRs and the subsequent reactivity effect of CR exposure on discharged SNF, NRC staff have indicated a need for greater understanding in these areas. In response, this paper presents results of a parametric study of the effect of CR exposure on the reactivity of discharged SNF for various CR designs (including Axial Power Shaping Rods), fuel enrichments, and exposure conditions (i.e., burnup and axial insertion). The study is performed in two parts. In the first part, two-dimensional calculations are performed, effectively assuming full axial CR insertion. These calculations are intended to bound the effect of CR exposure and facilitate comparisons of the various CR designs. In the second part, three-dimensional calculations are performed to determine the effect of various axial insertion conditions and gain a better understanding of reality. The results from the study demonstrate that the reactivity effect increases with increasing CR exposure (e.g., burnup) and decreasing initial fuel enrichment (for a fixed burnup). Additionally, the results show that even for significant burnup exposures, minor axial CR insertions (e.g., eff of a spent fuel cask

  7. Azcatl-CRP: An ant colony-based system for searching full power control rod patterns in BWRs

    International Nuclear Information System (INIS)

    We show a new system named AZCATL-CRP to design full power control rod patterns in BWRs. Azcatl-CRP uses an ant colony system and a reactor core simulator for this purpose. Transition and equilibrium cycles of Laguna Verde Nuclear Power Plant (LVNPP) reactor core in Mexico were used to test Azcatl-CRP. LVNPP has 109 control rods grouped in four sequences and currently uses control cell core (CCC) strategy in its fuel reload design. With CCC method only one sequence is employed for reactivity control at full power operation. Several operation scenarios are considered, including core water flow variation throughout the cycle, target different axial power distributions and Haling conditions. Azcatl-CRP designs control rod patterns (CRP) taking into account safety aspects such as k eff core value and thermal limits. Axial power distributions are also adjusted to a predetermined power shape

  8. Azcatl-CRP: An ant colony-based system for searching full power control rod patterns in BWRs

    Energy Technology Data Exchange (ETDEWEB)

    Ortiz, Juan Jose [Dpto. Sistemas Nucleares, ININ, Carr. Mexico-Toluca Km. 36.5, Salazar, Edo. de Mexico (Mexico)]. E-mail: jjortiz@nuclear.inin.mx; Requena, Ignacio [Dpto. Ciencias Computacion e I.A. ETSII Informatica, University of Granada, C. Daniel Saucedo Aranda s/n, 18071 Granada (Spain)]. E-mail: requena@decsai.ugr.es

    2006-01-15

    We show a new system named AZCATL-CRP to design full power control rod patterns in BWRs. Azcatl-CRP uses an ant colony system and a reactor core simulator for this purpose. Transition and equilibrium cycles of Laguna Verde Nuclear Power Plant (LVNPP) reactor core in Mexico were used to test Azcatl-CRP. LVNPP has 109 control rods grouped in four sequences and currently uses control cell core (CCC) strategy in its fuel reload design. With CCC method only one sequence is employed for reactivity control at full power operation. Several operation scenarios are considered, including core water flow variation throughout the cycle, target different axial power distributions and Haling conditions. Azcatl-CRP designs control rod patterns (CRP) taking into account safety aspects such as k {sub eff} core value and thermal limits. Axial power distributions are also adjusted to a predetermined power shape.

  9. A feasibility study on the innovative control rod driving mechanism, (1)

    International Nuclear Information System (INIS)

    The objectives of this study are to establish innovative Control Rod Driving Mechanisms (CRDMs) in order to achieve a highly safe and economic Advanced Marine Reactor (Ship). The innovative CRDMs to be carried this ship with Advanced Marine Reactor are to be installed in the reactor pressure vessel, since the internal CRDMs can eliminate 'Control Rod Ejection' which had been one of the design bases accidents on the licensing issues for conventional LWR with external CRDMs. This report presents the following works. After the discussion of the design requirement on the innovative CRDMs which had been selected as 'Concept Design of Advanced Marine Reactor (1), System Design (1), Design Study for Simplification of System', the scheme of the selected concept were considered and set up tentatively. Subsequently the development plan of the innovative CRDMs was made out. The major assets required for the internal CRDMs were heat-resistance and insulation-resistance of those electro-parts. The concept on the following key parts of the innovative CRDMs' electro-parts were embodied. Built-in-Motor, Scram-Magnet. Based on the concept, some kinds of typical heat-resistance wires estimated to be eligible used in the reactor pressure vessel were nominated. After the basic characteristics test screened out them to be a few, the heat-resistance wire eligible for the internal CRDMs' electro-parts were specified by the trial manufacturing and performance test of miniature coils made of selected wires. (author)

  10. Evaluation of Heat Removal from RBMK-1500 Core Using Control Rods Cooling Circuit

    Directory of Open Access Journals (Sweden)

    A. Kaliatka

    2008-01-01

    Full Text Available The Ignalina nuclear power plant is a twin unit with two RBMK-1500, graphite moderated, boiling water, multichannel reactors. After the decision was made to decommission the Ignalina NPP, Unit 1 was shut down on December 31, 2004, and Unit 2 is to be operated until the end of 2009. Despite of this fact, severe accident management guidelines for RBMK-1500 reactor at Ignalina NPP are prepared. In case of beyond design basis accidents, it can occur that no water sources are available at the moment for heat removal from fuel channels. Specificity of RBMK reactor is such that the channels with control rods are cooled with water supplied by the system totally independent from the reactor cooling system. Therefore, the heat removal from RBMK-1500 reactor core using circuit for cooling of rods in control and protection system can be used as nonregular mean for reactor cooldown in case of BDBA. The heat from fuel channels, where heat is generated, through graphite bricks is transferred in radial direction to cooled CPS channels. This article presents the analysis of possibility to remove heat from reactor core in case of large LOCA by employing CPS channels cooling circuit. The analysis was performed for Ignalina NPP with RBMK-1500 reactor using RELAP5-3D and RELAP5 codes. Results of the analysis have shown that, in spite of high thermal inertia of graphite, this heat removal from CPS channels allows to slow down effectively the core heat-up process.

  11. Digital driver of alternate current motors of the control rods in a nuclear research reactor

    International Nuclear Information System (INIS)

    The updating of the instruments as the operation console of the TRIGA Mark III Salazar Reactor is based on the use of a personal computer that works as data acquisition and control device. The power changes on the reactor have been made through the inserting or extraction of four control rods, that they are operated by mechanisms based in alternate current motors. That is with the object to handling each of the bars and so avoiding too the degradation about the performance of the computer of process. Also it is using four drives of smart kind which do the basic duties for generating the control signals and verifying the sensors state of the limits in continuous form. The computer and drivers are organized as a ring net using the serial port R S-232. The computer of process sends the orders and the identification of destination instrument throughout the net. (Author)

  12. Test simulation of drop of control rods in Trillo NPP with CASMO-4 and SIMULATE-3; Simulacion de prueba de caida de barras de control en CNT con CASMO/SIMULATE

    Energy Technology Data Exchange (ETDEWEB)

    Encinas, L.; Bermejo, J. A.; Ortego, A.

    2011-07-01

    In this paper we describe the rod drop test that was performed in control CN Trillo for justifier the modification of design about of the restriction system functions, in particular the function of detection of inadvertent drop of control rods.

  13. ROBUST CONTROL OF AN ELECTRO-HYDRAULIC PROPORTIONAL SPEED CONTROL SYSTEM WITH A SINGLE-ROD HYDRAULIC ACTUATOR

    Institute of Scientific and Technical Information of China (English)

    Yang Jian; Xu Bing; Yang Huayong

    2005-01-01

    A robust control algorithm is proposed to focus on the non-linearity and parameters'uncertainties of an electro-hydraulic proportional speed control system (EHPSCS) with a single-rod hydraulic actuator. The robust controller proposed does not need to design stable compensator in advance, is simple in design and has large scope of uncertainty applications. The feedback gains of the robust controller proposed are small, so it is easily implemented in engineering applications.Experimental research on the speed control under the different conditions is carried out for an EHPSCS. Experimental results show that the robust controller proposed has better robustness subject to parametric uncertainties, and adaptability of parameters' variation of control system itself and plant parameter variation.

  14. Cost analysis of magnetically controlled growing rods compared with traditional growing rods for early-onset scoliosis in the US: an integrated health care delivery system perspective

    Science.gov (United States)

    Polly, David W; Ackerman, Stacey J; Schneider, Karen; Pawelek, Jeff B; Akbarnia, Behrooz A

    2016-01-01

    Purpose Traditional growing rod (TGR) for early-onset scoliosis (EOS) is effective but requires repeated invasive surgical lengthenings under general anesthesia. Magnetically controlled growing rod (MCGR) is lengthened noninvasively using a hand-held magnetic external remote controller in a physician office; however, the MCGR implant is expensive, and the cumulative cost savings have not been well studied. We compared direct medical costs of MCGR and TGR for EOS from the US integrated health care delivery system perspective. We hypothesized that over time, the MCGR implant cost will be offset by eliminating repeated TGR surgical lengthenings. Methods For both TGR and MCGR, the economic model estimated the cumulative costs for initial implantation, lengthenings, revisions due to device failure, surgical-site infections, device exchanges (at 3.8 years), and final fusion, over a 6-year episode of care. Model parameters were estimated from published literature, a multicenter EOS database of US institutions, and interviews. Costs were discounted at 3.0% annually and represent 2015 US dollars. Results Of 1,000 simulated patients over 6 years, MCGR was associated with an estimated 270 fewer deep surgical-site infections and 197 fewer revisions due to device failure compared with TGR. MCGR was projected to cost an additional $61 per patient over the 6-year episode of care compared with TGR. Sensitivity analyses indicated that the results were sensitive to changes in the percentage of MCGR dual rod use, months between TGR lengthenings, percentage of hospital inpatient (vs outpatient) TGR lengthenings, and MCGR implant cost. Conclusion Cost neutrality of MCGR to TGR was achieved over the 6-year episode of care by eliminating repeated TGR surgical lengthenings. To our knowledge, this is the first cost analysis comparing MCGR to TGR – from the US provider perspective – which demonstrates the efficient provision of care with MCGR. PMID:27695352

  15. Fuzzy logic modeling and control of steel rod quenching after hot rolling

    Science.gov (United States)

    Giorleo, G.; Memola Capece Minutolo, F.; Sergi, V.

    1997-10-01

    Reinforced concrete rod produced by European Community countries must comply with standards that establish minimum strength and tensile properties along with other technological and geometrical characteristics; however, possible variability within the assigned limits is not specified. Consequently, a number of manufacturing methods are now used, with the result that over time the mechanical properties of these products vary widely. Increased competition has led to the development of new procedures incorporating both process and quality control. One example is a process based on the heat treatment undergone by the metal bars leaving the final stand of the rolling mill train. In this way, the mechanical and technological properties can be graduated, thereby enhancing strength (particularly yield point) without altering the deformability of the material. This procedure does away with the need to alter the chemical composition of the steel used to manufacture the rods. Process adjustment still relies on the experience of the production manager, however. This paper examines the possibility of applying fuzzy logic computer techniques to the heat treatment process in order to render it more rational and independent of operator unreliability.

  16. A response matrix method for improved modeling of neutron streaming in large control rod holes of prismatic VHTR cores - 217

    International Nuclear Information System (INIS)

    This paper presents a response matrix method developed for accurate modeling of neutron streaming through empty, large control rod holes in VHTRs. In this approach, a response matrix based on transport solution is derived for each control rod channel region and embedded in the whole-core transport solution scheme. Depending on the region geometry only, each element of the response matrix represents the outgoing partial current at a surface due to a unit incoming partial current at another surface. In order to improve the axial solution accuracy, this response matrix approach was incorporated into the DeCART code that solves whole-core transport problems by coupling two-dimensional MOC and one-dimensional nodal solutions. Verification test results showed very good agreements in control rod worth and axial power distributions with MCNP5 solutions. (authors)

  17. Internal Leakage Fault Detection and Tolerant Control of Single-Rod Hydraulic Actuators

    Directory of Open Access Journals (Sweden)

    Jianyong Yao

    2014-01-01

    Full Text Available The integration of internal leakage fault detection and tolerant control for single-rod hydraulic actuators is present in this paper. Fault detection is a potential technique to provide efficient condition monitoring and/or preventive maintenance, and fault tolerant control is a critical method to improve the safety and reliability of hydraulic servo systems. Based on quadratic Lyapunov functions, a performance-oriented fault detection method is proposed, which has a simple structure and is prone to implement in practice. The main feature is that, when a prescribed performance index is satisfied (even a slight fault has occurred, there is no fault alarmed; otherwise (i.e., a severe fault has occurred, the fault is detected and then a fault tolerant controller is activated. The proposed tolerant controller, which is based on the parameter adaptive methodology, is also prone to realize, and the learning mechanism is simple since only the internal leakage is considered in parameter adaptation and thus the persistent exciting (PE condition is easily satisfied. After the activation of the fault tolerant controller, the control performance is gradually recovered. Simulation results on a hydraulic servo system with both abrupt and incipient internal leakage fault demonstrate the effectiveness of the proposed fault detection and tolerant control method.

  18. A methodology for obtaining the control rods patterns in a BWR using systems based on ants colonies

    International Nuclear Information System (INIS)

    In this work the AZCATL-PBC system based on a technique of ants colonies for the search of control rods patterns of those reactors of the Nuclear Power station of Laguna Verde (CNLV) is presented. The technique was applied to a transition cycle and one of balance. For both cycles they were compared the kef values obtained with a Haling calculation and the control rods pattern proposed by AZCATL-PBC for a burnt one fixed. It was found that the methodology is able to extend the length of the cycle with respect to the Haling prediction, maintaining sure to the reactor. (Author)

  19. The differential characteristics of control rods of VVER-1000 core simulator at a low number of axial mesh points

    Science.gov (United States)

    Bolsunov, A. A.; Karpov, S. A.

    2013-12-01

    An algorithm for refining the differential characteristics of the control rods (CRs) of the control and protection system (CPS) for a neutronics model of the VVER-1000 simulator at a low number of axial mesh points of the core is described. The problem of determining the constants for a cell with a partially inserted CR is solved. The cell constants obtained using the proposed approach ensure smoothing of the differential characteristics of an absorbing rod. The algorithm was used in the VVER-1000 simulators (Bushehr NPP, unit no. 1; Rostov NPP, unit no. 1; and Balakovo NPP, unit no. 4).

  20. Ex-core detector response caused by control rod misalignment observed during operation of the reactor on the nuclear ship Mutsu

    Energy Technology Data Exchange (ETDEWEB)

    Itagaki, Masafumi; Miyoshi, Yoshinori (Japan Atomic Energy Research Inst., Ibaraki (Japan)); Gakuhari, Kazuhiko; Okada, Noboru (Mitsubishi Atomic Power Industries, Inc., Tokyo (Japan)); Sakai, Tomohiro (Japan Research Inst., Ltd., Tokyo (Japan))

    1993-04-01

    Unexpected deviations of ex-core neutron detector signals were observed during a voyage of the Japanese nuclear ship, Mutsu. From detailed three-dimensional analyses, this phenomenon was determined to be caused by an asymmetrical neutron source distribution in the core due to a small misalignment between the two control rods of a control rod group. A systematic ex-core detector response experiment was performed during the Mutsu's third experimental voyage to gain some understanding of the relationship between the control rod pattern and the detector response characteristics. Results obtained from analyses of the experiment indicate that the Crump-Lee technique, using calculated three-dimensional source distributions for various control rod patterns, provides good agreement between the calculated and measured detector responses. Xenon transient analyses were carried out to generate accurate three-dimensional source distributions for predicting the time-dependent detector response characteristics. Two types of ex-core detector responses are caused by changes in the control rod pattern in the Mutsu reactor: the detector response ratio tends to decrease with the withdrawal of a group of control rods as a pair, and a difference in the positions of the control rods in a group causes signal deviations among the four ex-core detectors. Control rod misalignment does not greatly affect the mean value of the four detector signals, and the deviation can be minimized if the two rods within a group are set at the same elevation at the time of detector calibration.

  1. Analysis of reactivity initiated transient from control rod failure events of a molten salt reactor

    International Nuclear Information System (INIS)

    In a molten salt reactor (MSR), the fuel is dissolved in fluoride salt. In this paper, the reactivity worth and reactivity initiated transient of Molten-Salt Reactor Experiment (MSRE) in the control rod failure events are analyzed. The point kinetic coupling heat-transfer model with decay character of six-group delayed neutron precursors due to the fuel motion is applied. The relative power and temperature transient under reactivity step and ramp initiated at different power levels are studied. The results show that the reactor power and temperature increase to a maximum, where they begin to decrease to stable values. Comparing with full power level, the transient result at low power level is more serious. The results are of help in our study on safety characteristics of an MSR system. (authors)

  2. Development of direct digital reactivity computer system (DDRCS) for dynamic control rod reactivity measurement(DCRM)

    Energy Technology Data Exchange (ETDEWEB)

    Woo, I. T.; Ryu, S. J.; Sin, H. C.; Lee, E. K.; Bae, S. M.; Lee, C. S. [Korea Electric Power Research Institute, Taejon (Korea, Republic of)

    2002-10-01

    Neutron Flux level may be rapidly decreased to 1/100{approx}1/1000th order of magnitude during DCRM(Dynamic Control rod Reactivity Measurement) test. Because the conventional DRCS(Digital Reactivity Computer System) converts NIS current signal to analog one with the range from 0 to 2 volt, and computes reactivity, the DRCS can not measure the widely changed flux level during DCRM test. The DDRCS(Direct Digital Reactivity Computer System) which is developed in this study can measure the current of all the range directly and reduce the burden to maintain the equipments, because of its simplified structure. The function of DDRCS was fully validated through three times of plant low power physics tests. The software program to handle all the items of low power physics test will be developed.

  3. The effect of aging upon CE and B ampersand W control rod drives

    International Nuclear Information System (INIS)

    The effect of aging upon the Babcock ampersand Wilcox (B ampersand W) and Combustion Engineering (CE) Control Rod Drive (CRD) systems has been evaluated as part of the USNRC Nuclear Plant Aging Research (NPAR) program. Operating experience data for the 1980--1990 time period was reviewed to identify predominant failure modes, causes, and effects. These results, in conjunction with an assessment of component materials and operating environment, conclude that both systems are susceptible to age degradation. System failures have resulted in significant plant effects, including power reductions, plant shutdowns, scrams, and Engineered Safety Feature (ESF) actuation. Current industry inspection and maintenance practices were assessed. Some of these practices effectively address aging, while others do not

  4. Heat Removal Performance of Hybrid Control Rod for Passive In-Core Cooling System

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kyung Mo; Jeong, Yeong Shin; Kim, In Guk; Bang, In Cheol [UNIST, Ulsan (Korea, Republic of)

    2015-10-15

    The two-phase closed heat transfer device can be divided by thermosyphon heat pipe and capillary wicked heat pipe which uses gravitational force or capillary pumping pressure as a driving force of the convection of working fluid. If there is a temperature difference between reactor core and ultimate heat sink, the decay heat removal and reactor shutdown is possible at any accident conditions without external power sources. To apply the hybrid control rod to the commercial nuclear power plants, its modelling about various parameters is the most important work. Also, its unique geometry is coexistence of neutron absorber material and working fluid in a cladding material having annular vapor path. Although thermosyphon heat pipe (THP) or wicked heat pipe (WHP) shows high heat transfer coefficients for limited space, the maximum heat removal capacity is restricted by several phenomena due to their unique heat transfer mechanism. Validation of the existing correlations on the annular vapor path thermosyphon (ATHP) which has different wetted perimeter and heated diameter must be conducted. The effect of inner structure, and fill ratio of the working fluid on the thermal performance of heat pipe has not been investigated. As a first step of the development of hybrid heat pipe, the ATHP which contains neutron absorber in the concentric thermosyphon (CTHP) was prepared and the thermal performance of the annular thermosyphon was experimentally studied. The heat transfer characteristics and flooding limit of the annular vapor path thermosyphon was studied experimentally to model the performance of hybrid control rod. The following results were obtained: (1) The annular vapor path thermosyphon showed better evaporation heat transfer due to the enhanced convection between adiabatic and condenser section. (2) Effect of fill ratio on the heat transfer characteristics was negligible. (3) Existing correlations about flooding limit of thermosyphon could not reflect the annular vapor

  5. Activation calculation of steel of the control rods of TRIGA Mark III reactor; Calculo de activacion del acero de las barras de control del reactor TRIGA Mark III

    Energy Technology Data Exchange (ETDEWEB)

    Garcia M, T.; Cruz G, H. S.; Ruiz C, M. A.; Angeles C, A., E-mail: teodoro.garcia@inin.gob.mx [ININ, Carretera Mexico-Toluca sn, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2014-10-15

    In the pool of TRIGA Mark III reactor of the Instituto Nacional de Investigaciones Nucleares (ININ), there are control rods that were removed from the core, and which are currently on shelves of decay. These rods were part of the reactor core when only had fuel standard (from 1968-1989). To conduct a proper activation analysis of the rods, is very important to have well-characterized the materials which are built, elemental composition of the same ones, the atomic densities and weight fractions of the elements that constitute them. To determine the neutron activation of the control rods MCNP5 code was used, this code allows us to have well characterized the radionuclides inventory that were formed during irradiation of the control rods. This work is limited to determining the activation of the steel that is part of the shielding of the control rods, the nuclear fuel that is in the fuel follower does not include. The calculation model of the code will be validated with experimental measurements and calculating the activity of fission products of the fuel follower which will take place at the end of 2014. (Author)

  6. Electromagnet Tests on Primary Control Rod Drive Mechanism of a Prototype Gen-IV SFR

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jaehan; Koo, Gyeonghoi [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    The primary control system is used for power control, burn-up compensation and reactor shutdown in response to demands from the plant control or protection systems. This paper describes the lifting and holding force tests of the electromagnetic equipment of a primary control rod drive mechanism (CRDM). The supply currents above 1.5 A and 15A on coil are required for holding the CRA with a 1mm gap, and lifting the CRA with 10mm gap, respectively. The currents cover all the loads to be expected in driveline. The S10C carbon steel can be replaced with the SS410 stainless steel by increasing the supply current about 30%. The assist spring, pushing down the tension tube with a compressed force, plays an important role when the operation load is smaller than 20kgf. The spring force can cease a time delay on the free drop of the tension tube carrying a light driving mass because a residual electromagnetic force may exist for a while even though the supply power is cut off. The holding current can be reduced by closing the gap size of 1mm between inner core and armature.

  7. The JASPER system, an innovating, competitive tool for rod cluster control assembly (RCCA) in-service inspection

    International Nuclear Information System (INIS)

    Taking benefit from the experience of the AREVA NP group, a new tool for the inspection of rod control cluster assemblies (RCCA) was jointly developed by Intercontrole and AREVA NP Fuel Division. The valuable know-how of R/D Tech (today Zetec) engineers in the field of UT signal spectrum analysis was a key factor of success in the development. JASPER (an acronym for Joint Advanced System for Performant Examination of RCCA) combines three measurements, one of which is an innovation: - Profilometry using a time of flight measurement, for the outer dimensions of the clad - Eddy current detection of cracks - Direct measurement of the rod wall thickness by spectrum analysis of the UT echoes, thus adding considerable interest in the examination process. UT measurements are performed on the whole length of the rod, including the weld of the lower cap. ET measurements are performed on the lower length of the rod. UT data are systematically recorded and analysed for detection and characterization of indications, with no retest. ET measurements are triggered upon request of the Utility, depending on the age of the rod. Data acquisition and processing thus require a constant duration for each assembly; the inspection duration is actually shortened by a 20% factor. The technique was qualified in-house by a number of tests. (orig.)

  8. Proceedings of the specialist meeting on nuclear fuel and control rods: operating experience, design evolution and safety aspects

    International Nuclear Information System (INIS)

    Design and management of nuclear fuel has undergone a strong evolution process during past years. The increase of the operating cycle length and of the discharge burnup has led to the use of more advanced fuel designs, as well as to the adoption of fuel efficient operational strategies. The analysis of recent operational experience highlighted a number of issues related to nuclear fuel and control rod events raising concerns about the safety aspects of these new designs and operational strategies, which led to the organisation of this Specialists Meeting on fuel and control rod issues. The meeting was intended to provide a forum for the exchange of information on lessons learned and safety concern related to operating experience with fuel and control rods (degradation, reliability, experience with high burnup fuel, and others). After an opening session 6 papers), this meeting was subdivided into four sessions: Operating experience and safety concern (technical session I - 6 papers), Fuel performance and operational issues (technical session II - 7 papers), Control rod issues (technical session III - 9 papers), Improvement of fuel design (technical session IV.A - 4 papers), Improvement on fuel fabrication and core management (technical session IV.B - 6 papers)

  9. Sucker rods

    Energy Technology Data Exchange (ETDEWEB)

    Hoffmann, J.; Preis, L.

    1987-12-08

    The sucker rod system in a deep well sucker rod pump consists of a plurality of unidirectionally reinforced composite fiber rods extending substantially parallel but not in contact with each other, the cross-sectional area of which rods is less than 1 cm/sup 2/. This enables the advantageous material properties to be utilized to a high degree. The sucker rod system can be assembled on site. The individual composite fiber rods can be monitored when they are in the working position.

  10. The Application of Paret/ANL Code for Accident Analysis on Inadvertent Control Rod Withdrawal for RSG GAS Reactor

    International Nuclear Information System (INIS)

    The analysis is intended to take a look the condition of safety parameters such as fuel and clad temperature, and minimum safety margin against flow instability (S) in the occurrence of inadvertent control rod withdrawal at nominal power, which is performed by PARET/ANL Code. The accident is initiated when all control rods are simultaneously withdrawn with maximum speed of 0.0564 cm/s which consequently gives maximum reactivity insertion rate σρ/σt = 2.82 x 10-4/s, resulting in the Reactor Protection System (RPS) respond to scram the reactor by dropping the control rods into the core. The primary cooling system is assumed to be in normal operation. It is postulated that the first trip signal from over power is not effective to scram the reactor, but only the second signal from Floating Limit Value eventually causes a scram with 0.5 s delays. During the occurrence of inadvertent control rods withdrawal at 30 MW of initial power, the maximum fuel and clad temperature reach 181.29oC and 137.62oC, respectively and the peak power of 37.11 MW. Meanwhile the minimum value of S reaches 2.62. Therefore, during the occurrence of control rods withdrawal at initial power of 30 MW, the integrity of fuel and clad can be maintained secure since they do not exceed the maximum limit of fuel and clad temperature of 207oC and 145oC, respectively and the minimum value of S is still higher than the design limit of 1.48 for anticipated transient

  11. Wave propagation visualization in an experimental model for a control rod drive mechanism assembly

    International Nuclear Information System (INIS)

    Highlights: → We fabricate a full-scale mock-up of the control rod drive mechanism (CRDM) assembly in the upper reactor head of the nuclear power plant. → An ultrasonic propagation imaging method using a scanning laser ultrasonic generator is proposed to visualize and simulate ultrasonic wave propagation around the CRDM assembly. → The ultrasonic source location and frequency are simulated by changing the sensor location and the band pass-filtering zone. → The ultrasonic propagation patterns before and after cracks in the weld and nozzle of the CRDM assembly are analyzed. - Abstract: Nondestructive inspection techniques such as ultrasonic testing, eddy current testing, and visual testing are being developed to detect primary water stress corrosion cracks in control rod drive mechanism (CRDM) assemblies of nuclear power plants. A unit CRDM assembly consists of a reactor upper head including cladding, a penetration nozzle, and J-groove dissimilar metal welds with buttering. In this study, we fabricated a full-scale CRDM assembly mock-up. An ultrasonic propagation imaging (UPI) method using a scanning laser ultrasonic generator is proposed to visualize and simulate ultrasonic wave propagation around the thick and complex CRDM assembly. First, the proposed laser UPI system was validated for a simple aluminium plate by comparing the ultrasonic wave propagation movie (UWPM) obtained using the system with numerical simulation results reported in the literature. Lamb wave mode identification and damage detectability, depending on the ultrasonic frequency, were also included in the UWPM analysis. A CRDM assembly mock-up was fabricated in full-size and its vertical cross section was scanned using the laser UPI system to investigate the propagation characteristics of the longitudinal and Rayleigh waves in the complex structure. The ultrasonic source location and frequency were easily simulated by changing the sensor location and the band pass filtering zone

  12. Rodding Surgery

    Science.gov (United States)

    ... a rod or nail into the internal cavity (medullary canal) of a long bone. Purpose of Rodding ... Osteogenesis Imperfecta: A Translational Approach to Brittle Bone Disease 1 st edition. New York, NY: Elsevier Academic ...

  13. Development of a HTSMA-Actuated Surge Control Rod for High-Temperature Turbomachinery Applications

    Science.gov (United States)

    Padula, Santo, II; Noebe, Ronald; Bigelow, Glen; Culley, Dennis; Stevens, Mark; Penney, Nicholas; Gaydosh, Darrell; Quackenbush, Todd; Carpenter, Bernie

    2007-01-01

    In recent years, a demand for compact, lightweight, solid-state actuation systems has emerged, driven in part by the needs of the aeronautics industry. However, most actuation systems used in turbomachinery require not only elevated temperature but high-force capability. As a result, shape memory alloy (SMA) based systems have worked their way to the forefront of a short list of viable options to meet such a technological challenge. Most of the effort centered on shape memory systems to date has involved binary NiTi alloys but the working temperatures required in many aeronautics applications dictate significantly higher transformation temperatures than the binary systems can provide. Hence, a high temperature shape memory alloy (HTSMA) based on NiTiPdPt, having a transformation temperature near 300 C, was developed. Various thermo-mechanical processing schemes were utilized to further improve the dimensional stability of the alloy and it was later extruded/drawn into wire form to be more compatible with envisioned applications. Mechanical testing on the finished wire form showed reasonable work output capability with excellent dimensional stability. Subsequently, the wire form of the alloy was incorporated into a benchtop system, which was shown to provide the necessary stroke requirements of approx.0.125 inches for the targeted surge-control application. Cycle times for the actuator were limited to 4 seconds due to control and cooling constraints but this cycle time was determined to be adequate for the surge control application targeted as the primary requirement was initial actuation of a surge control rod, which could be completed in approximately one second.

  14. Tensile and impact testing of an HFBR [High Flux Beam Reactor] control rod follower

    International Nuclear Information System (INIS)

    The Materials Technology Group of the Department of Nuclear Energy (DNE) at Brookhaven National Laboratory (BNL) undertook a program to machine and test specimens from a control rod follower from the High Flux Beam Reactor (HFBR). Tensile and Charpy impact specimens were machined and tested from non-irradiated aluminum alloys in addition to irradiated 6061-T6 from the HFBR. The tensile test results on irradiated material showed a two-fold increase in tensile strength to a maximum of 100.6 ksi. The impact resistance of the irradiated material showed a six-fold decrease in values (3 in-lb average) compared to similar non-irradiated material. Fracture toughness (KI) specimens were tested on an unirradiated compositionally and dimensionally similar (to HFBR follower) 6061 T-6 material with Kmax values of 24.8 ± 1.0 Ksi√in (average) being obtained. The report concludes that the specimens produced during the program yielded reproducible and believable results and that proper quality assurance was provided throughout the program. 9 figs., 6 tabs

  15. Improved dashpot constructions for a nuclear reactor control rod guide thimble

    International Nuclear Information System (INIS)

    A dashpot in a control rod guide thimble for a nuclear fuel assembly includes a lower tubular portion of an elongated main tube of the guide thimble and an auxiliary hollow tube of smaller internal diameter associated with the lower portion of the main tube, and an end plug attached to a lower end portion of the auxiliary tube. In one embodiment the auxiliary tube is inserted into the main tube lower end and has an outside diameter slightly less than an inside diameter of the main tube to permit a close fitting relationship between an exterior surface of the auxiliary tube and an interior surface of the main tube lower portion. In a second embodiment, the auxiliary tube is butt-welded to the lower end of the main tube. The auxiliary tube also has an upper end portion with an inside surface portion in axial cross-section flaring upwardly and outwardly to provide a tapered transition extending between and connecting an interior surface of the auxiliary tube with that of the main tube. (author)

  16. Nuclear Data Library Effects on Fast to Thermal Flux Shapes Around PWR Control Rod Tips

    Science.gov (United States)

    Vasiliev, A.; Ferroukhi, H.; Zhu, T.; Pautz, A.

    2014-04-01

    The development of a high-fidelity computational scheme to estimate the accumulated fluence at the tips of PWR control rods (CR) has been initiated at the Paul Scherrer Institut (PSI). Both the fluence from high-energy (E>1 MeV) neutrons as well as for the thermal range (Erod surfaces, especially for fully withdrawn CRs. In that case, the tip is located a short distance above the core/reflector interface and since this situation corresponds to a large part of reactor operation, the accumulated fluence will highly depend on the achieved calculation accuracy and precision in this non-fueled zone. The objective of the work presented in this paper is to quantify the influence of nuclear data on the calculated fluxes at the CR tips by (1) conducting a systematic comparison of modern neutron cross-section libraries, including JENDL-4.0, JEFF-3.1.1 and ENDF/B-VII.0, and (2) by quantifying the uncertainties in the neutron flux calculations with the help of available neutron cross-section variances/covariances data. For completeness, the magnitude of these nuclear data-based uncertainties is also assessed in relation to the influence from other typical sources of modeling uncertainties/biases.

  17. Reliability centered maintenance applied to the control rod drives of a nuclear power reactor

    International Nuclear Information System (INIS)

    Reliability Centered Maintenance (RCM) offers a hybrid reliability analysis methodology to evaluate the level of maintenance and direct the resources in an effective manner. The RCM analysis consists of several steps, including a Logic Tree Analysis (LTA) for identification of applicable and efficient preventive maintenance task. As a result of the analysis the total amount of maintenance of equipment is decreased or increased, depending on whether the failures are potential having adverse effects on plant safety, availability or economics. A RCM analysis results thus in an improved preventive maintenance program, which is supposed to decrease the maintenance and outage costs and at the same time increase the system safety and reliability. A problem is the relatively large and slow analysis effort, the fact which tends to inhibit the large scale and continuous application of this useful method. A case study is described, in which the control rod equipment at the TVO I/II nuclear units was analyzed and the problematics of the RCM-method is discussed. (au)

  18. Disposal Of Irradiated Cadmium Control Rods From The Plumbrook Reactor Facility

    International Nuclear Information System (INIS)

    Innovative mixed waste disposition from NASA's Plum Brook Reactor Facility was accomplished without costly repackaging. Irradiated characteristic hardware with contact dose rates as high as 8 Sv/hr was packaged in a HDPE overpack and stored in a Secure Environmental Container during earlier decommissioning efforts, awaiting identification of a suitable pathway. WMG obtained regulatory concurrence that the existing overpack would serve as the macro-encapsulant per 40CFR268.45 Table 1.C. The overpack vent was disabled and the overpack was placed in a stainless steel liner to satisfy overburden slumping requirements. The liner was sealed and placed in shielded shoring for transport to the disposal site in a US DOT Type A cask. Disposition via this innovative method avoided cost, risk, and dose associated with repackaging the high dose irradiated characteristic hardware. In conclusion: WMG accomplished what others said could not be done. Large D and D contractors advised NASA that the cadmium control rods could only be shipped to the proposed Yucca mountain repository. NASA management challenged MOTA to find a more realistic alternative. NASA and MOTA turned to WMG to develop a methodology to disposition the 'hot and nasty' waste that presumably had no path forward. Although WMG lead a team that accomplished the 'impossible', the project could not have been completed with out the patient, supportive management by DOE-EM, NASA, and MOTA. (authors)

  19. The Design, Fabrication, and Characteristic Experiment of the Electromagnet of Bottom-mounted Control Rod Drive Mechanism for Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Huh, Hyung; Cho, Yeong Garp; Choi, Myoung Hwan; Kim, Ji Ho; Kim, Jong In [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2011-10-15

    A control rod drive mechanism (CRDM) is located in the reactor pool top (Top-mounted) or the room below the reactor pool (Bottom-mounted). The function of the CRDM is to insert, withdraw, or maintain neutron absorbing material at any required position in the reactor core in order to maintain reactivity control of the core. There are so many kinds of CRDMs, such as magnetic-jack type, hydraulic type, rack and pinion type, chain type and linear or rotary step motor and so on. As a part of a new project, we are investigating the bottom-mounted control rod drive mechanism as shown in Fig. 1. To have a better knowledge of the electromagnetic and magnetic characteristics, numerical models of bottom-mounted CEDM are investigated. In this study, we clarified thrust force characteristics of the electromagnet by experiment and simulation, and verified the propriety of the FEM analysis by comparing it with the results

  20. Genetic algorithm based active vibration control for a moving flexible smart beam driven by a pneumatic rod cylinder

    Science.gov (United States)

    Qiu, Zhi-cheng; Shi, Ming-li; Wang, Bin; Xie, Zhuo-wei

    2012-05-01

    A rod cylinder based pneumatic driving scheme is proposed to suppress the vibration of a flexible smart beam. Pulse code modulation (PCM) method is employed to control the motion of the cylinder's piston rod for simultaneous positioning and vibration suppression. Firstly, the system dynamics model is derived using Hamilton principle. Its standard state-space representation is obtained for characteristic analysis, controller design, and simulation. Secondly, a genetic algorithm (GA) is applied to optimize and tune the control gain parameters adaptively based on the specific performance index. Numerical simulations are performed on the pneumatic driving elastic beam system, using the established model and controller with tuned gains by GA optimization process. Finally, an experimental setup for the flexible beam driven by a pneumatic rod cylinder is constructed. Experiments for suppressing vibrations of the flexible beam are conducted. Theoretical analysis, numerical simulation and experimental results demonstrate that the proposed pneumatic drive scheme and the adopted control algorithms are feasible. The large amplitude vibration of the first bending mode can be suppressed effectively.

  1. Wavelet filter based de-noising of weak neutron flux signal for dynamic control rod reactivity measurement

    Energy Technology Data Exchange (ETDEWEB)

    Park, Moon Ghu; Bae Sung Man; Lee, Chang Sup [Korea Electric Power Research Institute, Taejon (Korea, Republic of)

    2002-10-01

    The measurement and validation of control rod bank (group) worths are typically required by the start-up physics test standard programs for Pressurized Water Reactors (PWR). Recently, the method of DCRM{sup TM} (Dynamic Control rod Reactivity Measurement) technique is developed by KEPRI and will be implemented in near future. The method is based on the fast and complete bank insertion within the short period of time which makes the range of the reactivity variation very large from the below of the background gamma level to the vicinity of nuclear heating point. The weak flux signal below background gamma level is highly noise contaminated, which invokes the large reactivity fluctuation. This paper describes the efficient noise filtering method with wavelet filters. The performance of developed method is demonstrated with the measurement data at YGN-3 cycle 7.

  2. Laser Ultrasonic System for Surface Crack Visualization in Dissimilar Welds of Control Rod Drive Mechanism Assembly of Nuclear Power Plant

    OpenAIRE

    Yun-Shil Choi; Hyomi Jeong; Jung-Ryul Lee

    2014-01-01

    In this paper, we propose a J-groove dissimilar weld crack visualization system based on ultrasonic propagation imaging (UPI) technology. A full-scale control rod drive mechanism (CRDM) assembly specimen was fabricated to verify the proposed system. An ultrasonic sensor was contacted at one point of the inner surface of the reactor vessel head part of the CRDM assembly. Q-switched laser beams were scanned to generate ultrasonic waves around the weld bead. The localization and sizing of the cr...

  3. A Position Estimation Method of the Control Rod Guide Tube with Matched Filters

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae C.; Seop, Jun H.; Choi, Yu R.; Kim, Jae H. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2005-07-01

    which we are looking for in that input image. In addition to the presence of the guide tube and pins, the cross correlation value will have a maximum at the exact position of the object. As a result, we can perform the inspection without any troublesome jobs such as a guide rail installation. We studied this algorithm for applying it to the control rod guide tubes inspection robot and tried an inspection without on operator's intervention.

  4. Concept of the core for a small-to-medium-sized BWR that does not use control rods during normal operation

    International Nuclear Information System (INIS)

    A small-to-medium-sized boiling water reactor (BWR) with a natural circulation system is being developed for countries where initial investment funds for construction are limited and electricity transmission networks have not been fully constructed. To lighten operators' work load, a core that does not use control rods during normal operation (control rod-free core) was developed by using a neutronics calculation system coupled with core flow evaluation. The control rod-free core had large core power fluctuation with conventional burnable poison design. The target of core power fluctuation was set to less than 10% and was achieved by optimization of burnable poison arrangement. (author)

  5. Control rod effects on reaction rate distributions in tight pitched PuO2-UO2 fuel assembly

    International Nuclear Information System (INIS)

    Investigations were made for the heterogeneity effects caused by insertion or withdrawal of a B4C control rod on fine structure of reaction rates distributions in a tight pitched PuO2-UO2 fuel assembly. Analysis was carried out by using the VIM and SRAC codes with the libraries based on JENDL-2 for the hexagonal fuel assembly basically corresponding to the PROTEUS-LWHCR experimental core. The reaction rates are affected more remarkably by the withdrawal of the control rod rather than its insertion. The changes of the reaction rates were decomposed into three terms of spectrum shifts, the changes of effective cross sections with fine groups, and their higher order components. From the analysis, it is concluded that most changes of reaction rates are caused by spectral shifts. The SRAC code with fine group constants can predict the distribution of reaction rates and their ratios with the accuracy of about 5 % except for the values related to Pu-242 capture rate, as compared with the VIM results. To increase the accuracy, it is necessary to generate the effective cross sections of the fuel near control rods with consideration of the heterogeneities in the fuel assembly. (author)

  6. Opposed piston engine having fuel inlet through rod controlled piston port

    Energy Technology Data Exchange (ETDEWEB)

    Lively, E.P. Sr.

    1991-07-09

    This patent describes an internal combustion engine. It comprises at least one of each of an intake port, exhaust port and fuel inlet port; a pair of opposed pistons within a cylinder of the engine defining a combustion chamber; one of the pair of pistons opening and closing the at least one exhaust port, the one piston including the fuel inlet port therethrough; a connecting rod operatively connecting the one piston to a driven shaft, the connecting rod having an end portion which opens and closes the fuel inlet port.

  7. TRIGA control rod position and reactivity transient Monitoring by Neural Networks

    International Nuclear Information System (INIS)

    Plant sensors drift or malfunction and operator actions in nuclear reactor control can be supported by sensor on-line monitoring, and data validation through soft-computing process. On-line recalibration can often avoid manual calibration or drifting component replacement. DSP requires prompt response to the modified conditions. Artificial Neural Network (ANN) and Fuzzy logic ensure: prompt response, link with field measurement and physical system behaviour, data incoming interpretation, and detection of discrepancy for mis-calibration or sensor faults. ANN (Artificial Neural Network) is a system based on the operation of biological neural networks. Although computing is day by day advancing, there are certain tasks that a program made for a common microprocessor is unable to perform. A software implementation of an ANN can be made with Pros and Cons. Pros: A neural network can perform tasks that a linear program can not; When an element of the neural network fails, it can continue without any problem by their parallel nature; A neural network learns and does not need to be reprogrammed; It can be implemented in any application; It can be implemented without any problem. Cons: The architecture of a neural network is different from the architecture of microprocessors therefore needs to be emulated; it requires high processing time for large neural networks; and the neural network needs training to operate. Three possibilities of training exist: Supervised learning: the network is trained providing input and matching output patterns; Unsupervised learning: input patterns are not a priori classified and the system must develop its own representation of the input stimuli; Reinforcement Learning: intermediate form of the above two types of learning, the learning machine does some action on the environment and gets a feedback response from the environment. Two TRIGAN ANN applications are considered: control rod position and fuel temperature. The outcome obtained in this

  8. TRIGA control rod position and reactivity transient Monitoring by Neural Networks

    Energy Technology Data Exchange (ETDEWEB)

    Rosa, R.; Palomba, M.; Sepielli, M. [ENEA - Casaccia TRIGA Reactor (Italy)

    2008-10-29

    Plant sensors drift or malfunction and operator actions in nuclear reactor control can be supported by sensor on-line monitoring, and data validation through soft-computing process. On-line recalibration can often avoid manual calibration or drifting component replacement. DSP requires prompt response to the modified conditions. Artificial Neural Network (ANN) and Fuzzy logic ensure: prompt response, link with field measurement and physical system behaviour, data incoming interpretation, and detection of discrepancy for mis-calibration or sensor faults. ANN (Artificial Neural Network) is a system based on the operation of biological neural networks. Although computing is day by day advancing, there are certain tasks that a program made for a common microprocessor is unable to perform. A software implementation of an ANN can be made with Pros and Cons. Pros: A neural network can perform tasks that a linear program can not; When an element of the neural network fails, it can continue without any problem by their parallel nature; A neural network learns and does not need to be reprogrammed; It can be implemented in any application; It can be implemented without any problem. Cons: The architecture of a neural network is different from the architecture of microprocessors therefore needs to be emulated; it requires high processing time for large neural networks; and the neural network needs training to operate. Three possibilities of training exist: Supervised learning: the network is trained providing input and matching output patterns; Unsupervised learning: input patterns are not a priori classified and the system must develop its own representation of the input stimuli; Reinforcement Learning: intermediate form of the above two types of learning, the learning machine does some action on the environment and gets a feedback response from the environment. Two TRIGAN ANN applications are considered: control rod position and fuel temperature. The outcome obtained in this

  9. On the Rod Drop technique in integral reactivity measures in control banks and reactor safety; Sobre a tecnica de Rod Drop em medidas de reatividade integral em bancos de controle e seguranca de reatores

    Energy Technology Data Exchange (ETDEWEB)

    Stefani, Giovanni Laranjo

    2013-07-01

    This work presents a study on the effect of shading in neutron detectors, when used in measures of reactivity with the rod drop technique. Shading can be understood as a change in the efficiency of the detectors, when it is given in detected neutrons fission occurred in the reactor, more evident in the detectors closest to the bank being inserted. The method of analysis was based on simulations of reactor IPEN/MB-01, using the code CITATION and MCNP program. In both cases, the results were static, showing Neutronic flows in only two situations: before insertion of the control rod and after insertion. The measure of reactivity in this case was achieved using the expression derived from the source jerk technique. In addition to theoretical study, data from a rod drop experiment conducted in the reactor IPEN/MB-01 were also used. In this case, the reactivity was obtained using inverse kinetic method, since experimental data were set of values that vary with time. In all cases, correction factors for the shadowing effect have been proposed. (author)

  10. Analysis of a control rod ejection accident in a 900 MWe PWR recycling plutonium with a gray control mode

    International Nuclear Information System (INIS)

    This research thesis addresses the study of the control rod cluster ejection accident in a 900 MWe PWR recycling plutonium and operating in grey mode, a class-IV accident in the safety report, which results from the failure of the cluster mechanism pressure enclosure, and results in a quick introduction of a reactivity within the core, and then in a violent power transient during which fuel strength can be put into question again. Two aspects are thus notably addressed: plutonium recycling, and grey mode operation. The objective is to qualitatively and quantitatively assess the evolution of physical parameters during the accident in order to determine the most severe scenarios and to be able to assess the severity of consequences. The author first studies all possible scenarios by means of a 2D+1D+0D calculation scheme in order to determine the most penalizing ones. Then, he develops a precise calculation based on 3D steady calculations, neutron kinetics calculations and thermal kinetics calculations in order to study the previously retained scenarios

  11. Common cause failure analysis of hydraulic scram and control rod systems in the Swedish and Finnish BWR plants

    International Nuclear Information System (INIS)

    The main task of the project included the analysis of the operating experiences at the BWRs of ABB Atom design, comprising 9 units in Sweden and 2 in Finland. International experience and reference information were also surveyed. A reference application was done for the Barsebaeck plant. This pilot study covered all systems which contribute to the reactor shutdown, including also the actuation relays at the interface to the reactor protection system. The Common Load Model was used as the quantification method, which proved to be a practicable approach. This method provides a consistent handling of failure combinatorics and workable extension to evaluate localized dependence between adjacent control rod and drive assemblies (CRDAs). As part of this project, instructions of handbook style were prepared for the CCF analysis of high redundancy systems. The primary focus in the analysis of operating experience was placed on the scram valves and CRDAs. Due to the limited component population, the experiences for the scram valve constitute only a few single failures and some potential but none actual CCF event. These insights are compatible with the generic data for these valves. The experiences for the CRDAs include several single failures, and some actual and many potential CCF events of varying degree of functional impact. Special emphasis was placed to identify any multiple failure or degradation indicating that adjacent rods would be more vulnerable to failure, because such phenomena are far more critical for the scram function as compared to failure of randomly placed rods. 17 refs

  12. Getting into shape: How do rod-like bacteria control their geometry?

    OpenAIRE

    Amir, Ariel; Teeffelen, Sven van

    2014-01-01

    Rod-like bacteria maintain their cylindrical shapes with remarkable precision during growth. However, they are also capable to adapt their shapes to external forces and constraints, for example by growing into narrow or curved confinements. Despite being one of the simplest morphologies, we are still far from a full understanding of how shape is robustly regulated, and how bacteria obtain their near-perfect cylindrical shapes with excellent precision. However, recent experimental and theoreti...

  13. A Calculation of the radioactivity induced in PWR cluster control rods with the origin and casmo codes

    International Nuclear Information System (INIS)

    The radioactivity induced in PWR cluster control rods during reactor operation has been calculated using the computer programme ORIGEN. Neutron fluxes and spectrum conditions as well as the strongly shielded cross sections for the absorber materials Ag, In and Cd have been obtained by running the cell and assembly code CASMO for a couple of typical cases. The results show that Ag-110m, Fe-55 and Co-60 give the largest activity contributions in the interval 1-10 years after the end of irradiation, and Ni-63 and Cd-113m in a longer time perspective. (author)

  14. Nondestructive and Destructive Examination Studies on Removed-from-Service Control Rod Drive Mechanism Penetrations

    Energy Technology Data Exchange (ETDEWEB)

    Cumblidge, Stephen E.; Crawford, Susan L.; Doctor, Steven R.; Seffens, Rob J.; Schuster, George J.; Toloczko, Mychailo B.; Harris, Robert V.; Bruemmer, Stephen M.

    2007-06-07

    Studies conducted at the Pacific Northwest National Laboratory (PNNL) in Richland, Washington, focused on assessing the effectiveness of nondestructive examination (NDE) techniques for inspecting control rod drive mechanism (CRDM) nozzles and J-groove weldments. The primary objectives of this work are to provide information to the U.S. Nuclear Regulatory Commission (NRC) on the effectiveness of NDE methods as related to the in-service inspection of CRDM nozzles and J-groove weldments and to enhance the knowledge base of primary water stress corrosion cracking (PWSCC) through destructive characterization of the CRDM assemblies. Two CRDM assemblies were removed from service, decontaminated, and then used in a series of NDE and destructive examination (DE) measurements; this report addresses the following questions: 1) What did each NDE technique detect? 2) What did each NDE technique miss? 3) How accurately did each NDE technique characterize the detected flaws? 4) Why did the NDE techniques perform or not perform? Two CRDM assemblies including the CRDM nozzle, the J-groove weld, buttering, and a portion of the ferritic head material were selected for this study. This report focuses on a CRDM assembly that contained suspected PWSCC, based on in-service inspection data and through-wall leakage. The NDE measurements used to examine the CRDM assembly followed standard industry techniques for conducting in-service inspections of CRDM nozzles and the crown of the J-groove welds and buttering. These techniques included eddy current testing (ET), time-of-flight diffraction ultrasound, and penetrant testing. In addition, laboratory-based NDE methods were employed to conduct inspections of the CRDM assembly with particular emphasis on inspecting the J-groove weld and buttering. These techniques included volumetric ultrasonic inspection of the J-groove weld metal and visual testing via replicant material of the J-groove weld. The results from these NDE studies were used to

  15. Determination of the Effectiveness of Control Rods in the VVER Reactor Fuel Assemblies

    International Nuclear Information System (INIS)

    The paper describes experiments done in homogeneous mock-ups of the fuel assemblies from the VVER Reactor (at one level of enrichment) to determine the effectiveness of absorbing systems comprising shim fuel assemblies or water cavities and of absorbing rods clad in jackets made of differing materials. The paper also gives data on some experiments that have been done in mock-ups of assemblies with differing levels of enrichment. These experiments make it possible to verify the methods used in calculation and to evaluate the prospects of using them for heterogeneous reactors. (author)

  16. Development of in-vessel type control rod drive mechanism for a innovative small reactor (Contract research)

    Energy Technology Data Exchange (ETDEWEB)

    Yoritsune, Tsutomu; Ishida, Toshihisa [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-03-01

    Although the control rod drive mechanism of an existing large scale light water reactor is generally installed outside the reactor vessel, an in-vessel type control rod drive mechanism (INV-CRDM) is installed inside the reactor vessel. The INV-CRDM contributes to compactness and simplicity of the reactor system, and it can eliminate the possibility of a rod ejection accident. Therefore, INV-CRDM is an important technology adopted in an innovative small reactor. Japan Atomic Energy Research Institute (JAERI) has developed this type of CRDM driven by an electric motor, which can work under high temperature and high pressure water for the advanced marine reactor. On the basis of this research result, a driving motor coil and a bearing were developed to be used under the high temperature steam, severe condition for an innovative small reactor. About the driving motor, we manufactured the driving motor available for high temperature steam and carried out performance test under room temperature atmosphere to confirm the electric characteristic and coolability of the driving coil. With these test results and the past test results under high temperature water, we analyzed and evaluated the electric performance and coolability of the driving coil under high temperature steam. Concerning bearing, we manufactured the test pieces using some candidate material for material characteristic test and carried out the rolling wear test under high temperature steam to select the material. Consequently, we confirmed that performance of the driving coil for the advanced type driving motor, is enough to be used under high temperature steam. And, we evaluated the performance of the bearing and selected the material of the bearing, which can be used under high temperature steam. From these results, we have obtained the prospect that the INV-CRDM can be used for an innovative small reactor under steam atmosphere could be developed. (author)

  17. The Controlled Synthesis of Carbon Tubes and Rods by Template-Assisted Twin Polymerization

    Directory of Open Access Journals (Sweden)

    Falko Böttger-Hiller

    2013-01-01

    Full Text Available The application of porous carbon is versatile. It is used for high-performance catalyst support, electrode material in batteries, and gas storage. In each of these application fields nanostructuring improves the material properties. Supercapacitors store a high energy density. Exactly adapted carbon structures increase the life of lithium batteries and protect catalysts with increasing reaction rate and selectivity. Most of porous carbon materials have a spherical shape. To the best of our knowledge, there is no procedure to synthesize nanostructured cylindrical porous carbon systematically. Here, template glass fibres and SiO2-tubes were modified with nanostructured SiO2/phenolic resin and SiO2/poly(furfuryl alcohol layers by surface twin polymerization (TP of 2,2′-spirobi[4H-1,3,2-benzodioxasiline] and tetrafurfuryloxysilane. Afterwards the SiO2/polymer layer on the template is thermally transformed into a defect-free nanostructured SiO2/carbon layer. After completely removing the SiO2 components microporous carbon tubes or rods are finally achieved. The diameters of the carbon rods and the inner as well as the outer diameter of the carbon tubes are adjustable according to the shape and size of the template. Thus, a huge variety of microporous carbon materials can be easily produced by template-assisted TP.

  18. Disposal of waste channels and control rods and radioactive waste; Gestion de canales usados y barras de control como residuos radiactivos

    Energy Technology Data Exchange (ETDEWEB)

    Lopez Alvarez, L.

    2015-07-01

    Iberdrola and ENRESA are jointly developing a project for the characterization and conditioning of around 200 control rods and 70 used channel from Cofrentes Nuclear Power Plant. This treatment line for high level waste with a radiologic inventory that avoids using the El Cabril low level waste repository is new in Spain and incorporates specific features like the option to carry on with the conditioning stage prior to having a licensed package and available storage facility for this type of waste. (Author)

  19. Design and manufacture of an ultrasonic inspection device for the friction welds in reactor vessel control rod drive mechanism housings

    International Nuclear Information System (INIS)

    The control rod drive mechanism housings of a PWR reactor vessel consist of a stainless steel flange and a Ni-Cr-Fe alloy tube, assembled by friction welding. The properties of the interface and the nature of the adjacent materials require the development of a specific ultrasonic inspection technique which could be easily automated, considering the number of parts involved (77 parts per 1300 MWe reactor vessel). The part has the general shape of a tube (inside diameter: 70 mm, outside diameter: 103 mm). The transition between both forged parent materials (stainless steel/Ni-Cr-Fe alloy) is obtained by a very thin interface, whose general orientation is normal to the tube centerline. The heat affected zone has generally a coarser and more irregular structure than that observed in the parent materials. The design and development were carried out using a prototype machine on test-pieces representative of a control rod drive mechanism housing, and containing the following artificial reflectors: notches obtained by electro-discharge machining on the inside and outside surfaces, on each side of the interface; planar artificial defects, parallel to the interface. These defects, obtained from 2 flat bottomed holes, drilled into the mock-up constituent parts, were conveyed to the interface during friction welding

  20. Study of temperature distribution of fuel, clad and coolant in the VVER-1000 reactor core during group-10 control rod scram by using diffusion and point kinetic methods

    International Nuclear Information System (INIS)

    In this paper, through the application of two different methods (point kinetic and diffusion), the temperature distribution of fuel, clad and coolant has been studied and calculated during group-10 control rod scram, in the Bushehr Nuclear Power Plant (Iran) with a VVER-1000 reactor core. In the reactor core of Bushehr NPP, 10 groups of control rods are used of which, group-10 control rods contain the highest amount of injected negative reactivity in terms of quantity as compared to other groups of control rods. In this paper we explain impacts of negative reactivity, caused by a complete or minor scram of group-10 control rods, on thermoneutronic parameters of the VVER-1000 nuclear reactor core. It should be noted that through these calculations and by using the results, we can develop a sound understanding of impacts of this controlling element in optimum control of the reactor core and, on this basis, with careful attention and by gaining access to a reliable simulation (on the basis of results of calculations made in this survey) we can monitor the VVER-1000 reactor core through a smart control system. In continuation, for a more accurate survey and for comparing results of different calculation systems (point kinetic and diffusion), by using COSTANZA-R,Z calculation code (in which neutronic calculations are based on diffusion model) and using WIMS code at different areas and temperatures (for calculation of constant physical coefficients and temperature coefficients needed in COSTANZAR, Z code) for the VVER-1000 reactor core of Bushehr NPP, calculation of temperature distribution of fuel elements and coolant by using diffusion model is made in the course of group-10 control rods scram and afterwards. (author)

  1. The Sloping Scram Time Simulation of the Nuclear Reactor Control Rod%核反应堆控制棒倾斜落棒时程模拟

    Institute of Scientific and Technical Information of China (English)

    许珍; 白立新; 代飞

    2015-01-01

    事故情况下控制棒组件会在自身重力作用下下落,下落时间是衡量核电站安全运行的重要参数之一。通过建立落棒物理模型,分析落棒的物理过程以及控制棒下落过程中的各阶段的受力情况,得到了落棒控制方程。采用VS2010软件对控制方程数值求解,完成了控制棒组件竖直及倾斜下落过程的时程的分析,对比了倾斜落棒角度对落棒时程的影响,得出最大倾斜落棒角度,为反应堆控制棒组件的设计提供依据。%Control rod assembly will drop because of its own gravity under accident conditions , and the dropping time is one of the important parameters of the safe operation of nuclear power plant .The mathematical model of rod dropping process is established by analyzing the physical process of rod drop and the force at each stage .U-sing VS2010 to solve control equation , the vertical and inclined dropping time of the control rod component and how the angle of tilt rod drop influences the dropping time have been analyzed .The largest angle of tilt rod drop is obtained , which can provide reference for the design of the reactor control rod components .

  2. Benchmark analysis of the NUREC code with OECD/NEA and U.S.NRC PWR MOX/UO{sub 2} control rod ejection problem

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Hyun Chul; Yoo, Jae Woon; Noh, Jae Man; Joo, Hyung Kook [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2004-07-01

    The NUREC code has been developed based on the refined AFEN method for the analysis of LWR cores with mixed-oxide (MOX) fuel. The code was verified against the NEACRP-L336 MOX benchmark problem and the experimental data of YeongKwang Unit 3 and 4. The transient calculation capability of the code was also tested against the NEACRP-L335 rod ejection problem proposed by Finnemann. However, the core in the rod ejection problem was composed of only UO{sub 2} fuels. In this paper, the NUREC code was verified against the OECD/NEA and U.S.NRC PWR MOX/UO{sub 2} control rod ejection problem which was proposed recently by comparing its results with those of the U.S.NRC PARC code. This benchmark problem employed many characteristics of the NEACRP-L335 rod ejection problem but some complexities were added to model a rod ejection accident in a core fueled partially with weapons grade MOX. Some subroutines of the NUREC code were modified to model the transient initiated by a rod ejection in a core loaded with MOX fuels. They include subroutines for reading the cross sections, interpolating the cross sections, treating the delayed neutron fractions, and treating the thermal conductivity of the fuel.

  3. Absorber materials, control rods and designs of shutdown systems for advanced liquid metal fast reactors. Proceeding of a technical committee meeting

    International Nuclear Information System (INIS)

    Thirty-five specialists from France, Germany, India, Japan, the Republic of Kazakhsan, the Russian Federation and the Republic of Georgia (observer) attended the meeting. The meeting had seven sessions. The main topics of discussions were: Status of control rod designs for fast reactors and experience with operation; properties and behaviour of absorber materials for control rods; results of post-irradiation examination of absorber materials, and mechanisms affecting their properties and behaviour; design of a backup reactivity shutdown system utilizing passive mechanisms: Curie point electromagnetic mechanism; enhancement of thermal expansion of absorber rdo drive lines; hydraulically suspended control rods; gas expansion modules in the core; and the possibility of optimizing the reactivity coefficients and the efficiency of Pu burning by using absorber and moderator materials in the core. A total of 23 papers were presented, and a technical tour of the IPPE also took place. Refs, figs, tabs

  4. Feasibility study of the university of Utah TRIGA reactor power upgrade - Part I: Neutronics-based study in respect to control rod system requirements and design

    Directory of Open Access Journals (Sweden)

    Ćutić Avdo

    2013-01-01

    Full Text Available We present a summary of extensive studies in determining the highest achievable power level of the current University of Utah TRIGA core configuration in respect to control rod requirements. Although the currently licensed University of Utah TRIGA power of 100 kW provides an excellent setting for a wide range of experiments, we investigate the possibility of increasing the power with the existing fuel elements and core structure. Thus, we have developed numerical models in combination with experimental procedures so as to assess the potential maximum University of Utah TRIGA power with the currently available control rod system and have created feasibility studies for assessing new core configurations that could provide higher core power levels. For the maximum determined power of a new University of Utah TRIGA core arrangement, a new control rod system was proposed.

  5. Analysis of coolability of the control rods of a Savannah River Site production reactor with loss of normal forced convection cooling

    International Nuclear Information System (INIS)

    An analytical study of the coolability of the control rods in the Savannah River Site (SRS) K-Production Reactor under conditions of loss of normal forced convection cooling has been performed. The study was performed as part of the overall safety analysis of the reactor supporting its restart. The analysis addresses the buoyancy-driven flow over the control rods that occurs when forced cooling is lost, and the limit of critical heat flux that sets the acceptance criteria for the study. The objective of the study is to demonstrate that the control rods will remain cooled at powers representative of those anticipated for restart of the reactor. The study accomplishes this objective with a very tractable simplified analysis for the modest restart power. In addition, a best-estimate calculation is performed, and the results are compared to results from sub-scale scoping experiments. 5 refs

  6. IAEA benchmark calculations on control rod withdrawal test performed during PHENIX end-of-life experiments. JAEA's calculation results

    International Nuclear Information System (INIS)

    This paper describes details of the IAEA/CRP benchmark calculation by JAEA on the control rod withdrawal test in the Phenix End-of-Life Experiments. The power distribution deviation by the control rod insertion/withdrawal, which is the major target of the benchmark, is well simulated by calculation. In addition to the CRP activities, neutron and photon heat transport effect is evaluated in the nuclear heating calculation of the benchmark analysis. It is confirmed that the neutron and photon heat transport effect contributes to the improvement of the absolute power calculation results in the breeder blanket region. (author)

  7. Benchmark of Atucha-2 PHWR RELAP5-3D control rod model by Monte Carlo MCNP5 core calculation

    International Nuclear Information System (INIS)

    Atucha-2 is a Siemens-designed PHWR reactor under construction in the Republic of Argentina. Its geometrical complexity and peculiarities require the adoption of advanced Monte Carlo codes for performing realistic neutronic simulations. Therefore core models of Atucha-2 PHWR were developed using MCNP5. In this work a methodology was set up to collect the flux in the hexagonal mesh by which the Atucha-2 core is represented. The scope of this activity is to evaluate the effect of obliquely inserted control rod on neutron flux in order to validate the RELAP5-3DC/NESTLE three dimensional neutron kinetic coupled thermal-hydraulic model, applied by GRNSPG/UNIPI for performing selected transients of Chapter 15 FSAR of Atucha-2. (authors)

  8. Benchmark of Atucha-2 PHWR RELAP5-3D control rod model by Monte Carlo MCNP5 core calculation

    Energy Technology Data Exchange (ETDEWEB)

    Pecchia, M.; D' Auria, F. [San Piero A Grado Nuclear Research Group GRNSPG, Univ. of Pisa, via Diotisalvi, 2, 56122 - Pisa (Italy); Mazzantini, O. [Nucleo-electrica Argentina Societad Anonima NA-SA, Buenos Aires (Argentina)

    2012-07-01

    Atucha-2 is a Siemens-designed PHWR reactor under construction in the Republic of Argentina. Its geometrical complexity and peculiarities require the adoption of advanced Monte Carlo codes for performing realistic neutronic simulations. Therefore core models of Atucha-2 PHWR were developed using MCNP5. In this work a methodology was set up to collect the flux in the hexagonal mesh by which the Atucha-2 core is represented. The scope of this activity is to evaluate the effect of obliquely inserted control rod on neutron flux in order to validate the RELAP5-3D{sup C}/NESTLE three dimensional neutron kinetic coupled thermal-hydraulic model, applied by GRNSPG/UNIPI for performing selected transients of Chapter 15 FSAR of Atucha-2. (authors)

  9. Preliminary analysis of control rod accidents in the CRCN-R1 multipurpose reactor core of Recife in Brazil

    International Nuclear Information System (INIS)

    The paper shows some results of the neutronic accident analyses arisen by uncontrolled control rod withdrawal, based on the Conceptual Project of the CRCN-R1 MultiPurpose Reactor of Recife. In that reactor, a project of the CNEN/Brazil, under the leadership of the IPEN/Sao Paulo, is verified the thermal hydraulic limits in the reactor core during transients that simulate startup and power operation accidents. It has utilized a computer program that solved the kinetic equations based on multigroup diffusion theory, in our case we have used 4 energy groups, Two-Dimensional X-Y in the space, and 6 groups of delayed neutrons. A simple model of feedback is admitted in the capture and scattering macroscopic cross sections, in the fuel regions, temperature and coolant densities dependents. Based on those models, the results demonstrated that the reactor exhibits good degree of safety. (author)

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

    International Nuclear Information System (INIS)

    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

  11. Study for on-line system to identify inadvertent control rod drops in PWR reactors using ex-core detector and thermocouple measures

    Energy Technology Data Exchange (ETDEWEB)

    Souza, Thiago J.; Medeiros, Jose A.C.C.; Goncalves, Alessandro C., E-mail: tsouza@nuclear.ufrj.br, E-mail: canedo@lmp.ufrj.br, E-mail: alessandro@nuclear.ufrj.br [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear

    2015-07-01

    Accidental control rod drops event in PWR reactors leads to an unsafe operating condition. It is important to quickly identify the rod to minimize undesirable effects in such a scenario. In this event, there is a distortion in the power distribution and temperature in the reactor core. The goal of this study is to develop an on-line model to identify the inadvertent control rod dropped in PWR reactor. The proposed model is based on physical correlations and pattern recognition of ex-core detector responses and thermocouples measures. The results of the study demonstrated the feasibility of an on-line system, contributing to safer operation conditions and preventing undesirable effects, as its shutdown. (author)

  12. Rhodopsin kinase and arrestin binding control the decay of photoactivated rhodopsin and dark adaptation of mouse rods.

    Science.gov (United States)

    Frederiksen, Rikard; Nymark, Soile; Kolesnikov, Alexander V; Berry, Justin D; Adler, Leopold; Koutalos, Yiannis; Kefalov, Vladimir J; Cornwall, M Carter

    2016-07-01

    Photoactivation of vertebrate rhodopsin converts it to the physiologically active Meta II (R*) state, which triggers the rod light response. Meta II is rapidly inactivated by the phosphorylation of C-terminal serine and threonine residues by G-protein receptor kinase (Grk1) and subsequent binding of arrestin 1 (Arr1). Meta II exists in equilibrium with the more stable inactive form of rhodopsin, Meta III. Dark adaptation of rods requires the complete thermal decay of Meta II/Meta III into opsin and all-trans retinal and the subsequent regeneration of rhodopsin with 11-cis retinal chromophore. In this study, we examine the regulation of Meta III decay by Grk1 and Arr1 in intact mouse rods and their effect on rod dark adaptation. We measure the rates of Meta III decay in isolated retinas of wild-type (WT), Grk1-deficient (Grk1(-/-)), Arr1-deficient (Arr1(-/-)), and Arr1-overexpressing (Arr1(ox)) mice. We find that in WT mouse rods, Meta III peaks ∼6 min after rhodopsin activation and decays with a time constant (τ) of 17 min. Meta III decay slows in Arr1(-/-) rods (τ of ∼27 min), whereas it accelerates in Arr1(ox) rods (τ of ∼8 min) and Grk1(-/-) rods (τ of ∼13 min). In all cases, regeneration of rhodopsin with exogenous 11-cis retinal is rate limited by the decay of Meta III. Notably, the kinetics of rod dark adaptation in vivo is also modulated by the levels of Arr1 and Grk1. We conclude that, in addition to their well-established roles in Meta II inactivation, Grk1 and Arr1 can modulate the kinetics of Meta III decay and rod dark adaptation in vivo. PMID:27353443

  13. Design, Fabrication, and Characteristic Experiment of a Hybrid Electromagnet for Bottom-mounted Control Rod Drive Mechanism

    Energy Technology Data Exchange (ETDEWEB)

    Huh, Hyung; Lee, Jin-Haeng; Yoo, Yeon-Sik; Cho, Yeong-Garp; Ryu, Jeong-Soo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    A control rod drive mechanism (CRDM) is located in the reactor pool top (Top-mounted) or a reactivity control mechanism room under the reactor pool bottom (Bottom-mounted). The function of the CRDM is to insert, withdraw, or maintain neutron absorbing material at any required position in the reactor core in order to keep reactivity control of the core. There are so many kinds of CRDMs, such as magnetic-jack type, hydraulic type, rack and pinion type, chain type, and linear or rotary step motor and so on. As a part of a new project, we have completed the design, fabrication, and characteristic experiment of the prototype bottom-mounted CRDM (BMCRDM). The measured carrying capacity of proto-type hybrid electromagnet is approximately 2.8 (kgf) larger than that of 3D-FEM result. The major reasons of the disagreement between the measured and calculated results are as follows. A. B-H Curve differences of ferromagnetic materials B. Fabrication tolerance and the measured maximum temperature at the center of winding for proto-type hybrid electromagnet, 106 .deg. C, appeared to be 5 .deg. C higher than the analytical result. The major reasons of the disagreement between the measured and calculated results are as follows. A. Difficult for exact modeling of winding including impregnated epoxy, coil insulator, and isolator.

  14. Design, Fabrication, and Characteristic Experiment of a Hybrid Electromagnet for Bottom-mounted Control Rod Drive Mechanism

    International Nuclear Information System (INIS)

    A control rod drive mechanism (CRDM) is located in the reactor pool top (Top-mounted) or a reactivity control mechanism room under the reactor pool bottom (Bottom-mounted). The function of the CRDM is to insert, withdraw, or maintain neutron absorbing material at any required position in the reactor core in order to keep reactivity control of the core. There are so many kinds of CRDMs, such as magnetic-jack type, hydraulic type, rack and pinion type, chain type, and linear or rotary step motor and so on. As a part of a new project, we have completed the design, fabrication, and characteristic experiment of the prototype bottom-mounted CRDM (BMCRDM). The measured carrying capacity of proto-type hybrid electromagnet is approximately 2.8 (kgf) larger than that of 3D-FEM result. The major reasons of the disagreement between the measured and calculated results are as follows. A. B-H Curve differences of ferromagnetic materials B. Fabrication tolerance and the measured maximum temperature at the center of winding for proto-type hybrid electromagnet, 106 .deg. C, appeared to be 5 .deg. C higher than the analytical result. The major reasons of the disagreement between the measured and calculated results are as follows. A. Difficult for exact modeling of winding including impregnated epoxy, coil insulator, and isolator

  15. CEA contribution to the analysis of the control rod withdrawal test performed during PHENIX end-of-life experiments. IAEA common research program

    International Nuclear Information System (INIS)

    In 2007 the IAEA, within the framework of the Technical Working Group on Fast Reactors (TWG-FR) activities, decided to launch a Coordinated Research Project (CRP), devoted to benchmarking analyses on 'Control Rod Withdrawal Test' performed during the 'PHENIX End-of-Life Experiments'. This test program was conducted by the CEA, EDF and AREVA before the final shutdown of the prototype power fast reactor PHENIX in order to gather important data and knowledge about several aspects of the operation and safety of pool-type sodium-cooled fast reactors. The overall CRP objective was to improve the participants' analytical capabilities in various fields of research and design of sodium-cooled fast reactors. Among the accident sequences that are to be taken into account, inadvertent withdrawal of a control rod is considered. During operation at nominal power, such a sequence induces a general power increase and local deformations of the power shape. Afterwards, the local fuel temperature increases can thereby lead to fuel melting and clad failure. The quasi-static control rod withdrawal test was especially designed to gather power local data on fissile sub-assemblies and to complete validation databases of neutronic codes. The maximal deformation of the power shape reached ±12% and was obtained when two control rods were shifted in opposite directions. After a description of the test and the measurement methods, this paper presents some results obtained in the course of the test with special emphasis on control rod efficiencies and power deformation by subassemblies. This paper also discusses CEA results obtained in the course of the benchmark with the European neutronic code used for fast reactors design, ERANOS-2.2. (author)

  16. Implementation of a methodology to perform the uncertainty and sensitivity analysis of the control rod drop in a BWR

    Energy Technology Data Exchange (ETDEWEB)

    Reyes F, M. del C.

    2015-07-01

    A methodology to perform uncertainty and sensitivity analysis for the cross sections used in a Trace/PARCS coupled model for a control rod drop transient of a BWR-5 reactor was implemented with the neutronics code PARCS. A model of the nuclear reactor detailing all assemblies located in the core was developed. However, the thermohydraulic model designed in Trace was a simple model, where one channel representing all the types of assemblies located in the core, it was located inside a simple vessel model and boundary conditions were established. The thermohydraulic model was coupled with the neutronics model, first for the steady state and then a Control Rod Drop (CRD) transient was performed, in order to carry out the uncertainty and sensitivity analysis. To perform the analysis of the cross sections used in the Trace/PARCS coupled model during the transient, Probability Density Functions (PDFs) were generated for the 22 parameters cross sections selected from the neutronics parameters that PARCS requires, thus obtaining 100 different cases for the Trace/PARCS coupled model, each with a database of different cross sections. All these cases were executed with the coupled model, therefore obtaining 100 different outputs for the CRD transient with special emphasis on 4 responses per output: 1) The reactivity, 2) the percentage of rated power, 3) the average fuel temperature and 4) the average coolant density. For each response during the transient an uncertainty analysis was performed in which the corresponding uncertainty bands were generated. With this analysis it is possible to observe the results ranges of the responses chose by varying the uncertainty parameters selected. This is very useful and important for maintaining the safety in the nuclear power plants, also to verify if the uncertainty band is within of safety margins. The sensitivity analysis complements the uncertainty analysis identifying the parameter or parameters with the most influence on the

  17. Fuel rod

    International Nuclear Information System (INIS)

    The present invention provide a fuel rod used in a BWR type reactor, preventing the occurrence of defects of weld portions and improving the operationability of test and assembling operation to improve the quality of weld portions. Namely, the fuel rod is formed by loading a plurality of fuel pellets in a cladding tube. The outer diameter of a groove portion of a tightly sealing end plug to be inserted and welded to the open end of the cladding tube is made substantially identical with the inner diameter of the cladding tube. A neck portion having a diameter smaller than the outer diameter of the groove portion is disposed between an end plug main body and the groove portion. As a result, since the outer diameter of the groove portion is substantially identical with the inner diameter of the cladding tube, the positioning is facilitated. Since the neck portion having a smaller diameter than the outer diameter of the groove portion is disposed in the groove portion, a gap is formed in the welded portion thereby enabling to facilitate the confirmation of weld sag for confirming integrity of the weld by a non-destructive test. (I.S.)

  18. Modernization project of the rod control system and in-core instrumentation system for 34 units of the 900 MW French EDF fleet

    International Nuclear Information System (INIS)

    Rolls-Royce and Cegelec, in partnership, carry out a unique and considerable modernisation project of two Instrumentation and Control (I and C) systems for the entire 900 MWe fleet of Electricite De France (EDF). Both rod control (RCS) and reactor in-core measurement (RIC) systems are to be modernised in the frame of the third ten-year renovation of all 34 reactor units over 9 power plants. The RCS contributes to the control of nuclear power by actuating control rod drive mechanisms that allow insertion or withdrawal of control rods. The RCS has also monitoring functions such as controlling the actual rods' position as well as the functional consistency between commands and actual positions. The RIC system measures in-core neutron flux, providing useful information to the control room as well as to the reactor unit computer for further processing. The renovated systems shall replace the existing ageing analog technology by modern digital technology based on PLC (Programmable Logic Controllers) and FPGA (Field-Programmable Gate Array) in the case of power subassemblies of RCS. Both systems rely for certain functions on a common network linking the RCS and RIC networks, improving operations and maintenance thanks to a powerful Man Machine Interface at the different locations of the systems with an extensive suite of tools and diagnostic menus. The project whose design phase started in July 2006 is now in its deployment phase after the successful site implementation of both systems at the first of kind units of Tricastin and Fessenheim power plants, respectively in August 2009 and February 2010. With 20 units in operation in 2014, the deployment shall continue with the other 14 until 2020. Rolls-Royce has a broad range of civil nuclear expertise, including work related to licensing and safety reviews, engineering design, supply chain management, manufacturing, installation and commissioning of the nuclear island systems and equipment, as well as operational

  19. The Tested Control Rod Drive Mechanism System based on PLC%基于PLC的控制棒驱动机构复验

    Institute of Scientific and Technical Information of China (English)

    汪明珠; 范祖光; 解苑明; 吴军

    2011-01-01

    In order to test control rod drive mechanism would be match the functions and performance of the requirements of the design specifications, and complete the factory test, design this test device for test drive mechanism. Hie device based on programmable logic controller realized rod lifting, rod keeping, rod inserting and integrated alarm etc. Function. It introduces the working principle and realization method of the test device of control rod drive mechanism. Through connecting debugging with driving mechanism, The result proved that the test device would be match the requirements of the technical agreement, the function and performance of CRDM meet the requirements of the design specifications of the production.%为了检验控制棒驱动机构的功能和性能是否满足设计规范书的要求,完成驱动机构的出厂试验,设计了控制棒驱动机构试验装置.基于可编程逻辑控制器(以下简称PLC)的控制棒驱动机构试验装置实现了控制棒的提升、保持、下降和报警综合等功能.文中介绍了控制棒驱动机构试验装置的原理和实现方法,通过和驱动机构的联调试验,证明控制棒驱动机构试验装置符合技术协议的要求,驱动机构的功能和性能满足设计规范书的要求.

  20. A methodology for obtaining the control rod patterns in a BWR using genetic algorithms

    International Nuclear Information System (INIS)

    In this work the GACRP system based on the genetic algorithms technique for the obtaining of the drivers of control bars in a BWR reactor is presented. This methodology was applied to a transition cycle and a one of balance of the Laguna Verde nuclear power station (CNLV). For each one of the studied cycles, it was executed the methodology with a fixed length of the cycle and it was compared the effective multiplication factor of neutrons at the end of the cycle that it is obtained with the proposed drivers of control bars and the multiplication factor of neutrons obtained by means of a Haling calculation. It was found that it is possible to extend several days the length of both cycles with regard to the one Haling calculation. (Author)

  1. Controlled cooling technology for bar and rod mills -- Computer simulation and operational results

    Energy Technology Data Exchange (ETDEWEB)

    Mauk, P.J.; Kruse, M.; Plociennik, U. [SMS Schloemann-Siemag AG, Dusseldorf (Germany)

    1995-09-01

    The Controlled Cooling Technology (CCT) developed by SMS to simulate the rolling process and automatic control of the water cooling sections is presented. The Controlled Rolling and Cooling Technology (CRCT) model is a key part of the CCT system. It is used to simulate temperature management for the rolling stock on the computer before the actual rolling process takes place. This makes it possible to dispense with extensive rolling tests in the early stages of project planning and to greatly reduce the extent of such tests prior to the start of commercial production in a rolling mill. The CRCT model has been in use at Von Moos Stahl Ag for three years. It demonstrates that, by targeted improvement of the set-up values in both the technology and the plant, it is possible to improve microstructure quality and achieve better geometrical parameters in the rolled products. Also, the results gained with the CCT system in practical operation at the Kia Steel Bar Mill, Kunsan, Korea, are presented.

  2. Preliminary Test of a small heat pipe for hybrid control rod in-core passive decay heat removal system

    International Nuclear Information System (INIS)

    This paper introduces 'Hybrid control rod' combining its original function and heat removal ability. The high temperature operation and high resistance of radiation should be considered to adopt the hybrid heat pipe at the in-core condition. Other design consideration is to make extra inlet parts because it has a high risk of inlet boundary failure. It means that the introduction of heat pipe system is difficult to present nuclear power plants. The other concepts are presented to out-core cooling design but it has low performance compared with in-core heat removal system. Hybrid heat pipe for in-core heat removal system suggests the solution of these problems. Ultimate objective of this research is to develop the passive emergency decay heat removal system using hybrid heat pipes targeting design bases accidents such as station black-out (SBO) and small break loss of coolant accident (SBLOCA). The purpose of this work is to confirm the performance and heat transfer behavior of hybrid heat pipe. The hybrid heat pipe has special condition for operation. Therefore, it is hard to analyze their behavior in core. Table I shows the characteristics of hybrid heat pipe and consideration for manufacturing the heat pipe

  3. Laser Ultrasonic System for Surface Crack Visualization in Dissimilar Welds of Control Rod Drive Mechanism Assembly of Nuclear Power Plant

    Directory of Open Access Journals (Sweden)

    Yun-Shil Choi

    2014-01-01

    Full Text Available In this paper, we propose a J-groove dissimilar weld crack visualization system based on ultrasonic propagation imaging (UPI technology. A full-scale control rod drive mechanism (CRDM assembly specimen was fabricated to verify the proposed system. An ultrasonic sensor was contacted at one point of the inner surface of the reactor vessel head part of the CRDM assembly. Q-switched laser beams were scanned to generate ultrasonic waves around the weld bead. The localization and sizing of the crack were possible by ultrasonic wave propagation imaging. Furthermore, ultrasonic spectral imaging unveiled frequency components of damage-induced waves, while wavelet-transformed ultrasonic propagation imaging enhanced damage visibility by generating a wave propagation video focused on the frequency component of the damage-induced waves. Dual-directional anomalous wave propagation imaging with adjacent wave subtraction was also developed to enhance the crack visibility regardless of crack orientation and wave propagation direction. In conclusion, the full-scale specimen test demonstrated that the multiple damage visualization tools are very effective in the visualization of J-groove dissimilar weld cracks.

  4. Preliminary Test of a small heat pipe for hybrid control rod in-core passive decay heat removal system

    Energy Technology Data Exchange (ETDEWEB)

    Kim, In Guk; Ban, In Cheol [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

    2014-05-15

    This paper introduces 'Hybrid control rod' combining its original function and heat removal ability. The high temperature operation and high resistance of radiation should be considered to adopt the hybrid heat pipe at the in-core condition. Other design consideration is to make extra inlet parts because it has a high risk of inlet boundary failure. It means that the introduction of heat pipe system is difficult to present nuclear power plants. The other concepts are presented to out-core cooling design but it has low performance compared with in-core heat removal system. Hybrid heat pipe for in-core heat removal system suggests the solution of these problems. Ultimate objective of this research is to develop the passive emergency decay heat removal system using hybrid heat pipes targeting design bases accidents such as station black-out (SBO) and small break loss of coolant accident (SBLOCA). The purpose of this work is to confirm the performance and heat transfer behavior of hybrid heat pipe. The hybrid heat pipe has special condition for operation. Therefore, it is hard to analyze their behavior in core. Table I shows the characteristics of hybrid heat pipe and consideration for manufacturing the heat pipe.

  5. Insertion anomaly of control rod assemblies in nuclear reactors; Anomalies d'insertion des grappes de commande des reacteurs electronucleaires

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-10-06

    Some malfunction have been noticed in the insertion of the control rod assemblies in some PWR reactors of Electricite de France (EdF). These anomalies (insufficient speed, incomplete insertion) are due to a deformation of the most irradiated fuel assemblies. A prevention and monitoring program has been implemented by EdF, first in 1997 on its 1300 MW plants, and extended in 1999 to its 900 MW plants. The progress of this program is controlled by the French authority of nuclear safety (ASN) and the preventive and surveillance measures (falling time, rebound, and insertion stress measurements of control rods) are described in the appendixes of this document for the 900 MWe, 1300 MWe and 1450 MWe reactors, respectively. (J.S.)

  6. Analysis of Boron Depletion in HTR Control Rod%高温气冷堆控制棒硼燃耗特性分析

    Institute of Scientific and Technical Information of China (English)

    赵晶; 李富; 刘志宏; 石秀安

    2011-01-01

    控制棒价值及其燃耗规律是核反应堆物理设计关注的要点之一.球床式高温气冷堆控制棒位于侧反射层石墨孔道中,吸收体为圆环形的B4C,其燃耗特性具有特殊性.采用MCNP耦合燃耗计算模块的方法,对控制棒吸收体进行精细划分,分析了各子区域硼的详细燃耗特性及控制棒价值的变化规律.计算结果表明,由于强烈的空间自屏效应,虽然吸收体外层硼燃耗很多,但吸收体内层硼燃耗很少,因此,反应堆运行寿期末控制棒价值减少很小.%The worth of control rod and its change along depletion are one of important features for reactor neutronics design. The control rods in pebble bed high-temperature gas-cooled reactor (HTR) were located in the hole of side graphite reflector. Its absorber was made up of annular B4C, which had special depletion rules. The depletion characteristics in each absorber sub-zone and the changes of the control rod worth were analyzed, based on both the MCNP coupling another depletion module and the detailed modeling control rod absorber region. According to the analysis, the boron in inner region of absorber is depleted only a little, although most of boron in outer region of absorber is depleted, because of the strong space self-shielding effects in control rod absorber. Therefore, the worth of control rod in the HTR decreases just a little at the end of reactor life time.

  7. Microstructural characterization and properties of dissimilar joints used in coupling of PWR control rod driving

    International Nuclear Information System (INIS)

    The chemical, mechanical and microstructural characterizations of a dissimilar joint between SA336F347 austenitic and SA479Tp414 martensitic stainless steels were done, welded by TIG process, defining as a result of this characterization that the ER Ni Cr-3 Ni consumable seems to be the best applicable consumable compared to the ER309L consumable; The main variables of the process control were also evaluated, its weldability and properties for a future qualification of a welding procedure, besides to simulate possible situations to be found in this type of joint, such as, its weldability by the LASER process, welded joint without filler metal and without shielding gas, obtaining in this way enough data for the production of products that contains this type of joint. (author)

  8. A Fast Guide Tube Position Estimation Algorithm for a Control Rod Support Pin Inspection Robot

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae C.; Jeon, Hyeong S.; Choi, Yu R.; Kim, Jae H. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2005-07-01

    The risk that PWR guide tube support pins will crack has increased the necessity for the development of inspection methods and equipment. A special remote-controlled manipulator has been widely used to inspect the guide tube support pins. We presented a matched filter algorithm for detecting the existence and estimating the position of the guide tube support pins. But, the matched filter algorithm requires numbers of complex floating point calculations for the 2-D FFT and therefore it can not be fitted in to the small-sized embedded processors. We proposed a new simplified method for estimating the position of the guide tube support pins. It uses most of the operations with integers. We ported the proposed method in intel's xscale processor running at 400 mhz. We used gnu C language in embedded linux operating system. We can calculate the algorithm at a rate of 20 frames/sec. in a 160x120 image size.

  9. The whole process control of connecting rod production line equipment management%浅谈连杆生产线设备管理全过程控制

    Institute of Scientific and Technical Information of China (English)

    杨萍

    2013-01-01

    The connecting rod is one of the most basic and core components for engine. Combines with the practice of running environment, connecting rod components are effected by alternating load during operation long time and sustained by the force of compression, bending and stretching, thus the need to safeguard the connecting rod parts can have the quality characteristics of smaller, at the same time, has outstanding advantages on the rigidity and strength of the preparation, has the important meaning for the operation of the production line for connecting rod. Based on the actual situation, the equipment management connecting rod production line process control as the research object, to analyze the importance of connecting rod parts, and then combines with the operation rod production line, studies the equipment management measures of the whole process control, hoping to draw special attention to staff.%  连杆是发动机最为基础与核心的组成部件之一。结合实践运行环境来看,连杆部件在运行过程当中长时间且持续性的受到了压缩、弯曲以及拉伸等交变载荷的作用力影响,从而需要保障连杆部件能够具备较小的质量特性,同时在刚度与强度方面具备突出优势,这对于连杆生产线的运行而言也有着重要意义。本文依据这一实际情况,以连杆生产线设备管理全过程控制为研究对象,首先简要分析了连杆部件的重要性,进而结合连杆生产线的作业工序,详细研究了设备管理全过程控制的主要措施,希望能够引起各方工作人员的特别关注与重视。

  10. Comparison of different heuristic techniques for the optimization of control rod patterns

    International Nuclear Information System (INIS)

    Presently work a comparison, using different methodologies, with the results obtained for the optimization of the design of control bar patterns for boiling water reactors is carried out. The results were obtained considering the same conditions for all the used methodologies, which are part of the combinatory optimization, these were programmed in FORTRAN 77 language under the UNIX platform in an ALPHA work station. The techniques used to carry out the optimization are the following ones, Genetic Algorithms, Dispersed Search, Taboo Search, Ants Colonies and Neural Networks. The objective function used is the same one in all the cases, in this its are included the MFLPD thermal limits (maximum fraction of power density to the end of the operation cycle), MPGR (Maxim rate of power generation at the end of the operation cycle) and MFLCPR (maximum fraction for the critical power ratio at the end of the operation cycle), of equal form it is included the criticality condition for the reactor and it is needed the Power Axial Profile in each step of burnt to be adjusted to a proposed Power Axial Profile. The CM-PRESTO code (Scandpower) was used to evaluate the proposed designs. The approach used for the comparison is essentially the keff at the cycle end, as well as the thermal limits performance, nevertheless, it is also analyzed the exchanges number among the shallow and deep positions, and the total number of movements carried out during the complete cycle. (Author)

  11. Chemical State Mapping of Degraded B4C Control Rod Investigated with Soft X-ray Emission Spectrometer in Electron Probe Micro-analysis

    Science.gov (United States)

    Kasada, R.; Ha, Y.; Higuchi, T.; Sakamoto, K.

    2016-05-01

    B4C is widely used as control rods in light water reactors, such as the Fukushima Daiichi nuclear power plant, because it shows excellent neutron absorption and has a high melting point. However, B4C can melt at lower temperatures owing to eutectic interactions with stainless steel and can even evaporate by reacting with high-temperature steam under severe accident conditions. To reduce the risk of recriticality, a precise understanding of the location and chemical state of B in the melt core is necessary. Here we show that a novel soft X-ray emission spectrometer in electron probe microanalysis can help to obtain a chemical state map of B in a modeled control rod after a high-temperature steam oxidation test.

  12. Feasibility study of the university of Utah TRIGA reactor power upgrade - Part I: Neutronics-based study in respect to control rod system requirements and design

    OpenAIRE

    Ćutić Avdo; Choe Dongok; Jevremović Tatjana

    2013-01-01

    We present a summary of extensive studies in determining the highest achievable power level of the current University of Utah TRIGA core configuration in respect to control rod requirements. Although the currently licensed University of Utah TRIGA power of 100 kW provides an excellent setting for a wide range of experiments, we investigate the possibility of increasing the power with the existing fuel elements and core structure. Thus, we have developed numerical models in combination w...

  13. 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.

  14. Calculation of the Phenix end-of-life test “control rod withdrawal” with the ERANOS code

    International Nuclear Information System (INIS)

    The Institute of Radioprotection and Nuclear Safety (IRSN) being established as technical support organization for French public authorities is in charge of safety assessment of both operating and under construction reactors and nuclear facilities. It provides safety studies of advanced and innovative projects like fast sodium cooled reactors as well. In this context, one of the IRSN objectives is to evaluate comprehensively the accuracy of numerical tools and their performance on studies of safety relay items. Reactor physics studies step in the safety assessment support from different points of view, among which the design of core and its protection system. They are essential in the cores behavior analysis in normal, perturbed and accidental conditions in order to assess the integrity of the first barrier and the exclusion of prompt criticality and re-criticality risks. The codes capability to compute in an accurate manner the fission power distribution in the core during the whole reactor lifetime could indicate the codes' accuracy for many so-called spatial dependent values calculations. The IAEA Coordinated Research Project on the Phenix end-of-life test “Control Rod Withdrawal” has been a good opportunity to check the capability of calculation tools by comparison with the measured radial power distributions on fast reactor. IRSN participated to this benchmark with the ERANOS code package developed by CEA for fast reactors studies. The challenge for this code package was that in the considered core configurations the neutron fields were notably deformed. This paper presents the results obtained in the framework of the benchmark activity. A relatively good agreement has been found with available measures considering the approximations done in the modeling. The work underlines the importance of precise knowledge of the details of burn-up distribution as it could impact the calculations of the power distribution. (author)

  15. Coupled thermal-hydraulic and neutronic simulations of Phenix control rod withdrawal tests with SIMMER-IV

    International Nuclear Information System (INIS)

    The “end-of-life” tests performed in the Phenix reactor before its final shutdown in 2009, in particular the Control Rod (CR) withdrawal experiments provide an excellent opportunity for the validation and verification of the reactor physics computer codes and modeling approaches. SIMMER-IV, a modern three-dimensional reactor safety code, has been recently employed at Karlsruhe Institute of Technology (KIT) for simulating Phenix experiments in the framework of a benchmark exercise organized under the IAEA project. In this paper, we report and discuss main results obtained with SIMMER-IV at KIT. Particular attention is devoted to the coupling features of thermal-hydraulics and neutronics and their mutual influences. The reactor reactivity, power and neutron flux distributions calculated with SIMMER-IV are in good agreement both with experimental results and with calculations with advanced neutronics codes, such as ERANOS, while the CR reactivity worth is overestimated due to neglecting heterogeneity effects. Because of its multi-physics capabilities SIMMER also calculates the temperature distributions which are in a good agreement with the experimental test results. In this work we describe the improvements in SIMMER neutronics model by employing a correction that is based on the results of cell calculations performed with ERANOS. The study confirms that the 3D SIMMER-IV code can accurately predict major fast reactor neutronics and thermal hydraulic parameters, provided that a special treatment is employed for CR modeling. The results of calculations are analyzed in frames of SIMMER-IV validation and verification assessment. (author)

  16. A methodology for obtaining the control rods patterns in a BWR using systems based on ants colonies; Una metodologia para obtener los patrones de barras de control en un BWR usando sistemas basados en colonias de hormigas

    Energy Technology Data Exchange (ETDEWEB)

    Ortiz S, J.J. [Depto. de Sistemas Nucleares, ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico); Requena R, I. [Universidad de Granada, 18071 Granada (Spain)]. e-mail: jjortiz@nuclear.inin.mx

    2003-07-01

    In this work the AZCATL-PBC system based on a technique of ants colonies for the search of control rods patterns of those reactors of the Nuclear Power station of Laguna Verde (CNLV) is presented. The technique was applied to a transition cycle and one of balance. For both cycles they were compared the k{sub ef} values obtained with a Haling calculation and the control rods pattern proposed by AZCATL-PBC for a burnt one fixed. It was found that the methodology is able to extend the length of the cycle with respect to the Haling prediction, maintaining sure to the reactor. (Author)

  17. Modernization project of the rod control system and in-core instrumentation system for 34 units of the 900 MW French EDF fleet

    International Nuclear Information System (INIS)

    Rolls-Royce and Cegelec, in partnership, carry out a unique and considerable modernisation project of two Instrumentation and Control (I and C) systems for the entire 900 MWe fleet of Electricite De France (EDF). Both rod control (RCS) and reactor in-core measurement (RIC) systems are to be modernised in the frame of the third ten-year renovation of all 34 reactor units over 9 power plants. The RCS contributes to the control of nuclear power by actuating control rod drive mechanisms that allow insertion or withdrawal of control rods. The RCS has also monitoring functions such as controlling the actual rods' position as well as the functional consistency between commands and actual positions. The RIC system measures in-core neutron flux, providing useful information to the control room as well as to the reactor unit computer for further processing. The renovated systems shall replace the existing ageing analog technology by modern digital technology based on PLC (Programmable Logic Controllers) and FPGA (Field-Programmable Gate Array) in the case of power subassemblies of RCS. Both systems rely for certain functions on a common network linking the RCS and RIC networks, improving operations and maintenance thanks to a powerful Man Machine Interface at the different locations of the systems with an extensive suite of tools and diagnostic menus. The project whose design phase started in July 2006 is now in its deployment phase after the successful site implementation of both systems at the first units of Tricastin and Fessenheim power plants, respectively in August 2009 and February 2010. The deployment shall continue with the other 32 units until 2020. Rolls-Royce has a broad range of civil nuclear expertise, including work related to licensing and safety reviews, engineering design, supply chain management, manufacturing, installation and commissioning of the nuclear island systems and equipment, as well as operational management through life support. Cegelec, with

  18. Automatic safety rod for reactors. [LMFBR

    Science.gov (United States)

    Germer, J.H.

    1982-03-23

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

  19. Common Cause Failure Analysis of Control Rods and Drives in the Swedish and Finnish BWR Plants. Operating Experiences in 1983 - 2003

    Energy Technology Data Exchange (ETDEWEB)

    Mankamo, Tuomas [Avaplan Oy, Espoo (Finland)

    2006-11-15

    The control rod and drives in a Boiling Water Reactor (BWR) constitute a highly redundant system. The reliability of the system is determined by how well the design withstands dependencies, as Common Cause Failures (CCFs). This report upgrades an earlier data collection on CCFs of control rod and drives (SKI Report 1996:77) to more recent years, with the objective to report the data to ICDE project (International Common Cause Failure Data Exchange) and to the safety analysts in the Nordic countries. The operating experiences were analyzed at the BWRs of former Asea-Atom design, comprising 9 units in Sweden and Olkiluoto 1 and 2 in Finland, covering years 1983 - 2003. A new logical scheme was developed to classify interconnected failure modes of the two redundant functions for reactivity shutdown, fast hydraulic insertion and slower screw insertion of control rods. The scheme makes an explicit distinction between the different attributes of the failure event: - affected function - affected movement direction - detectability - criticality, i.e. inoperable control rod function versus only degraded functionality Another novel idea emerged for grouping the events according to generic failure mechanism. The generic classes will help to organize and structure the information efficiently, because in most cases within a class, the failure modes prove to be same, or there are only a few alternatives to chose from. From the set of 72 candidate cases, altogether 27 actual or more significant potential CCFs were screened out. Special emphasis was placed to identify any multiple failure or degradation indicating that adjacent rods would be more vulnerable to failure, because such phenomena are far more critical for reactivity shutdown as compared to failure of randomly placed rods. Only slight tendency of position dependence could be determined. Another positive insight is that the events, where foreign objects caused the jamming of rod insertion, were separated by both

  20. Modeling of primary water stress corrosion cracking at control rod drive mechanism nozzles of pressurized water reactors

    International Nuclear Information System (INIS)

    One of the main failure mechanisms that cause risks to pressurized water reactors is the primary water stress corrosion cracking (PWSCC) occurring in alloys. It can occurs, besides another places, at the control reactor displacement mechanism nozzles. It is caused by the joint effect of tensile stress, temperature, susceptible metallurgical microstructure and environmental conditions of the primary water. These cracks can cause accidents that reduce nuclear safety by blocking the rod's displacement and may cause leakage of primary water, reducing the reactor's life. In this work it is proposed a study of the existing models and a modeling proposal to primary water stress corrosion cracking in these nozzles in a nickel based Alloy 600. It is been superposed electrochemical and fracture mechanics models, and validated using experimental and literature data. The experimental data were obtained at CDTN-Brazilian Nuclear Technology Development Center, in a recent installed slow strain rate testing equipment. In the literature it is found a diagram that indicates a thermodynamic condition for the occurrence of some PWSCC sub modes in Alloy 600: it was used potential x pH diagrams (Pourbaix diagrams), for Alloy 600 in high temperature primary water (300 deg C till 350 deg C). Over it, were located the PWSCC sub modes, using experimental data. It was added a third parameter called 'stress corrosion strength fraction'. However, it is possible to superpose to this diagram, other parameters expressing PWSCC initiation or growth kinetics from other models. Here is the proposition of the original contribution of this work: from an original experimental condition of potential versus pH, it was superposed, an empiric-comparative, a semi-empiric-probabilistic, an initiation time, and a strain rate damage models, to quantify respectively the PWSCC susceptibility, the failure time, and in the two lasts, the initiation time of stress corrosion cracking. It was modeling from our

  1. Introduction to Quality Control of Oil Cylinder Piston Rod%浅谈油缸活塞杆的质量控制

    Institute of Scientific and Technical Information of China (English)

    孙俊玲

    2014-01-01

    The hydraulic cylinder is an important assembly in hydraulic systems ,is widely used in engineering machinery ,and the piston rod is an important part of hyduaulic cylinder .Combining with an example of piston rod failure ,this paper points out the quality of coating on piston rod directly affects the service life and reliability of hydraulic cylinder .The causes for failure of electric plating are analyzed ,the study results can provide a basis for the quality control of coating .%液压缸是液压系统中重要的执行部件,在工程机械中有着广泛的应用,而支持活塞做功的连接部件---活塞杆是液压缸的重要部件。结合一活塞杆电镀层受损实例,阐述了镀层质量的好坏直接影响到整个液压缸的寿命和可靠性,同时分析了电镀层失效的原因,为后续镀层的质量控制提供了依据。

  2. Demonstration of control rod holding stability of the self actuated shutdown system in Joyo for enhancement of fast reactor inherent safety

    International Nuclear Information System (INIS)

    Self actuated shutdown system (SASS) with a Curie point electromagnet (CPEM) has been developed for use in a large-scale liquid metal cooled fast breeder reactor (LMFBR) in order to establish the passive shutdown capability against anticipated transient without scram (ATWS) events. The basic characteristics of SASS have already been investigated by various out-of-pile tests for material elements. As the final stage of the development, the stability of SASS needs to be confirmed under the actual reactor-operational environment with high temperature, high neutron flux, and flowing sodium to ensure the high plant availability factor. For this purpose, the demonstration test of holding stability using the reduced-scale experimental equipment of SASS was conducted in the 1st and 2nd operational cycles of the experimental fast reactor Joyo MK-III. The rod-holding stability and the rod-recovering functions of the driving system to re-connect and pull out the separated control rod were fully confirmed. The results also indicate there is no essential problem for the practical use of SASS about its operational trouble involving the unexpected drop during reactor operation. (author)

  3. Application of IGBT in Rod Control System of Nuclear Power Station%IGBT在核电站棒控系统中的应用

    Institute of Scientific and Technical Information of China (English)

    许育周; 李涛; 王春生

    2014-01-01

    针对目前基于可控硅的核电站棒控系统存在输出电流纹波大、电流上升和下降时间长等特点,提出了基于绝缘栅双模型晶体管( IGBT)的捧控系统。 IGBT作为新型大功率开关器件,具有电压型控制、输入阻抗大、驱动功率小、控制电路简单、开关损耗小、工作频率高和元件容量大等优点。实际应用表明,IGBT减小了棒控系统输出的电流纹波和电流上升、下降时间,十分适合应用于核电站棒控系统中。%At present, most of the rod control system in nuclear power stations are using SCR, so the ripple of output current is big, and the rising time and calling time of the current are long, to overcome this demerits, the rod control system based on IGBT is proposed. IGBT is a new type of large power switching device; it features voltage control, high input impedance, small driving power, simple control circuit, low switching loss, high working frequency and large capacity. Practical application shows that the ripple and current rising and falling time are decreased, this shows that IGBT is suitable for application of rod control system in nuclear power station.

  4. Verification Test of Movable Magnetic Coils Control Rod Drive Line%可动线圈控制棒电磁驱动线验证试验

    Institute of Scientific and Technical Information of China (English)

    张之华; 钱达志; 邓勇军; 薄涵亮; 徐显启; 吴莘馨; 米向秒

    2011-01-01

    The movable magnetic coils control rod drive line, a new kind control rod drive line for reactors, regulates and controls reactor power by driving control rods with electromagnetic force. A series of verification tests were conducted to break through the key technology in development process. The theoretical basis of equipment development was verified by the principle test, making a breakthrough in the feasibility. The stroke test and life test not only verified the stability, reliability of the equipment, but also got its operation characteristic parameters. The anti-seismic test verified the security function in extreme conditions. The verification test data provide reference for the installation, regulation and safe operation of the drive line.%可动线圈控制棒电磁驱动线是新研制的一种反应堆用驱动线,靠电磁力驱动控制棒实现反应堆的开堆、停堆和功率调节.为突破其研制过程中的关键技术,进行了一系列的验证试验.通过原理性能试验,验证了设备研制的理论基础,实现了可行性的突破;通过行程试验和寿命考验,验证了设备的稳定性、可靠性,并得到了设备的运行特性参数;通过抗震试验,验证了设备在极端条件下的安全功能.性能试验得到的试验数据,可为该驱动线的安装、调试、安全运行提供依据.

  5. Nuclear reactor protection system in case of control rod drop. Systeme de protection d'un reacteur nucleaire en cas de chute d'un element antireactif

    Energy Technology Data Exchange (ETDEWEB)

    Bourin, J.M.; Mourlevat, J.L.; Sengler, G.

    1989-07-28

    Protection against the consequences of an accidental drop of a control rod in a nuclear reactor is assured by 4 independent guardlines containing for each 2 neutron flux detectors, 2 primary analysing circuits associated with these 2 detectors each providing a primary signal of rod drop when a rapid decrease of flux is measured by the corresponding detector and a gate providing an output signal in the presence of at least one such primary signal. The detectors are distributed angularly around the vertical axis of the core. A secondary circuit trips the reactor when it receives a rod drop signal from at least 2 guardlines.

  6. Connexin 36 and rod bipolar cell independent rod pathways drive retinal ganglion cells and optokinetic reflexes.

    Science.gov (United States)

    Cowan, Cameron S; Abd-El-Barr, Muhammad; van der Heijden, Meike; Lo, Eric M; Paul, David; Bramblett, Debra E; Lem, Janis; Simons, David L; Wu, Samuel M

    2016-02-01

    Rod pathways are a parallel set of synaptic connections which enable night vision by relaying and processing rod photoreceptor light responses. We use dim light stimuli to isolate rod pathway contributions to downstream light responses then characterize these contributions in knockout mice lacking rod transducin-α (Trα), or certain pathway components associated with subsets of rod pathways. These comparisons reveal that rod pathway driven light sensitivity in retinal ganglion cells (RGCs) is entirely dependent on Trα, but partially independent of connexin 36 (Cx36) and rod bipolar cells. Pharmacological experiments show that rod pathway-driven and Cx36-independent RGC ON responses are also metabotropic glutamate receptor 6-dependent. To validate the RGC findings in awake, behaving animals we measured optokinetic reflexes (OKRs), which are sensitive to changes in ON pathways. Scotopic OKR contrast sensitivity was lost in Trα(-/-) mice, but indistinguishable from controls in Cx36(-/-) and rod bipolar cell knockout mice. Mesopic OKRs were also altered in mutant mice: Trα(-/-) mice had decreased spatial acuity, rod BC knockouts had decreased sensitivity, and Cx36(-/-) mice had increased sensitivity. These results provide compelling evidence against the complete Cx36 or rod BC dependence of night vision's ON component. Further, the findings suggest the parallel nature of rod pathways provides considerable redundancy to scotopic light sensitivity but distinct contributions to mesopic responses through complicated interactions with cone pathways. PMID:26718442

  7. Development of multivariable control for bar and wire rod rolling and control system for close-tolerance bar rolling. Boko/senzai atsuen no tahensu seigyo to seimitsu atsuen system no kaihatsu

    Energy Technology Data Exchange (ETDEWEB)

    Noguchi, Y.; Okamura, K.; Ogai, H.; Baba, K.; Naganuma, Y.; Ishii, H. (Nippon Steel Corp., Tokyo (Japan))

    1992-11-30

    Regarding an application of the modern control theory to the continuous rolling, the multivariable control simulations of the cold rolling mill have been reported so far. Since the requirements for a higher dimensional accuracy of the bar and wire rod products also have been raised by the users today, for corresponding to it, the development of sophisticated dimension control technology for the continuous rolling has been desired. In order to realize the rolling with a higher accuracy, by the authors, using the rolling state equations and the optimal regulator theory, the state equations required for designing the control system for the bar mill, intermediate wire rod mill and wire rod finishing block mill have been established and a multivariable control system has been developed. In addition, at the bar mill in the Muroran steelwork of the Nippon Steel Corp., a precision bar rolling system consisted of the inter-billet dimension control and the in-billet dimension control by this control method has been put into practice. By using it, a stable precision rolling has been realized, and moreover, the automation of rolling and the skillfree operation have been achieved. 19 refs., 13 figs., 2 tabs.

  8. The Eclipse system for surveying the guide tubes of control rod clusters; Systeme Eclipse pour l'expertise des tubes guides de commande de grappes

    Energy Technology Data Exchange (ETDEWEB)

    Pace, Y.M. [Areva NP, 92 - Paris la Defense (France)

    2008-07-15

    Eclipse is a new system developed by Areva to assess the wear of the guide tubes of control rod clusters. This system is based on the projection of a shadow on a light plan in order to record the profile and the internal diameter of a hollow tube. This system allows us to quantify the wear and it can be included in a program dedicated to monitor the wear and master its kinetics. This system has been validated on the guide tubes from the Ringhals units. (A.C.)

  9. Development and field experiences in ultrasonic and eddy current inspection of inaccessible welds in the control rod housings of boiling water reactor vessels

    International Nuclear Information System (INIS)

    The methodology and inspection techniques are described developed by Tecnatom (Spain) for detecting intergranular stress corrosion cracking at welded penetration joints, specifically, the J-shaped stub-tube/control rod housing joint in BWR's. They are based on ultrasound complemented with eddy current testing. A mixed analog-digital data acquisition and processing system is used for the evaluation of results. The use of machines for the said testing showed that the detection of defects was even possible in the above awkward places, and the correlation and repeatability of the results confirmed the reliability of the system. The use of multifrequency probes was shown to be advantageous. (L.O.). 10 figs

  10. Calibrating and Controlling the Quantum Efficiency Distribution of Inhomogeneously Broadened Quantum Rods by Using a Mirror Ball

    DEFF Research Database (Denmark)

    Hansen, Per Lunnemann; Rabouw, Freddy T.; van Dijk-Moes, Relinde J. A.;

    2013-01-01

    near a mirror, not only allows an extraction of calibrated ensemble-averaged rates, but for the first time also to quantify the full inhomogeneous dispersion of radiative and non radiative decay rates across thousands of nanocrystals. We apply the technique to novel ultrastable CdSe/CdS dot......-in-rod emitters. The emitters are of large current interest due to their improved stability and reduced blinking. We retrieve a room-temperature ensemble average quantum efficiency of 0.87 ± 0.08 at a mean lifetime around 20 ns. We confirm a log-normal distribution of decay rates as often assumed in literature...

  11. Investigation of combined free and forced convection in a 2 x 6 rod bundle during controlled flow transients

    Energy Technology Data Exchange (ETDEWEB)

    Bates, J.M.; Khan, E.U.

    1980-10-01

    An experimental study was performed to obtain local fluid velocity and temperature measurements in the mixed (combined free and forced) convection regime for specific flow coastdown transients. A brief investigation of steady-state flows for the purely free-convection regime was also completed. The study was performed using an electrically heated 2 x 6 rod bundle contained in a flow housing. In addition a transient data base was obtained for evaluating the COBRA-WC thermal-hydraulic computer program (a modified version of the COBRA-IV code).

  12. A Study on the Electromagnet Thrust force Characteristics of Newly Proposed Hybrid Bottom-mounted Control Rod Drive Mechanism for Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Huh, Hyung; Cho, Yeong Garp; Choi, Myoung Hwan; Yu, Je Yong; Kim, Ji Ho; Kim, Jong In [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2011-05-15

    The control rod drive mechanism (CRDM) is the part of reactor regulating system (RRS), which is located in the reactor pool top (Top-mounted) or the room below the reactor pool (Bottom-mounted). The function of the CRDM is to insert, withdraw or maintain neutron absorbing material at any required position within the reactor core, in order to the reactivity control of the core. There are so many kinds of CRDM, such as magneticjack type, hydraulic type, rack and pinion type, chain type and linear or rotary step motor and so on. As a part of a new project, we have investigated the movable coil electromagnetic drive mechanism (MCEDM) which is new scheme for the reactor control rod adopted by China Advanced Research Reactor (CARR) as shown in Fig.1. To improve a better function of the electromagnetic and magnetic characteristics, new model CRDM, which is named a hybrid bottommounted CRDM (HBCRDM), is proposed. Especially in order to achieve improved thrust force, numerical magnetic field calculations between MCEDM and HBCRDM have been carried out and the HBCRDM FEM results have been compared with the MCEDM FEM results, and FEM results are summarized in the following sections

  13. CARR控制棒驱动机构堆外调试试验%Out of Pile Commissioning Test of CARR Control Rod Driving Mechanism

    Institute of Scientific and Technical Information of China (English)

    张应超; 高永光; 张明葵; 康亚伦; 季松涛; 黄道立; 马明武

    2013-01-01

    中国先进研究堆(CARR)控制棒由磁力驱动。为进行CARR控制棒堆外调试试验,建造了模拟CARR热工水力条件的试验回路。试验中发现驱动机构存在一些问题,提出改进建议,得到CARR工程部和设计者认可。对驱动机构进行了一些改进,试验测量了线圈温度、落棒时间、极限流量和极限载荷等重要参数,试验证明经改进的控制棒驱动机构达到了设计要求。%China Advanced Research Reactor (CARR) control rods are driven by the magnetic force produced by electrical coils . An out of pile testing loop was specially constructed to simulate CARR thermal-hydraulic conditions for the test .Some problems were found during the experiment ,and several proposals of improvement were offered for the modification of the driving mechanism and accepted by the CARR engineering department and the designer .The modification to the driving mechanism was implemented ,and the important parameters were measured ,such as the balance temperature of the coils ,the rod drop time ,the coolant flow limit ,and the load limit . T he results show that the modified control rod driving mechanism meets design requirements .

  14. Modelling of sucker rod string

    Energy Technology Data Exchange (ETDEWEB)

    Hojjati, M.H. [Mazandaran Univ., (Iran, Islamic Republic of). Dept. of Mechanical Engineering; Lukasiewicz, S.A. [Calgary Univ., AB (Canada). Dept. of Mechanical and Manufacturing Engineering

    2005-12-01

    Rod pumping is used extensively in the oil well industry as a method of artificial lift. In order to analyze the performance of oil wells, the force and displacement at the polished rod are measured using a dynamometer. The data is applied to the boundary conditions when calculating the forces and displacement at the bottom of the rod string that defines the conditions of the pump, pumping effectiveness and production rate. This study proposed a transfer matrix method to model the dynamic behavior of the sucker string rod. The main reason for developing the method was to simplify the currently used mathematical method with a simple matrix operation in which the bottom-hole force-displacement values are obtained as a product of data vectors at the polished rod end by a transfer matrix. The problem was solved using D'Alembert's systems solution equation and the adaptive filter matrix method. The proposed method reduces calculation time because a more efficient matrix operation is used without losing accuracy. This study showed that it is possible to use the transfer matrix to calculate load-displacement relations a hundred or more times in one stroke, which is beneficial when developing tools to control oil wells, such as wellhead controllers. 9 refs., 3 tabs., 8 figs.

  15. Piston rod seal

    Energy Technology Data Exchange (ETDEWEB)

    Lindskoug, S.

    1984-06-05

    In a piston rod seal of the type comprising a gland through which the piston rod is passed the piston is provided with a sleeve surrounding the piston rod and extending axially so as to axially partly overlap the gland when the piston is in its bottom dead center position. 4 figs.

  16. Piston rod seal

    Energy Technology Data Exchange (ETDEWEB)

    Lindskoug, Stefan (Malmo, SE)

    1984-01-01

    In a piston rod seal of the type comprising a gland through which the piston rod is passed the piston is provided with a sleeve surrounding the piston rod and extending axially so as to axially partly overlap the gland when the piston is in its bottom dead center position.

  17. Tie rod insertion test

    CERN Multimedia

    B. LEVESY

    2002-01-01

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

  18. 1987 Sucker rod tables

    Energy Technology Data Exchange (ETDEWEB)

    1987-03-01

    This reference identifies manufacturers qualified to produce API sucker rods and related equipment, lists chemical and mechanical properties of the various types of rods and provides dimensional characteristics. In addition, similar information is given for non-API products such as fiberglass and hollow rods.

  19. 改进的源倍增方法测量控制棒价值%The Control Rod Worth Measurement With Improved Neutron Source Multiplication Method

    Institute of Scientific and Technical Information of China (English)

    史永谦; 李义国; 鲁谨; 洪景彦; 吴小波; 彭旦

    2015-01-01

    该文给出了改进的源倍增方法测量控制棒价值的原理,在高富集度235 U 燃料元件转换为低富集度235 U 的微型中子源零功率反应堆上进行研究,实验测量微型中子源零功率反应堆中心控制棒的价值,与周期方法相比在2%内符合,但减少了测量时间。该方法为今后加速器驱动次临界系统 ADS 的次临界在线监督提供一种可能的方法。%Control rod worth measurement principle with improved neutron source multiplication (INSM)method is described and its worth measurement was completed on the miniature neutron source reactor by this method. Worth of control rod located in the center of the core is in accordance with the value measured by period method within the error 2%,but the measuring time is quite short. INSM can be used to monitor the sub-critical reactivity of accelerator driven sub-critical system (ADS)on line in the future.

  20. 磁悬浮控制棒驱动线性能试验研究%Experimental Study on Performance of Magnetic Levitation Control Rod Driving Line

    Institute of Scientific and Technical Information of China (English)

    张之华; 薄涵亮; 米向秒; 徐显启; 王家英; 张新荣; 吴莘馨

    2012-01-01

    为了测试磁悬浮控制棒驱动线的特性,验证其性能的稳定性和可靠性,进行了驱动线的综合性能研究.通过运行特性试验,得到控制棒驱动机构及驱动线的静态、动态性能参数;通过运行寿命考验,验证控制棒驱动线的稳定性、可靠性,并对其综合性能进行检验.%To test the performance of magnetic levitation control rod driving line and verify the stability and reliability, the comprehensive study on drive line is conducted. The operational performance test gives out the static and dynamic parameters of the driving system and driving line. The operational life test proves the stability and reliability of the control rod driving line and verifies the comprehensive performance.

  1. Sucker rod construction

    Energy Technology Data Exchange (ETDEWEB)

    Anderson, R.A.; Goodman, J.L.; Tickle, J.D.; Liskey, A.K.

    1987-03-31

    A sucker rod construction is described comprising: a connector member being formed to define a rod receptacle having a closed axially inner end and an open axially outer end, the rod receptacle having axially spaced, tapered annular surfaces, a cylindrical fiberglass rod having an end having an outer surface being received within the rod receptacle through the outer end and cooperating therewith to define an annular chamber between the outer surface of the end of the rod and the tapered annular surfaces, and a bonding means positioned in the annular chamber for bonding to the outer surface of the end of the rod to confront the tapered annular surfaces, each annular surface having an angle of taper with respect to the outer surface of the fiberglass rod, and each angle of taper being progressively and uniformly less toward the open end by an amount between one and one-half degrees and two degrees, inclusive, and a collet connected to the connector member adjacent the open axially outer end of the rod receptacle and having an axial bore therethrough retaining the end of the rod in coaxial position within the rod receptacle.

  2. Sucker rod guide

    Energy Technology Data Exchange (ETDEWEB)

    Edwards, B.J.; Starks, J.A.

    1989-08-22

    This patent describes a sucker rod guide for mounting on a sucker rod and spacing the sucker rod from the tubing in an oil well. The guide comprising a generally cylindrically-shaped, extruded, ultra-high density polyethylene body having a substantially smooth outside surface; a longitudinal bore provided centrally of the body. The bore having a smaller diameter than the diameter of the sucker rod; a plurality of grooves provided in circumferential relationship in the bore; and a tapered slot extending longitudinally through the body from the outside surface to the bore. The tapered slot further comprising a slot mouth located at the outside surface and a slot throat spaced from the slot mouth. The slot throat lying adjacent to the sucker rod bore and wherein the slot throat is wider than the slot mouth for mounting the sucker rod guide on the sucker rod.

  3. Turbodrill rod angular velocity indicator

    Energy Technology Data Exchange (ETDEWEB)

    Rogachev, O.K.; Belozerova, L.P.; Konenkov, A.K.

    1984-01-01

    This paper outlines shortcomings of existing types of telemetry systems which resulted in production of the IChT-1 unit. Unit is intended for control of angular velocity of serially produced turbodrill rods, during drilling of wells up to 5000 m deep, and bottomhole temperatures to 100C. The paper provides a detailed description and diagrams for installing this unit.

  4. Fuel rod welding (LWBR development program)

    International Nuclear Information System (INIS)

    Procedures were developed to weld both ends of approximately 25,000 fuel rods for the Light Water Breeder Reactor (LWBR) core. The rods were welded using the gas tungsten arc (GTA) method in high-purity helium at 1 atmosphere. Welding parameters, including weld current, arc gap, and speed of rotation, were established to control the size of the weld. Electrode and chill positioning with respect to the endclosure/tube joint controlled the location of the weld. Weld quality of the fuel rods was ensured by 100-percent nondestructive testing by ultrasonic and radiographic inspection and the destructive evaluation of process control samples in each weld lot

  5. Low fluid level in pulse rod shock absorber

    International Nuclear Information System (INIS)

    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)

  6. Fuel rod bowing

    International Nuclear Information System (INIS)

    The purpose of this investigation was to quantify the extent of fuel rod bowing in Westinghouse pressurized water reactors and to assess the effects of fuel rod bowing on plant safety and reliability. An empirical bow correlation was developed based on data from irradiated assemblies. Analyses conducted with these conservative empirical predictions show that: (1) generically identified DNBR margins are adequate to offset DNBR reductions due to rod bow, (2) the present design practice of increasing the highest calculated core peaking factor is sufficient to account for all deviations, including the effects of rod bow, and (3) fretting and corrosion of bowed rods are negligible. These conclusions indicate that fuel rod bowing results in no impact on plant safety or reliability

  7. Ionic nitriding as remedy for control rod elements wear; La nitruration ionique, remede a l`usure des crayons de grappes de commande

    Energy Technology Data Exchange (ETDEWEB)

    Hertz, D.; Peyran, J.C. [Societe Franco-Americaine de Constructions Atomiques (FRAMATOME), 92 - Paris-La-Defense (France)

    1994-12-31

    As soon as the reactor control mode enabling grid follow was implemented, FRAMATOME focussed on the impact of RCCA motion on the wear resistance of the RCCA rod lets. This led to the choice of the ionitriding process from a range of candidate surface treatments. By means of process adaptations, the wear resistance of austenitic stainless steels is enhanced with no significant degradation of their corrosion performance. Apart from the surface treatment principle, this paper outlines the wear and corrosion characteristics of the product, together with the experience feedback from the roughly 350 RCCA`s manufactured in 1993 and more than 200 RCCA`s loaded into reactors, the first of which have exceeded 5 cycles of irradiation. (authors). 2 figs., 1 tab., 5 refs.

  8. Control assembly materials for water reactors: Experience, performance and perspectives. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    The safe, reliable and economic operation of water cooled nuclear power reactors depends to a large extent upon the reliable operation of control assemblies for the regulation and shutdown of the reactors. These consist of neutron absorbing materials clad in stainless steel or zirconium based alloys, guide tubes and guide cards, and other structural components. Current designs have worked extremely well in normal conditions, but less than ideal behaviour limits the lifetimes of control materials, imposing an economic penalty which acts as a strong incentive to produce improved materials and designs that are more reliable. Neutron absorbing materials currently in use include the ceramic boron carbide, the high melting point metal hafnium and the low melting point complex alloy Ag-In-Cd. Other promising neutron absorbing materials, such as dysprosium titanate, are being evaluated in the Russian Federation. These control materials exhibit widely differing mechanical, physical and chemical properties, which must be understood in order to be able to predict the behaviour of control rod assemblies. Identification of existing failure mechanisms, end of life criteria and the implications of the gradual introduction of extended burnup, mixed oxide (MOX) fuels and more complex fuel cycles constitutes the first step in a search for improved materials and designs. In the early part of this decade, it was recognized by the International Working Group on Fuel Performance and Technology (IWGFPT) that international conferences, symposia and published reviews on the materials science aspects of control assemblies were few and far between. Consequently, the IWGFPT recommended that the IAEA should rectify this situation with a series of Technical Committee meetings (TCMs) devoted entirely to the materials aspects of reactor control assemblies. The first was held in 1993 and in the intervening five years considerable progress has been made. In bringing together experts in the

  9. Validation of the ultrasonic and Eddy current techniques to inspect the accommodation of the elements of (CRDH) control rod drive; Validacion de las tecnicas de ultrasonidos y corrientes inducidas para inspeccionar los alojamientos

    Energy Technology Data Exchange (ETDEWEB)

    Garcia, A.; Gomez, P.; Sanchez, J.; Resa, P.

    2013-07-01

    Tecnatom development in the past with ultrasonic inspection procedures to examine vessels BWR of several Central nuclear (CRDH) control rod drive elements, accommodations. In each case, inspection techniques have relied on both the volume of required test postulated defects. Also, taking into account the possible access to the component, developed mechanical equipments of different characteristics.

  10. Control of the saturation temperature in magnetic heating by using polyethylene-glycol-coated rod-shaped nickel-ferrite (NiFe2O4) nanoparticles

    Science.gov (United States)

    Iqbal, Yousaf; Bae, Hongsub; Rhee, Ilsu; Hong, Sungwook

    2016-02-01

    Polyethylene-glycol (PEG)-coated nickel-ferrite nanoparticles were prepared for magnetic hyperthermia applications by using the co-precipitation method. The PEG coating occurred during the synthesis of the nanoparticles. The coated nanoparticles were rod-shaped with an average length of 16 nm and an average diameter of 4.5 nm, as observed using transmission electron microscopy. The PEG coating on the surfaces of the nanoparticles was confirmed from the Fourier-transform infrared spectra. The nanoparticles exhibited superparamagnetic characteristics with negligible coercive force. Further, magnetic heating effects were observed in aqueous solutions of the coated nanoparticles. The saturation temperature could be controlled at 42 ℃ by changing the concentration of the nanoparticles in the aqueous solution. Alternately, the saturation temperature could be controlled for a given concentration of nanoparticles by changing the intensity of the magnetic field. The Curie temperature of the nanoparticles was estimated to be 495 ℃. These results for the PEG-coated nickel-ferrite nanoparticles showed the possibility of utilizing them for controlled magnetic hyperthermia at 42 ℃.

  11. A Study About Inversion Performance of Magnetic Suspension Control Rod Drive Mechanism%磁悬浮式控制棒驱动机构静水转向性能研究

    Institute of Scientific and Technical Information of China (English)

    朱学微; 甄建霄; 王玉林; 焦迪楠

    2014-01-01

    在中国先进研究堆( CARR)调试阶段,发现其控制棒驱动机构在转向过程中存在驱动线圈与控制棒运动不同步的现象。为掌握控制棒驱动机构的转向性能参数,在静水条件下进行了控制棒驱动机构转向性能测试试验。试验表明:驱动线圈与控制棒运动不同步是磁悬浮式控制棒驱动机构的固有特性;棒位、棒速和驱动线圈通电电流大小等因素对静水条件下转向性能影响不大,且四组驱动机构静水条件下转向性能具有一致性。%During the commissioning phase of China Advanced Research Reactor ( CARR) , operator found the a-synchronism between drive coils and control rod when the control rod drive mechanism ( CRDM ) taking inver-sion.With the aim of acquiring inversion performance of the CRDM , a CRDM inversion performance test has been taken under the condition of static coolant .The result of test shows that asynchronism between drive coils and control rod is the intrinsic characteristic of magnetic suspension CRDM ,the factors , such as control rod po-sition, control rod speed and drive coil current , have limited influence on static coolant inversion performance , and the inversion performance of four groups CRDM have uniformity under static coolant condition .

  12. KINIK, Absorber Rod Calibration Kinetics

    International Nuclear Information System (INIS)

    1 - Description of program or function: KINIK is an inverse kinetic code that solves the inverse form of the point kinetic equations using the Runge-Kutta method. An optimization procedure is involved to control the time step and to reduce the running time. Up to 24 delayed neutron groups of different types (in case of heavy water as moderator or beryllium as reflector) are considered. KINIK is commonly applied to determine reactivity worths and to calibrate absorber rods. Following a rod drop, neutron flux or power is recorded as a function of time and used as input. 2 - Method of solution: The inverse point kinetic equations are numerically solved for each time step using the Runge-Kutta method. The input data resulting from measurements are first approximated by polynomials of maximum degree 10 using a least-squares approach

  13. Flexible sucker rod unit

    Energy Technology Data Exchange (ETDEWEB)

    Allen, L.F.

    1987-02-03

    This patent describes a deep well having: a. an education tube with an inside diameter extending from the surface of the earth to far below the surface, b. a reciprocating pump housing attached to the bottom of the education tube, c. pump jack means at the surface for reciprocating the pump, d. a light sucker rod connected to the pump jack means and extending into the education tube, and e. a series of heavy sinker bars having a large cross sectional area in the education tube connecting the light sucker rod to the pump; f. an improved integral metal flexible rod unit interconnecting the sinker bars comprising in combination with the above: g. a coupling on each end of the integral metal flexible rod unit connecting the flexible rod unit to the contiguous sinker bar, h. a segment which is flexible as compared to the sinker bars connecting one of the couplings to i. an integral metal bearing adjacent to the other of the couplings, the bearing having j. a cylindrical surface with k. a diameter i. only slightly smaller than the inside diameter of the education tube thereby forming a sliding fit therewith, and ii. greater than the diameter of any other portion of the flexible rod unit and the sinker bar, and l. grooves in the cylindrical surface for the passage of fluid between in the education tube around the bearing.

  14. Ultrasonic Phased Array Assessment of the Interference Fit and Leak Path of the North Anna Unit 2 Control Rod Drive Mechanism Nozzle 63 with Destructive Validation

    Energy Technology Data Exchange (ETDEWEB)

    Crawford, Susan L.; Cinson, Anthony D.; MacFarlan, Paul J.; Hanson, Brady D.; Mathews, Royce

    2012-08-01

    The objective of this investigation was to evaluate the efficacy of ultrasonic testing (UT) for primary water leak path assessments of reactor pressure vessel (RPV) upper head penetrations. Operating reactors have experienced leakage when stress corrosion cracking of nickel-based alloy penetrations allowed primary water into the annulus of the interference fit between the penetration and the low-alloy steel RPV head. In this investigation, UT leak path data were acquired for an Alloy 600 control rod drive mechanism nozzle penetration, referred to as Nozzle 63, which was removed from the North Anna Unit 2 reactor when the RPV head was replaced in 2002. In-service inspection prior to the head replacement indicated that Nozzle 63 had a probable leakage path through the interference fit region. Nozzle 63 was examined using a phased-array UT probe with a 5.0-MHz, eight-element annular array. Immersion data were acquired from the nozzle inner diameter surface. The UT data were interpreted by comparing to responses measured on a mockup penetration with known features. Following acquisition of the UT data, Nozzle 63 was destructively examined to determine if the features identified in the UT examination, including leakage paths and crystalline boric acid deposits, could be visually confirmed. Additional measurements of boric acid deposit thickness and low-alloy steel wastage were made to assess how these factors affect the UT response. The implications of these findings for interpreting UT leak path data are described.

  15. Analysis of unprotected transients with control and safety rod drive mechanism expansion feedback in a medium sized oxide fuelled fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sathiyasheela, T., E-mail: sheela@igcar.gov.in; Natesan, K.; Srinivasan, G.S.; Devan, K.; Puthiyavinayagam, P.

    2015-09-15

    Highlights: • Possibilities of enhancing safety under ULOF and UTOP accidents. • CSRDM expansion feedbacks under unprotected transients. • CSRDM expansion feedback enhances the safety of fast reactors. • CSRDM expansion feedbacks ensuring enough time for initiating safety actions. - Abstract: Possibilities of enhancing core safety under unprotected loss of flow (ULOF) and unprotected transient over power (UTOP) accidents with control and safety rod drive mechanism (CSRDM) expansion feedbacks are explored in a medium sized oxide fuelled fast breeder reactor. This feedback is expected to take the reactor to a safe shutdown under ULOF and to an another steady state under UTOP where there is no significant fuel melting. Under ULOF, with CSRDM feedback net reactivity was maintained negative throughout the transient (up to 2000 s) and the power dropped to a level of heat removal capacity of decay heat removal system based on natural circulation. Similarly, under UTOP with the above feedback reactor power goes to a lower peak value. The fuel temperature is just touching the melting temperature and the melt fraction does not cross 5%. With CSRDM expansion feedbacks both ULOF and UTOP transients prolong beyond 2000 s. It ensures, availability of time for initiating any safety actions against the transients, and thus it helps to preclude core disruptive accidents (CDA) in a medium sized oxide fuelled reactors.Classification: L. safety and risk analysis.

  16. Effects of Repair Weld of Reactor Pressure Vessel Upper Head Control Rod Drive Mechanism Penetration Nozzle on J-Groove Weldment Using Finite Element Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Ju Hee; Yoo, Sam Hyeon [Korea Military Academy, Seoul (Korea, Republic of); Kim, Yun Jae [Korea University, Seoul (Korea, Republic of)

    2014-06-15

    In pressurized water reactors, the upper head of the reactor pressure vessel (RPV) contains numerous control rod drive mechanism (CRDM) nozzles. These nozzles are fabricated by welding after being inserted into the RPV head with a room temperature shrink fit. The tensile residual stresses caused by this welding are a major factor in primary water stress corrosion cracking (PWSCC). Over the last 15 years, the incidences of cracking in alloy 600 CRDM nozzles have increased significantly. These cracks are caused by PWSCC and have been shown to be driven by the welding residual stresses and operational stresses in the weld region. Various measures are being sought to overcome these problems. The defects resulting from the welding process are often the cause of PWSCC acceleration. Therefore, any weld defects found in the RPV manufacturing process are immediately repaired by repair welding. Detailed finite element simulations for the Korea Nuclear Reactor Pressure Vessel were conducted in order to predict the magnitudes of the repair weld residual stresses in the tube materials.

  17. Rod worth measurement innovation at Westinghouse

    International Nuclear Information System (INIS)

    Bank worth measurement of control rods and shut-down rods is required for every cycle startup of a nuclear power plant for design validation. For Pressurized Water Reactors (PWR), the bank worth measurement is part of the Low Power Physics Tests (LPPT) program. In almost all instances, this program is on critical path during ascension to power. There is a strong incentive for the utility industry to have a fast and reliable method of measuring the bank worth. Over the past decade, Westinghouse has been developing new advanced rod worth measurement methods to provide faster, safer, more accurate and easier to use products. The advancement of 3D core simulation codes has made it possible to make revolutionary developments for a new generation of rod worth measurement methods

  18. Analysis of Control Rod Worth Features of MC Simulation in AP1000 Reactor%AP1000核反应堆控制棒价值特性的MC模拟

    Institute of Scientific and Technical Information of China (English)

    谢明亮; 于雷; 陈玉清

    2016-01-01

    针对当前AP000堆芯采用的两类控制棒束,基于MCNP5程序建立堆芯仿真计算模型,分析了含不同硼浓度对堆芯kef与硼微分价值的影响,同时对AP1000棒组价值进行模拟计算,对比分析了黑棒与灰棒插入堆芯对kef的影响。结果表明:基于MC N P5程序建立的模型是正确的,硼微分价值(绝对值)随硼浓度增加呈现下降趋势,其值在-9.16~-13.60范围内变化,符合反应性设计要求,有效增殖系数kef随控制棒插入呈现非线性变化,得到了控制棒的价值变化曲线与拟合关系式,为控制棒在反应堆内紧急控制与功率调节提供参考。%In view of current use of two kinds of control rods in AP1000 reactor,the simulation calculation model was set up based on MCNP5 code,kef and differential value of boron with several different boron concentration were analyzed in the reactor,and simulation of group value rods of AP1000 were calculated at the same time,and the contrast analysis of kef effects was accomplished when black and grey rod insert-ed into the core.Results show that:the model based on MCNP5 code is correct,and the differential value of boron (absolute value)present a downward trend with the increase of boron concentration,and its value changed between 9.16 and 13.60,which is conforms to the requirements of reactive design,and the effec-tive multiplication factor kef present nonlinear variation with rod inserted into the core,and it gets the curves of control rod worth and fitting relation,which provides the reference for emergency control and power regulation with control rod in the reactor.

  19. Study on Anti-Seismic Test of Control Rod Driving System Suspended by Magnetic Force%磁悬浮控制棒驱动线抗震试验研究

    Institute of Scientific and Technical Information of China (English)

    张之华; 钱达志; 张征明; 吴莘馨; 徐显启; 黄洪文; 胡晓

    2012-01-01

    To verify the stability, reliability and security function in extreme conditions, the anti-seismic test of control rod drive line was conducted. Drop-time of control rod drive line in different earthquake intensities was got. The response and strain values of control rod drive line acceleration on SL-1, SL-2 level were measured. Safety functions of control rod drive line were validated in different work conditions. Anti-seismic test data shows that the driving system can keep the structure's integrality and realize operation function under OBE and SSE.%为验证设备的稳定性、可靠性以及在极端条件下的安全功能,在地震模拟振动台上,采用一组控制棒驱动线实物作为足尺模型,进行了控制棒驱动线的抗震试验研究.得到了不同的地震输入对控制棒驱动线落棒时间的影响;测量了运行安全地震( SL-1)、极限安全地震(SL-2)水平下,控制棒驱动线的加速度响应值和应变值;验证了不同工况下控制棒驱动线的安全功能.试验数据表明,该驱动线在运行基准地震( OBE)、安全停堆地震(SSE)工况下,均能保持结构的完整性,并能实现运行功能.

  20. Improvement to the pattern of control rods of the equilibrium cycle of 18 months for the CLV using bio-inspired algorithms

    International Nuclear Information System (INIS)

    Nowadays in the National Institute of Nuclear Research are carried out studies with some bio-inspired optimization techniques to improve the performance of the fuel cycles of the boiling water reactors of the Laguna Verde power plant (CLV). In the present work two bio-inspired techniques were applied with the purpose of improving the performance of a balance cycle of 18 months developed for the CLV: genetic algorithms (AG) and systems based on ants colonies (SCH). The design of the reference cycle it represents in several aspects an optimal cycle proposed starting from the experience of several operation decades with the boiling water reactors (BWR initials for Boiling Water Reactor) in the world. To try to improve their performance is beforehand a difficult challenge and it puts on test the feasibility of the optimization methods in the reloads design. The study of the bio-inspired techniques was centered exclusively on the obtaining of the control rod patterns (PBC) trying to overcome the capacity factor reached in the design of the reference cycle. It was fixed the cycle length such that the decrease of the coast down period would represent an increase of the capacity factor of the cycle; so that, it diminishes the annual cost associated with the capital cost of the plant. As consequence of the study, was found that the algorithm based on the ants colonies reaches to diminish the coast down period in five and half days respect to the original balance cycle, what represents an annual saving of $US 74,000. Since the original cycle was optimized, the above-mentioned, shows the ability of the SCH for the optimization of the cycle design. With the AG it was reach to approach to the original balance cycle with a coast down period greater in seven days estimating an annual penalization of $US 130,000. (Author)

  1. Control rod cluster drop time anomaly Guandong nuclear power station (Daya bay) and Electricite de France nuclear power stations (1450 MWe N4 Series); Anomalie de temps de chute des grappes de controle centrale de guang dong (daya bay) et centrales d`electricite de France (Palier N4-1450 MWE)

    Energy Technology Data Exchange (ETDEWEB)

    Olivera, J.J.; Naury, S.; Tricot, N.; Tran Dai, P.; Gama, J.M.

    1996-12-31

    The anomaly of control rod cluster drop time revealed at Guandong Nuclear Power Station in Daya Bay and in the Chooz B1 pilot unit for the N4 series, led to the replacement of the M1 type control rod cluster guide tubes with 1300 MWe PWR type guide tubes, adapted to the geometry of the Guandong reactors and the 1450 MWe reactors of the N4 series. The comparison of the drop times obtained with the 1300 MWe type control rod cluster guide 1300 MWe type control rod cluster guide tubes gave satisfactory results. These met the safety criterion for N4 series control rod cluster drop times (2.15 under hot shutdown conditions). The drop time tests which will be carried out in middle of and at the end of cycle 1 of Chooz B1 should make it possible to finally validate the solution already successfully implemented at Guandong. However, this anomaly has revealed the limits of representativeness of the experimental test loops with regard to the real reactor configuration. In view of this, it has been deemed necessary to ask Electricite de France to pursue its analysis both on the understanding of the phenomena which led to this anomaly and on the limits of the representativeness of the experimental test loops. (authors).

  2. Disassembly of steel elements of the Stade nuclear power station. Disposal of control rod guide thimbles from KKI 1; Zerlegung von Stahlelementen aus dem KKW Stade. Entsorgung von Steuerelementfuehrungsrohren aus dem KKI 1

    Energy Technology Data Exchange (ETDEWEB)

    Friske, A.; Radzuweit, J. [GNS Gesellschaft fuer Nuklear-Service mbH, Essen (Germany); Stechmann, L. [E.ON Kernkraft, Stade (Germany). Kernkraftwerk Stade; Schwarz, W. [E.ON Kernkraft, Essenbach (Germany). Kernkraftwerk Isar 1

    2008-01-15

    In the course of decommissioning and dismantling PWR plants of older lines, activated steel elements arise, in addition to the usual core components, which had been installed in fuel element positions of the core edge region to protect the reactor pressure vessel from neutron-induced embrittlement. Twelve steel elements had to be conditioned for waste management within the decommissioning and disassembly phase of the Stade nuclear power station (KKS). The steel rods were dismantled from the skeletons under water, cut up, and loaded into MOSAIK {sup registered} inserts provided for this purpose. The packaging concept also included provisions for later loading of the skeleton parts into the MOSAIK {sup registered} inserts. In a second phase, the steel element skeletons were disassembled with underwater shears and added to the partly filled MOSAIK {sup registered} inserts. The full MOSAIK {sup registered} II-15 U EI casks were emplaced in the present MOSAIK {sup registered} cask store. They are to be transported to the KKS interim store. Removal to an external interim store or a future repository is planned for a later date. BWR plants give rise to control rod guide thimbles for disposal chiefly in the course of decommissioning. The small number of these parts arising during plant operation often create problems in fuel element storage pools because of their large volumes. Two control rod guide thimbles had to be disposed of in the Isar 1 nuclear power station. An underwater saw was developed for this purpose which was used to disassemble two control rod guide thimbles in KKI 1 in 2005. The thimbles were cut up, packed in drums, and these drums were emplaced in high-density concrete casks of the KONRAD-I type. The activated top sections were packed in a MOSAIK {sup registered} II-15 U EI cask under water and, after drainage and drying of the cask, shipped to the Karlsruhe Research Center, disassembled again, packaged, and compacted under high pressure. (orig.)

  3. 控制黑棒和灰棒对AP1000反应堆Keff值影响的M-C模拟%M-C Simulation on Keff Value for Control and Gray Rod Effect in AP1000 Reactor

    Institute of Scientific and Technical Information of China (English)

    魏强林; 刘义保; 杨波; 吴和喜

    2013-01-01

    采用正在三门建造的AP1000核电厂堆芯参数,使用MCNP5程序建立AP1000堆芯数学模型.考虑了燃料棒、黑棒与灰棒7种不同排布方式,分3种情况通过调节黑棒和灰棒在堆芯中的深度来研究有效增值因数Keff值的变化情况.模拟结果表明:随着黑棒和灰棒在反应堆堆芯中的插入,Keff值在1.44-1.22之间变化.为了验证其合理性,并用1 000 x10-6(ppm)的硼酸溶液进行了化学补偿模拟试验,计算得Keff值在1.17-1.07之间,基本能够满足降低过剩反应性的要求.%Reactor core parameter of API000 nuclear power plant is used. Taking into account 7 different ways arrangement of the fuel rod, control and gray rod, AP1000 reactor core mathematical model is established by the MCNP5 code. The effective multiplication factor Keff is studied through 3 different ways by adjusting the depth of control and gray rod in the reactor core. Simulated results show that the Keff value changed between 1.44 and 1. 22 with insertion of control and gray rod, In order to verify the reasonableness, the simulation experiment of chemical compensation is done by using 1 000 × 10-6 (ppm) boric acid, and the experiments show the Keff values changed between 1.17 and 1.07 , which could basically meet the requirements of reducing the excess reactivity.

  4. Piston for rod pumping

    Energy Technology Data Exchange (ETDEWEB)

    Pastushenko, G.I.

    1965-06-22

    A piston, or plunger, for rod pumping, is made up of a cylindrical housing with labryinthal seals, a nose piece, and a scraper. In order to remove paraffin from the inside surface of the production pipe, the housing is made in telescopic form. The scraper consists of an arrangement of springs installed on the outer surface of the housing.

  5. Sucker rod centralizer

    Energy Technology Data Exchange (ETDEWEB)

    Rezewski, J.

    1988-01-26

    This patent describes an oil well sucker rod guide consisting of an elongated body having a number of radial slots. Each slot is disposed at equiangular spaced positions, and contains a roller rotatably supported upon an axle transverse to the slot, such that the roller projects outside the periphery of the body from only one end of the slot

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

    Energy Technology Data Exchange (ETDEWEB)

    Bosselut, D. [Electricite de France (EDF), 92 - Clamart (France). Direction des Etudes et Recherches; Regnier, G. [Ecole Nationale Superieure des Arts et Metiers, 75 - Paris (France); Soulier, B. [DER Mecanique Pole Universitaire Leonard de Vinci, 92 - Paris (France)

    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) 9 refs.

  7. Analysis by the Monte Carlo method of doses around the pool of storage of the control rods irradiated in a BWR reactor; Analisis mediante el metodo de Monte Carlo de las dosis alrededor de la piscina de almacenamiento de las barras de control irradiadas en un reactror BWR

    Energy Technology Data Exchange (ETDEWEB)

    Rodenas, J.; Gallardo, S.

    2011-07-01

    The control rods of a boiling water reactor (BWR) are subject to a neutron flux and thus become activated during their stay in the reactor core. Activation occurs especially in the stainless steel components and impurities. The activity generated results in a dose around the bar, while it le unimportant in the reactor, but to be taken into account when removed f ron it. The bars drawn are stored on hangers placed in the storage pools of spent fuel f ron the plant. Each hanger 12 accommodates control rods and are arranged so that at least three meters of water abode the heads of the control rods. The dose received by potentially exposed workers who are in the vicinity of the storage must be calculated to ensure adequate protection of the came. This dose can be decreased significantly by changing the arrangement of the bars on hangers.

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

    KAUST Repository

    Moulton, D.E.

    2013-02-01

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

  9. Method for making sucker rods

    Energy Technology Data Exchange (ETDEWEB)

    Rasi-Zade, A.T.O.; Kurbanov, N.G.O.; Sutovsky, P.M.; Shikhlinsky, T.M.O.; Kakhramanov, K.T.; Rabinovich, A.M.; Karaev, I.K.O.; Timofeev, V.I.; Ibragimov, O.I.O.

    1991-07-20

    A method for making sucker rods used in oil well pumping is provided, which has the objectives of cutting down the cost of producing sucker rods and of improving their reliability in arduous operating conditions found in wells containing corrosive fluids. The method is characterized in that a rod-body blank is first welded together with rod-end blanks which are made from different materials than the rod-body blank. A welded sucker-rod blank is thus obtained, on which end heads are upset from the blank. The length of the rod-end blank is selected so that weld joints are established, after the upsetting procedure, across the maximum cross-section of the end heads. The method of the invention provides for a weld joint having as much as 3.5 to 4 times the area compared to the rod body, within the zone of the minimum effective stresses acting upon the rod, hence possessing a safety margin of many times the maximum stress applied. This assures high operational reliability and durability of the rods produced according to the invention. The method of the invention does not require precision accuracy in welding the sucker-rod blanks, and minimizes the consumption of expensive alloyed steel, which is used only for making the part of the rod that is subjected to the greatest loads. 7 figs.

  10. Crystal phase-controlled synthesis of rod-shaped AgInTe2 nanocrystals for in vivo imaging in the near-infrared wavelength region

    Science.gov (United States)

    Kameyama, Tatsuya; Ishigami, Yujiro; Yukawa, Hiroshi; Shimada, Taisuke; Baba, Yoshinobu; Ishikawa, Tetsuya; Kuwabata, Susumu; Torimoto, Tsukasa

    2016-03-01

    Rod-shaped AgInTe2 nanocrystals (NCs) exhibiting intense near-band edge photoluminescence in the near-infrared (NIR) wavelength region, were successfully prepared by the thermal reaction of metal acetates and Te precursors in 1-dodecanethiol. Increasing the reaction temperature resulted in the formation of larger AgInTe2 NCs with crystal structures varying from hexagonal to tetragonal at reaction temperatures of 280 °C or higher. The energy gap was increased from 1.13 to 1.20 eV with a decrease in rod width from 8.3 to 5.6 nm, accompanied by a blue shift in the photoluminescence (PL) peak wavelength from 1097 to 1033 nm. The optimal PL quantum yield was approximately 18% for AgInTe2 NCs with rod widths of 5.6 nm. The applicability of AgInTe2 NCs as a NIR-emitting material for in vivo biological imaging was examined by injecting AgInTe2 NC-incorporated liposomes into the back of a C57BL/6 mouse, followed by in vivo photoluminescence imaging in the NIR region.Rod-shaped AgInTe2 nanocrystals (NCs) exhibiting intense near-band edge photoluminescence in the near-infrared (NIR) wavelength region, were successfully prepared by the thermal reaction of metal acetates and Te precursors in 1-dodecanethiol. Increasing the reaction temperature resulted in the formation of larger AgInTe2 NCs with crystal structures varying from hexagonal to tetragonal at reaction temperatures of 280 °C or higher. The energy gap was increased from 1.13 to 1.20 eV with a decrease in rod width from 8.3 to 5.6 nm, accompanied by a blue shift in the photoluminescence (PL) peak wavelength from 1097 to 1033 nm. The optimal PL quantum yield was approximately 18% for AgInTe2 NCs with rod widths of 5.6 nm. The applicability of AgInTe2 NCs as a NIR-emitting material for in vivo biological imaging was examined by injecting AgInTe2 NC-incorporated liposomes into the back of a C57BL/6 mouse, followed by in vivo photoluminescence imaging in the NIR region. Electronic supplementary information (ESI) available

  11. Study on Dynamic Lifting Characteristics of Control Rod Drive Mechanism%控制棒驱动机构动态提升特性研究

    Institute of Scientific and Technical Information of China (English)

    沈小要

    2012-01-01

    基于控制棒驱动机构的磁路和电路方程以及对控制棒驱动机构动态提升过程分析,分别推导出系统静态过程和动态过程的磁路-电路-机械运动耦合方程.采用解析解的方法求解提升起始电流和提升起始时间.采用ASME规范推荐的动态分析的数值仿真方法模拟控制棒驱动机构动态提升过程,分析磁极和衔铁间不同设计间隙下系统的提升特性.结果表明,衔铁起始提升时间随着设计间隙增大而增大,且设计间隙越大,提升所需时间越长;提升速度随着时间的增加而增大,且随着时间的增加,提升加速度增大,设计间隙越小,提升结束时的冲击加速度越大.%Based on the equations of the electric circuit and the magnetic circuit and analysis of the dynamic lifting process for the control rod drive mechanism (CRDM), coupled magnetic-electric-mechanical equations both for the static status and the dynamic status are derived. The analytical method is utilized to obtain the current and the time when the lift starts. The numerical simulation method of dynamic analysis recommended by ASME Code is utilized to simulate the dynamic lifting process of CRDM, and the dynamic features of the system with different design gaps are studied. Conclusions are drawn as: (1) the lifting-start time increases with the design gap, and the time for the lifting process is longer with larger gaps; (2) the lifting velocity increases with time; (3) the lifting acceleration increases with time, and with smaller gaps, the impact acceleration is larger.

  12. 控制棒驱动线水力缓冲特性仿真研究%Simulation Researchon Control Rod Drive Line Hydraulic Buffer Characteristic

    Institute of Scientific and Technical Information of China (English)

    段春辉; 王留兵; 杜华; 赵伟; 王炳炎

    2015-01-01

    According to the principle of hydrodynamics and kinetics, mathematical model for a new type control rod drive line ( CRDL) buffer process was established. With MATLAB software and the mathematical model, a simulation model for CRDL buffer was constructed. Adapting the simulation model, the simulation analysis for the CRDL hydraulic buffering dynamic charateristic was carried out. With the simulation analysis results , reasonable structure parameters were selected. In order to verify the validity of the simulation model, a CRDL buffer for a nuclear power station was used as an example, and a simulation analysis for the CRDL buffer was carried out. According to the comparison result, CRDL buffer simulation results are close to the CRDL buffer test data, and the comparison verify that the simulation model is valid, and it can be used to select reasonable hydraulic buffer parameters.%以一种新型控制棒驱动线水力缓冲器为研究对象, 根据流体力学及动力学的基本原理, 建立驱动线缓冲器缓冲过程的数学模型. 利用MATLAB计算软件及所建立的数学模型建立控制棒驱动线水力缓冲器的仿真模型, 对控制棒驱动线水力缓冲动态特性进行仿真分析, 并根据仿真结果优选出合理的结构参数. 同时为验证建模思路及方法的正确性, 以某一核电堆型控制棒驱动线作为比对对象, 对其进行仿真建模及分析, 并将仿真结果与试验实测值进行了对比验证. 验证结果表明: 驱动线水力缓冲特性仿真结果与试验数据一致性较好, 仿真模型可用于指导水力缓冲参数的选取和优化.

  13. Modeling of control rod ejection transient for WWER-1000-model 446 using RELAP5m3.3/PARCSv2.6 coupled codes

    International Nuclear Information System (INIS)

    Highlights: • Capability to perform 3D neutronics/thermal–hydraulic analysis for WWER-1000 m446. • Good agreement was observed between the coupled codes results and FSAR data. • Capability to perform multi-dimensional analysis of complex transients such as a CREA. • WWER-1000 m446 shows a safe response during these transients performance. - Abstract: By using the Best Estimate (BE) method instead of conservative assumptions for the evaluation of reactor safety, significant economic considerations with optimal fuel burn-up could be obtained in addition to reactor safety. In this method, due to the detailed simulation and feedback considerations, special attention has been paid to the coupling of neutronic and thermo-hydraulic codes to achieve more reliable results. In this study, the Control Rod Ejection (CRE) transient has been simulated for Bushehr Nuclear Power Plant (BNPP) as a WWER-1000 power plant model 446 according to Final Safety Analysis Report (FSAR). CRE is a transient of Reactivity Initiated Accidents (RIA) category. In this study, the reactor thermo-hydraulic system has been simulated by RELAP5/mod3.3, while the neutron kinetic system of the reactor core has been simulated by the PARCSv2.6 code. These codes have been coupled utilizing Parallel Virtual Machine (PVM) interface software to consider the effects of thermal hydraulic and neutronic feedbacks. Thus, the power calculated by the PARCS code is used by the RELAP5 code and the obtained thermal hydraulic parameters are inserted to the PARCS code for macroscopic cross-section calculations. A computer program written by C++ has been used for the cycle execution of the WIMS code to produce the macroscopic cross-section library with the format required by the PARCS code. After the three-dimensional (3D) thermo-neutronic modeling of the reactor core, the Hot Zero Power (HZP) and Hot Full Power (HFP) versions of CRE transients, which have been considered in the plant’s FSAR, have been

  14. Sucker rod centralizer

    Energy Technology Data Exchange (ETDEWEB)

    Rivas, O.; Newski, A.

    1989-10-03

    This patent describes a device for centralizing at least one sucker rod within a production pipe downhole in a well and for reducing frictional forces between the pipe and at least one sucker rod. It comprises an elongate, substantially cylindrical body member having a longitudinal axis, a plurality of slots within the member and a rotatable member mounted within each slot, each of the plurality of slots has its major dimension along a first axis parallel to the longitudinal axis of the body member and is oriented with respect to the other seats so as to form a helicoidal array for maximizing the total surface contact area between the rotatable members and the pipe and for decreasing the forces acting on each rotatable member.

  15. Improvement in Jc performance below liquid nitrogen temperature for SmBa2Cu3Oy superconducting films with BaHfO3 nano-rods controlled by low-temperature growth

    Directory of Open Access Journals (Sweden)

    S. Miura

    2016-01-01

    Full Text Available For use in high-magnetic-field coil-based applications, the critical current density (Jc of REBa2Cu3Oy (REBCO, where RE = rare earth coated conductors must be isotropically improved, with respect to the direction of the magnetic field; these improvements must be realized at the operating conditions of these applications. In this study, improvement of the Jc for various applied directions of magnetic field was achieved by controlling the morphology of the BaHfO3 (BHO nano-rods in a SmBCO film. We fabricated the 3.0 vol. % BHO-doped SmBCO film at a low growth temperature of 720 °C, by using a seed layer technique (Ts = 720 °C film. The low-temperature growth resulted in a morphological change in the BHO nano-rods. In fact, a high number density of (3.1 ± 0.1 × 103 μm−2 of small (diameter: 4 ± 1 nm, discontinuous nano-rods that grew in various directions, was obtained. In Jc measurements, the Jc of the Ts = 720 °C film in all directions of the applied magnetic field was higher than that of the non-doped SmBCO film. The Jcmin (6.4 MA/cm2 of the former was more than 6 times higher than that (1.0 MA/cm2 of the latter at 40 K, under 3 T. The aforementioned results indicated that the discontinuous BHO nano-rods, which occurred with a high number density, exerted a 3D-like flux pinning at the measurement conditions considered. Moreover, at 4.2 K and under 17 T, a flux pinning force density of 1.6 TN/m3 was realized; this value was comparable to the highest value recorded, to date.

  16. Improvement in Jc performance below liquid nitrogen temperature for SmBa2Cu3Oy superconducting films with BaHfO3 nano-rods controlled by low-temperature growth

    Science.gov (United States)

    Miura, S.; Yoshida, Y.; Ichino, Y.; Xu, Q.; Matsumoto, K.; Ichinose, A.; Awaji, S.

    2016-01-01

    For use in high-magnetic-field coil-based applications, the critical current density (Jc) of REBa2Cu3Oy (REBCO, where RE = rare earth) coated conductors must be isotropically improved, with respect to the direction of the magnetic field; these improvements must be realized at the operating conditions of these applications. In this study, improvement of the Jc for various applied directions of magnetic field was achieved by controlling the morphology of the BaHfO3 (BHO) nano-rods in a SmBCO film. We fabricated the 3.0 vol. % BHO-doped SmBCO film at a low growth temperature of 720 °C, by using a seed layer technique (Ts = 720 °C film). The low-temperature growth resulted in a morphological change in the BHO nano-rods. In fact, a high number density of (3.1 ± 0.1) × 103 μm-2 of small (diameter: 4 ± 1 nm), discontinuous nano-rods that grew in various directions, was obtained. In Jc measurements, the Jc of the Ts = 720 °C film in all directions of the applied magnetic field was higher than that of the non-doped SmBCO film. The Jcmin (6.4 MA/cm2) of the former was more than 6 times higher than that (1.0 MA/cm2) of the latter at 40 K, under 3 T. The aforementioned results indicated that the discontinuous BHO nano-rods, which occurred with a high number density, exerted a 3D-like flux pinning at the measurement conditions considered. Moreover, at 4.2 K and under 17 T, a flux pinning force density of 1.6 TN/m3 was realized; this value was comparable to the highest value recorded, to date.

  17. Sucker rod guide

    Energy Technology Data Exchange (ETDEWEB)

    White, R.C.

    1988-10-25

    This patent describes an improved guide for use in a string of sucker rods for reciprocation in a tubing string in a borehole, the sucker rods having threaded male ends, the guide comprising: an elongated upright cylindrical member of external diameter less than the internal diameter of tubing in which it is to be used, the member having sucker rod receiving female threaded openings at the upper and lower ends, the threaded openings being coaxial of the member cylindrical axis whereby the member may be positioned in a string of sucker rods, and including a plurality of spaced-apart parallel sided slots within the member, each slot being of semi-circular configuration and of depth greater than the radius and less than the diameter of the cylindrical member, the sidewalls of each slot being parallel to and equally spaced from a plane of the member cylindrical axis; the member having an axle bore therein for each of the slots, the axle bores being parallel and spaced apart from each other, a plane of the axis of each bore being perpendicular the member cylindrical axis and the axis of each bore being displaced away from the member cylindrical axis; an axle received in each axle bore; and a wheel received on each axle the diameter of each wheel being approximately the diameter of the cylindrical member, the periphery of each wheel extending beyond the member cylindrical wall whereby the wheels are positioned to engage and roll on the internal cylindrical surface of tubing, the planes of adjacent slots in the member being rotationally displaced from each other, a portion of each wheel extending beyond the cylindrical surface of the member, the opposed portion of each wheel being within the confines of the member cylindrical surface whereby each wheel can contact a tubing wall at only one point on its cylindrical surface.

  18. Method for making sucker rods

    Energy Technology Data Exchange (ETDEWEB)

    Karaev, I.K.O.; Shikhlinsky, T.M.O.; Polikhronov, K.P.; Sutovsky, P.M.; Avakian, E.V.; Semkin, N.V.; Rabinovich, A.M.; Dzhabarov, R.D.

    1991-01-15

    A method for making sucker rods composed of a rod body and end heads is provided. The rod body end portions are subjected to an upsetting procedure which is carried out at a temperature that precludes softening of the rod body metal. A thickening is formed on each of the end portions, whose width in a direction square with the rod body axis is equal to or exceeds the head maximum diameter in the place of the weld joint, and whose length exceeds the width of the heat-affected zone involved in the welding process. A transition portion is shaped as a solid of revolution whose cross-section smoothly and continuously decreases from the thickening towards the rod body. The upsetting procedure is followed by pressure welding of each of the end heads together with the thickening on the rod body end portion and by turning the weld joint zone.

  19. Nuclear fuel rods

    International Nuclear Information System (INIS)

    Purpose: To enable a tight seal in fuel rods while keeping the sealing gas pressure at an exact predetermined pressure in fuel rods. Constitution: A vent aperture and a valve are provided to the upper end plug of a cladding tube. At first, the valve is opened to fill gas at a predetermined pressure in the fuel can. Then, a conical valve body is closely fitted to a valve seat by the rotation of a needle valve to eliminate the gap in the engaging thread portion and close the vent aperture. After conducting the reduced pressure test for the fuel rod in a water tank, welding joints are formed between the valve and the end plug through welding to completely seal the cladding tube. Since the welding is conducted after the can has been closed by the valve, the predetermined gas pressure can be maintained at an exact level with no efforts from welding heat and with effective gas leak prevention by the double sealing. (Kawakami, Y.)

  20. Sucker rod coupling

    Energy Technology Data Exchange (ETDEWEB)

    Klyne, A.A.

    1986-11-11

    An anti-friction sucker rod coupling is described for connecting a pair of sucker rods and centralizing them in a tubing string, comprising: an elongate, rigid, substantially cylindrical body member, each end of the body member forming means for threadably connecting the body member with a sucker rod. The body member further forms a transversely extending, substantially diametric, generally vertical slot extending therethrough. The body member further forms a pin bore, such pin bore extending transversely through the body member so as to intersect the slot substantially perpendicularly; a wheel member positioned within the slot to rotate in a generally vertical plane. The wheel member has a portion thereof extending beyond the periphery of the body member to engage the inner surface of the tubing string and centralize the coupling; and a pin mounted in the pin bore and supporting member thereon, whereby the wheel member is rotatable within the slot; the wheel member having sufficient clearance between its side surfaces and the wall surfaces of the slot, when the wheel member is centered in the slot on the pin, whereby the wheel member may shift along the pin to assist in ejecting sand and oil from the slot.

  1. Upper end plug of fuel rod

    International Nuclear Information System (INIS)

    The present invention concerns a seal-welding of an upper end plug of a fuel rod for nuclear fuels conducted in a final stage of molding fabrication of the fuel rod in a pressurized helium gas. A welding protrusion is formed at the periphery of a vent hole on the upper surface of the upper end plug, and the welding protrusion is melted by irradiation of laser beams. The melted protrusion intrudes into the end portion of the bent hole by capillary to close the vent hole. The upper end plug can be closed by an extremely simple operation of irradiating the laser beams to the protrusion. Control for electrode gap on every fuel rods and exchange for the electrodes as in TIG welding can be saved, thereby enabling to speed up and simplify the sealing operation for the upper end plug. (N.H.)

  2. Evaluation of the reduction of boron-10 in the control rods in the BWR of the Laguna Verde Central, through steady state calculations; Evaluacion de la reduccion del Boro-10 en las barras de control en los BWR de la CLV, mediante calculos en estado estacionario

    Energy Technology Data Exchange (ETDEWEB)

    Montes T, J.L.; Perusquia, R.; Hernandez, J.L.; Ramirez S, J.R. [Departamento de Sistemas Nucleares, ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    2003-07-01

    One of the more important aspects related with the safety and economy in the operation of a nuclear power reactor, it is without a doubt the control of the reactivity. During the normal operation of a reactor of boiling water (BWR-Boiling Water Reactor), the control of the reactivity in the nucleus it is strongly determined by the efficiency of the control rods. In the case of the Laguna Verde Nuclear power station (CNLV) the nucleus of the reactors has 109 control rods grouped in 4 sets. The CNLV at the moment uses the CCC method (Control Cell Core) in the design of the cycle. With this method only the A2 group is used for the control of the reactivity at full power. With the purpose of quantifying the effect of the decrease of the burnable poison (B{sub 4}C) of the control rods and in particular to the effect due to the postulated lost of 10% of Boron 10, it was carried out a series of calculations of the nucleus in stationary state by means of the system of HELIOS/CM-PRESTO codes. In this work the main derived results of these 3D simulations(three dimensions) of the reactors of the CNLV are presented. It was analyzed the one behavior of the infinite neutron multiplication factor (K{sub infinite}), at fuel assemble cell level used in an equilibrium cycle for the CNLV. It was also analyzed the effect in the shutdown margin (ShutDown Margin- SDM) in cold condition CZP (Cold Zero Power). Its are also included those results of the ARI cases (All Rods In) and SRO (Strong Rod Out). From the cases in condition HFP (Hot Full Power) the behavior of the effective multiplication factor (K{sub eff}) is presented. (Author)

  3. Fiber optic laser rod

    Science.gov (United States)

    Erickson, G.F.

    1988-04-13

    A laser rod is formed from a plurality of optical fibers, each forming an individual laser. Synchronization of the individual fiber lasers is obtained by evanescent wave coupling between adjacent optical fiber cores. The fiber cores are dye-doped and spaced at a distance appropriate for evanescent wave coupling at the wavelength of the selected dye. An interstitial material having an index of refraction lower than that of the fiber core provides the optical isolation for effective lasing action while maintaining the cores at the appropriate coupling distance. 2 figs.

  4. Cone rod dystrophies.

    Science.gov (United States)

    Hamel, Christian P

    2007-01-01

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

  5. Piston and connecting rod assembly

    Science.gov (United States)

    Brogdon, James William (Inventor); Gill, David Keith (Inventor); Chatten, John K. (Inventor)

    2001-01-01

    A piston and connecting rod assembly includes a piston crown, a piston skirt, a connecting rod, and a bearing insert. The piston skirt is a component separate from the piston crown and is connected to the piston crown to provide a piston body. The bearing insert is a component separate from the piston crown and the piston skirt and is fixedly disposed within the piston body. A bearing surface of a connecting rod contacts the bearing insert to thereby movably associate the connecting rod and the piston body.

  6. Investigating the optical XNOR gate using plasmonic nano-rods

    Science.gov (United States)

    Akhlaghi, Majid; Kaboli, Milad

    2016-04-01

    In this paper, a coherent perfect absorption (CPA)-type XNOR gate based on plasmonic nano particle is proposed. It consists of two plasmonic nano rod arrays on top of two parallel arms with quartz substrate. The operation principle is based on the absorbable formation of a conductive path in the dielectric layer of a plasmonic nano-particles waveguide. Since the CPA efficiency depends strongly on the number of plasmonic nano-rod and the nano rod location, an efficient binary optimization method based the Particle Swarm Optimization (PSO) algorithm is used to design an optimized array of the plasmonic nano-rod in order to achieve the maximum absorption coefficient in the 'off' state and the minimum absorption coefficient in the 'on' state. In Binary PSO (BPSO), a group of birds consists a matrix with binary entries, control the presence ('1‧) or the absence ('0‧) of nano rod in the array.

  7. Fuel rod plugs

    International Nuclear Information System (INIS)

    Purpose: To prevent the formation of voids to the inside of welded portion in fuel rod plugs. Constitution: A fuel rod is tightly sealed by welding end plugs at both ends of a fuel can charged with nuclear fuel material. For the welding of the end plug, laser welding has now been employed with the reason of increasing the welding efficiency and reducing the welding heat distortion. However, if the end plug is laser-welded to the end of the fuel can in the conventional form, there is a problem that voids are liable to be formed near the deepest penetration in the welding portion. That is, gases evolved near the deepest penetration remains in a key-hole like welded metal portion to result in voids there. Accordingly, grooves capable of passing the laser beam key hole therethrough are disposed along the circumferential direction of the pipe at the end plug welded portion in the fuel can. In this way, since gases generating near the deepest penetration are discharged into the grooves, the key hole-like welded metal is completely filled and voids are not formed. (Kamimura, M.)

  8. Biomechanical study comparing a new combined rod-plate system with conventional dual-rod and plate systems.

    Science.gov (United States)

    Sha, Mo; Ding, Zheng-Qi; Ting, Hu S; Kang, Liang-Qi; Zhai, Wen-Liang; Liu, Hui

    2013-02-01

    Most anterior spinal instrumentation systems are designed as either a plate or dual-rod system and have corresponding limitations. Dual-rod designs may offer greater adjustability; however, this system also maintains a high profile and lacks a locking design. Plate systems are designed to be stiffer, but the fixed configuration is not adaptable to the variety of vertebral body shapes. The authors designed a new combined rod-plate system (D-rod) to overcome these limitations and compared its biomechanical performance with the conventional dual-rod and plate system. Eighteen pig spinal specimens were divided into 3 groups (6 per group). An L1 corpectomy was performed and fixed with the D-rod (group A; n=6), Z-plate (Sofamor Danek, Memphis, Tennessee) (group B; n=6), or Ventrofix (Synthes, Paoli, Pennsylvania) (group C; n=6) system. T13-L2 range of motion was measured with a 6 degrees of freedom (ie, flexion-extension, lateral bending, and axial rotation) spine simulator under pure moments of 6.0 Nm. The D-rod and Ventrofix specimens were significantly stiffer than the Z-plate specimens (Pplate specimens were significantly stiffer than the Ventrofix specimens (Pplate and dual-rod systems, where the anterior rod exhibits the design of a low-profile locking plate, enhanced stability, and decreased interference of the surrounding vasculature. The posterior rods function in compression and distraction, and the dual-rod system offers greater adjustability and control over screw placement. The results indicate that it may provide adequate stability for anterior thoracolumbar reconstruction. PMID:23383624

  9. Optimization of fuel rod enrichment distribution for BWR fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Hirano, Yasushi; Hida, Kazuki; Sakurada, Koichi; Yamamoto, Munenari [Toshiba Corp., Kawasaki, Kanagawa (Japan). Nuclear Engineering Lab.

    1996-09-01

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

  10. 118-C-4 Horizontal Rod Cave characterization plan

    International Nuclear Information System (INIS)

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

  11. 台球杆皮头对杆法控制的影响分析%An analysis on the influence of the pool cue tip to rod method control

    Institute of Scientific and Technical Information of China (English)

    赵晓光

    2014-01-01

    Billiard is not only the competition of accuracy, but also the control of main ball blocking. The hardness of tip has an influence on main ball blocking at rod method control, especially in distance, the effect of mild tip is worse than hard one, and the hand feeling is better than hard one. Learning the influence factors of tip to rod method and rational techniques could improve the level.%台球不仅是准度的比拼,更关键是对主球走位的掌控。在杆法控制时皮头的硬度对主球的走位影响还是较大的,尤其是远距离的主球走位,软皮头效果不如较硬的皮头效果好,较硬的皮头在手感上又不如软性皮头好。了解皮头对杆法的影响能达到提高技术水平的目的。

  12. Computation of Resistance Force of Dropping Control Rod Assembly under Environment of Fluid%控制棒组件在流体环境中下落时所受阻力的计算

    Institute of Scientific and Technical Information of China (English)

    于建华; 魏泳涛; 孙磊; 李锡华

    2001-01-01

    基于流体力学和边界层理论,探讨了控制棒组件在下落过程中所受的流体阻力。对控制棒组件的 3种部件——控制棒束、驱动杆、星形爪在下落时所受的流体阻力分别按 SU/PG迎风修正的有限元数值方法及基于边界层理论的近似解析法进行求解,完成了控制棒组件下落过程的时程分析,并得出该组件的下落参数 (位置和速度 )与瞬时流体阻力的关系。%Basing on fluid mechanics and boundary theory,the principle of the fluid resistance force acting upon the Rod Cluster Control Assembly(RCCA) are concerned.The numerical method of SU/PG FEM and the approximate analytical solution based on the boundary theory are introduced respectively for the computation of resistance force of RCCA(including Control Rodlets,Drive Rod and Spider) when dropping.The time history analysis of RCCA during its drop have been done,and the relationship between the dropping parameter(position and velocity) and the transient fluid resistance force are obtained.

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

    Energy Technology Data Exchange (ETDEWEB)

    Hirano, Yasushi; Hida, Kazuki; Sakurada, Koichi; Yamamoto, Munenari [Toshiba Corp., Kawasaki, Kanagawa (Japan). Nuclear Engineering Lab.

    1997-01-01

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

  14. 基于SOPC的反应堆棒位信息监测技术研究%Research on Reactor Control Rod Position Indication Information Monitoring Technology Based on SOPC

    Institute of Scientific and Technical Information of China (English)

    郑晓; 蔡晨; 孙宇; 刘明星

    2013-01-01

    A new monitoring technology for control rod position is developed by utilizing the FPGA (Field-Programmable Gate Array) platform and using the SOPC (System On Programmable Chip) technology.In this SOPC system,Nios Ⅱ CPU,VGA (Video Graphics Array) display controller and CAN (Controller Area Network) bus controller are integrated in one FPGA chip.Thus the SOPC hardware platform with comprehensive functionality is constructed.Based on this SOPC platform,the real time data of control rod position indication system can be vividly displayed on the LCD and stored in the external nonvolatile RAM as history records.Therefore,the operator can obtain the overall operation status of control rod position indication system quickly and conveniently.The developed prototype has proved the feasibility of this technology.%运用可编程片上系统(SOPC)技术,以现场可编程门阵列(FPGA)为数字平台,研究一种反应堆棒位信息监测技术.该监测技术将Nios Ⅱ处理器、视频图形阵列(VGA)显示控制器、控制器局域网络(CAN)总线控制器等集成在一片FPGA芯片中,构建完成一个具有丰富功能的SOPC硬件平台.采用该硬件平台,可将棒位系统的运行状态信息实时、直观地进行数字化显示,同时以日志的形式存储在外部存储设备中,能够全方位地监测棒位系统的运行状态,使操纵人员能更方便、快捷地掌握整个棒位系统的运行情况.研制完成的原理样机验证了该技术的可行性.

  15. Sucker rod pump

    Energy Technology Data Exchange (ETDEWEB)

    Brewer, J.R.

    1992-04-14

    This patent describes a subsurface well pump, it comprises: a working barrel; a plunger which reciprocates along the vertical axis within the working barrel between an upper and lower position; a rod connected to the plunger and extending to a means for providing reciprocating force; a well string extending from the top of the working barrel to the surface; an outlet check valve which permits flow to exit the working barrel into the well string and does not permit flow to exit the well string into the working barrel; and an inlet check valve which permits flow into the working barrel from outside of the subsurface pump, the inlet check valve being above the top position of the plunger, the inlet check valve having a cross sectional flow area about equal to or greater than the horizontal cross sectional area of the working barrel, and the inlet check valve being a hinged flapper valve.

  16. H82B钢盘条冬季控冷工艺%Controlled cooling process of H82B steel wire rod in winter

    Institute of Scientific and Technical Information of China (English)

    王雷; 麻晗; 峰公雄

    2011-01-01

    冬季低温脆断是高碳钢盘条的普遍问题,对盘条及下游客户拉丝生产造成了严重影响。使用传统斯太尔摩控冷工艺生产的H82B钢盘条常常因心部马氏体、网状渗碳体超标及塑性不达标造成判次,这种现象在冬季尤为明显。通过改进斯太尔摩线控冷工艺等措施,明显改善了盘条的金相组织和力学性能,缩短了冬季自然时效时间,减少了冬季脆断现象。最后阐述了优化控冷工艺的原理。%Winter brittleness is a common problem for high carbon wire rods,which always leads to breakage during the cold drawing process.For H82B steel wire rod,the brittleness is even worse due to center martensite,cementite network and deficient plasticity caused by faster cooling in winter.By improving the Stelmor cooling condition,the microstructure and mechanical properties were improved,the winter natural aging time was shortened,and the winter brittleness was overcome.Process route to minimize the brittleness was designed with discussion on the mechanism involved.

  17. Determination of the rod-wire transition length in colloidal indium phosphide quantum rods.

    Science.gov (United States)

    Wang, Fudong; Buhro, William E

    2007-11-21

    Colloidal InP quantum rods (QRs) having controlled diameters and lengths are grown by the solution-liquid-solid method, from Bi nanoparticles in the presence of hexadecylamine and other conventional quantum dot surfactants. These quantum rods show band-edge photoluminescence after HF photochemical etching. Photoluminescence efficiency is further enhanced after the Bi tips are selectively removed from the QRs by oleic acid etching. The QRs are anisotropically 3D confined, the nature of which is compared to the corresponding isotropic 3D confinement in quantum dots and 2D confinement in quantum wires. The 3D-2D rod-wire transition length is experimentally determined to be 25 nm, which is about 2 times the bulk InP exciton Bohr radius (of approximately 11 nm).

  18. Eulerian formulation of elastic rods

    Science.gov (United States)

    Huynen, Alexandre; Detournay, Emmanuel; Denoël, Vincent

    2016-06-01

    In numerous biological, medical and engineering applications, elastic rods are constrained to deform inside or around tube-like surfaces. To solve efficiently this class of problems, the equations governing the deflection of elastic rods are reformulated within the Eulerian framework of this generic tubular constraint defined as a perfectly stiff normal ringed surface. This reformulation hinges on describing the rod-deformed configuration by means of its relative position with respect to a reference curve, defined as the axis or spine curve of the constraint, and on restating the rod local equilibrium in terms of the curvilinear coordinate parametrizing this curve. Associated with a segmentation strategy, which partitions the global problem into a sequence of rod segments either in continuous contact with the constraint or free of contact (except for their extremities), this re-parametrization not only trivializes the detection of new contacts but also transforms these free boundary problems into classic two-points boundary-value problems and suppresses the isoperimetric constraints resulting from the imposition of the rod position at the extremities of each rod segment.

  19. Status of rod consolidation, 1988

    International Nuclear Information System (INIS)

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

  20. Nuclear reactor fuel assembly with fuel rod removal means

    International Nuclear Information System (INIS)

    A fuel assembly is described for a nuclear reactor. The assembly has a bottom nozzle, at least one longitudinally extending control rod guide thimble attached to and projecting upwardly from the bottom nozzle and transverse grids spaced along the thimble. An organized array of elongated fuel rods are transversely spaced and supported by the grids and axially captured between the bottom nozzle and a top nozzle. The assembly comprises: (a) a transversely extending adapter plate formed by an arrangement of integral cross-laced ligaments defining a plurality of coolant flow openings; (b) means for mounting the adapter plate on an upper end portion of the thimble and spaced axially above and disposed transversely over the upper ends of all of the fuel rods present in the fuel assembly such that ones of the ligaments overlie corresponding ones of the fuel rods so as to prevent the fuel rods from moving upwardly through the coolant flow openings; and (c) removable plug means confined within the adapter plate and positioned over and spaced axially above selected ones of the fuel rods in providing access to at least one fuel rod for removal thereof upwardly through the axially spaced adapter plate without removing the top nozzle from the fuel assembly

  1. Numerical simulation of the sucker-rod pumping system

    Directory of Open Access Journals (Sweden)

    Oldrich Joel Romero

    2014-11-01

    Full Text Available The sucker rod pump is an artificial lift method frequently applied in onshore petroleum wells. This system can be described using a numerical simulation based on the behavior of a rod string. In the past, the elastic behavior of the rod string made it difficult to model the system. However, since the 1960s and with the advent of digital computers, it has been modeled numerically. The rod string be-haves like a slender bar, and thus, the propagation of elastic waves along the bar can be represented by a one-dimensional equation. Gibbs (1963 presented a mathematical model based on the wave equation, which is described on the basis of the analysis of forces on the rod string and is incorporated into a boundary value problem involving partial differential equations. The use of the finite differ-ence method allows for a numerical solution by the discretization of the wave equation developed in the mathematical formulation with appropriate boundary and initial conditions. This work presents a methodology for implementing an academic computer code that allows simulation of the upstroke and downstroke motion of the rod string described by the wave equation under ideal operating conditions, assuming a harmonic motion of the rod at one end and downhole pump at the other end. The goal of this study is to generate the downhole dynamometer card, an important and consolidated tool that controls the pump system by diagnosing oper-ational conditions of the downhole pump.

  2. Research on Lift Pole Thread Fatigue of Magnetic Lifting Control Rod Drive Mechanism%磁力提升型控制棒驱动机构提升磁极螺纹疲劳研究

    Institute of Scientific and Technical Information of China (English)

    唐向东; 杨博; 陈西南; 余志伟; 王德军

    2013-01-01

    Control Rod Drive Mechanism (CRDM) is a servomechanism of reactor control and protection system,and it is the only movable equipment in reactor body.Its reliability affects the reactor safety and operation directly.Especially in accident conditions,CRDM must release the control rod to insert into the core immediately.This paper presents the thread fatigue analysis of the CRDM lift pole through electromagnetic analysis,mechanical analysis and fatigue analysis method,and presents a fatigue analysis method for thread bearing impact load in Magnetic Lifting CRDM.This method can be used in the analysis and design of the similar type of CRDM.%对磁力提升型控制棒驱动机构的运动进行电磁-运动仿真.通过电磁分析、力学分析以及疲劳分析的方法,对控制棒驱动机构提升磁极螺纹进行了疲劳分析.结果表明,目前二代核电机型中驱动机构提升磁极在使用寿命内,螺纹结构满足使用要求,但对于第三代核电机型中驱动机构提升磁极在使用寿命末,第一牙螺纹的应力接近许用应力,疲劳寿命及疲劳安全系数偏低,需要进一步优化.

  3. Composites reinforcement by rods: a SAS study

    International Nuclear Information System (INIS)

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

  4. LOFT nuclear fuel rod behavior

    International Nuclear Information System (INIS)

    An overview of the calculational models used to predict fuel rod response for Loss-of-Fluid Test (LOFT) data from the first LOFT nuclear test is presented and discussed and a comparison of predictions with experimental data is made

  5. 中国先进研究堆控制棒驱动机构改进空程实验%Experiment on Idle Stroke of Improved Control Rod Drive Mechanism in China Advance Research Reactor

    Institute of Scientific and Technical Information of China (English)

    徐鹏程; 甄建霄

    2015-01-01

    There are some defects in the design of the control rod drive mechanism which is actively serving in China Advance Research Reactor (CARR). Because of the defects, a method was proposed to improve the mechanism. Theoretically, idle stroke of the improved mechanism is half of that of actively serving mechanism. To verify the improvement, experiments, which were about the influence elements like the speed of the rod, current, frictional force and the load of the mechanism, were designed with control variables method. The results showed that idle stroke were reduced about two-thirds after the improvement, which had achieved the design aim.Besides, the main influencing factors, the frictional force of the system and the coupling stiffness of armature and coil, of idle stroke were determined.%中国先进研究堆(CARR)现役控制棒驱动机构在设计上存在一定缺陷,因而对该驱动机构提出了改进。改进型机构空程的设计目标为达到现役机构空程一半以内,为验证改进设计,采用控制变量法分别针对棒速、电流、摩擦力、负载等因素进行实验。实验结果表明,改进型驱动机构的空程较现役驱动机构缩小了约一半,达到了改进设计目标;并且确定了该驱动机构空程的影响因素主要是衔铁和线圈的连接刚度以及系统摩擦力。

  6. 核电站控制棒驱动机构棘爪销孔堆焊技术研究%Surfacing technology of pawl and pin hole in control rod drive mechanism

    Institute of Scientific and Technical Information of China (English)

    吕永红; 周建明; 白冰; 卢朝晖; 向文元; 黄文有; 郑继雷

    2012-01-01

    As a key moving part in the control rod drive mechanism, the safety and reliability of the pawl should be ensured in the 60 years of the total life cycle, and it should have good abrasion resistance and toughness. Cobalt base alloy was deposited on the surface on the pawl and the pin hole with oxygen - acetylene welding process. Surfaced the pawl completes the life test of 10000000 times in a simulated nuclear reactor pressure vessel under high temperature and high pressure environment, which can be applied in engineering in quantity. The technical research in the least five years were summarized, and a detailed description of the constraints of control rod drive mechanism is a key technology in domestic.%为确保核电站控制棒驱动机构的核心运动部件——棘爪在60年全寿命周期的安全可靠性,以及运动时具有良好的耐磨性和韧性,采用氧乙炔堆焊方法对棘爪表面和销孔进行钴基合金堆焊.通过自主技术攻关,堆焊后的棘爪在模拟核电站反应堆压力容器高温高压的运行环境下,完成了1000万次的寿命考验,并具备了批量化工程应用的能力,打破了国外的技术垄断,填补了国内该领域的空白.文中主要对近5年的技术研究进行总结,详细描述了制约核电站控制棒驱动机构国产化的一种关键技术.

  7. Theoretic Analysis for Static Characteristic of Servo-tube Guided Hydraulic Control Rod Driving Mechanism%伺服管主导型控制棒水力驱动机构静态特性理论分析

    Institute of Scientific and Technical Information of China (English)

    韩文伟; 郭清; 韩伟实

    2014-01-01

    The analysis and calculation of static characteristic were carried out for servo-tube guided hydraulic control rod driving mechanism .The static holding flow rate and its variation law with temperature were acquired .The results indicate that the static holding flow rate needed is very small in steady working range of variable throttle orifice .The liquid density decreases with the increase of temperature ,and then the static holding flow rate increases accordingly .In inclining condition ,the range of static holding flow rate is augmented and the holding characteristic of control rod is more stable . Therefore , the resistance ability to perturbation is much stronger and is conformable to the criterion of nautical nuclear power device .%本文对伺服管主导型控制棒水力驱动机构的静态保持特性进行理论计算和分析,得到了水力驱动缸的静态保持流量及其随温度的变化规律。结果表明:在可变节流口间隙的稳态工作范围内,系统所需的静态保持流量很小;随温度的升高,流体密度降低,静态保持流量相应增大。倾斜工况下的系统静态保持流量范围较正常工况下的大,控制棒的保持特性更加稳定,抗扰动能力更强,满足舰船核动力装置规范的规定。

  8. Process-based tolerance assessment of connecting rod machining process

    Science.gov (United States)

    Sharma, G. V. S. S.; Rao, P. Srinivasa; Surendra Babu, B.

    2016-01-01

    Process tolerancing based on the process capability studies is the optimistic and pragmatic approach of determining the manufacturing process tolerances. On adopting the define-measure-analyze-improve-control approach, the process potential capability index ( C p) and the process performance capability index ( C pk) values of identified process characteristics of connecting rod machining process are achieved to be greater than the industry benchmark of 1.33, i.e., four sigma level. The tolerance chain diagram methodology is applied to the connecting rod in order to verify the manufacturing process tolerances at various operations of the connecting rod manufacturing process. This paper bridges the gap between the existing dimensional tolerances obtained via tolerance charting and process capability studies of the connecting rod component. Finally, the process tolerancing comparison has been done by adopting a tolerance capability expert software.

  9. Simulation of leaking fuel rods

    International Nuclear Information System (INIS)

    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 123I activity peak; 5) Using Paks NPP data a conservative value has been determined for the upper limit of the 123I release from failed fuel rods during transients

  10. Analysis of Double-encapsulated Fuel Rods

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-09-01

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

  11. Results of Post Irradiation Examinations of VVER Leaky Rods

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-06-15

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

  12. The Third ATLAS ROD Workshop

    CERN Multimedia

    Poggioli, L.

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

  13. Performance of the NRX shut-off rods

    International Nuclear Information System (INIS)

    A new type of shut-off rod of electromechanical design was developed by the American Machine and Foundry Company for use in the NRX reactor following the accident of 1952. The new rods were installed in May, 1956, as part of the control system conversion program which was completed in 1958. Some problems were encountered with limit switch adjustment but minor modifications in design led to much improved operation. he performance of the rods also improved as more experience was gained in the maintenance and adjustment of the various headgear components. Each headgear is now overhauled once a year on a routine basis. The present design of shut-off rod is considered to be very satisfactory. There has only been one occasion when a shut-off rod has failed to come fully down on a trip. Rods have failed to operate correctly on five other occasions but these occurred during shutdown periods or when the reactor was being shutdown manually. (author)

  14. Synthesis and Liquid-Crystal Behavior of Bent Colloidal Silica Rods.

    Science.gov (United States)

    Yang, Yang; Chen, Guangdong; Martinez-Miranda, Luz J; Yu, Hua; Liu, Kun; Nie, Zhihong

    2016-01-13

    The design and assembly of novel colloidal particles are of both academic and technological interest. We developed a wet-chemical route to synthesize monodisperse bent rigid silica rods by controlled perturbation of emulsion-templated growth. The bending angle of the rods can be tuned in a range of 0-50° by varying the strength of perturbation in the reaction temperature or pH in the course of rod growth. The length of each arm of the bent rods can be individually controlled by adjusting the reaction time. For the first time we demonstrated that the bent silica rods resemble banana-shaped liquid-crystal molecules and assemble into ordered structures with a typical smectic B2 phase. The bent silica rods could serve as a visualizable mesoscopic model for exploiting the phase behaviors of bent molecules which represent a typical class of liquid-crystal molecules. PMID:26700616

  15. Refabricated and instrumented fuel rods

    International Nuclear Information System (INIS)

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

  16. Disposal on Unqualified Items About Crape on Sealing Tube in Control Rod Driving System Suspended by Magnetic Force%磁悬浮控制棒驱动线密封管划伤的不符合项处理

    Institute of Scientific and Technical Information of China (English)

    张之华; 刘汉刚; 钱达志; 张新荣; 李江波; 徐显启; 方建国

    2012-01-01

    Serious unacceptable scrapes on the ektexine of some sealing tubes were discovered during integrative capability test of control rod driving framework suspended by magnetic force. Heating and disasembling tests on control rod driving framework were carried out. The results show that rudimental bits during maching, impertinency in local configuration and improper chosen material of gliding parts are main reasons. Filtering nitrogen, disposing of maching rudimental matter, acute angle passivation, replacing material, increasing guide flowing holes were performed, after which run and downfall tests were carried out. It is shown that no new scrapes appear and rob downfall time decreases. The method and results can be conference for maching and maintenance of congener control rob driving framework.%在进行磁悬浮控制棒驱动线的综合性能试验时,发现部分密封管的外壁产生严重的不可接受的划伤.通过控制棒驱动线运行的温升试验、解体试验,发现加工制造过程中的残留切屑、局部结构的不合理、滑动副部件的材料选取不当是造成密封管产生划痕的主要原因.通过对密封管外部进行渗氮处理、清理间隙槽内的加工残留物,钝化各锐角,更换限位套的材料,增加连杆上的导流孔等整改措施后,进行了驱动线的运行考验和快速落棒试验,未发现产生新的划痕并有效缩短了控制棒的快速落棒时间.本工作采取的措施及取得的试验结果,可为同类控制棒驱动线的加工制造和运行维护提供借鉴作用.

  17. The Effect of Nano seed Concentration on the Aspect Ratio of Gold Nano rod

    International Nuclear Information System (INIS)

    This paper reports the synthesis of gold nano rod with controlled aspect ratio prepared by varying the concentration of nano seed addition into the growth solution via the seed mediated growth method. In typical process, the gold nano rod with aspect ratio from ca. 2.2 to 4.2 can be successfully obtained. Owing to its simplicity, the present approach could be used to produce gold nano rod with special properties for SERS and catalyst application. (author)

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

    OpenAIRE

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

    2011-01-01

    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.

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

    CERN Document Server

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

    2013-01-01

    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.

  20. Analysis of reciprocating compressor piston rod failures

    Energy Technology Data Exchange (ETDEWEB)

    Tripp, H.A.; Drosjack, M.J.

    1984-02-01

    This report presents the analysis of five piston rod failures which occurred on reciprocating compressors. Calculations are shown for rod stress which includes nominal rod loading sources as well as additional loads due to unusual pressure losses in the compressor valves, flexure of the rods due to misalignment, and manufacturing errors. The additional loads were incorporated on the basis of field measurements. The stress values are used with Baquin's equation to produce fatigue life curves for the rods. Based on the calculations, recommendations for modified rods were made. The calculation procedures are described in a manner which will permit their application to other reciprocating compressors.

  1. The design and construction of the input and output isolation system for power channel and control rod position in the research reactor SR4 with analog device AD210

    International Nuclear Information System (INIS)

    The design and construction of analog input-output system with device AD210, SR4 (ICS Kartini Reactor) is used as interface and output signal isolation system for power channel meter NLW-2, NP1000 and control measurement with the slave computer system so that the measuring tools and computer operate independently. Device AD210 provides a very compact insulators and economically with highly accurate system performance. It is a DIP chip with 3-port model, input, output and integrated power supply. It can be applied as a multichannel data acquisition, or a single channel and others. Analog Device AD210 is placed between the output channel power measurement and control rod position measurement system / other support tools such as Computer Slave, ADC, multiplexer, and others. The design and construction of Analog Devices AD210 insulator use the input-output mode with the gain of 1, so that pulses or output voltage are equal to the input voltage pulses. The test results in graphs of output versus input are excellent with the linearity close to 100%. The use of this Device AD210 for NLW2 and NP1000 which are a vital instrument for the continuity of the operation of reactor, is expected to increase the Main Time Between Failure (MTBF) of the reactor as a whole. (author)

  2. Application of fiberglass sucker rods

    Energy Technology Data Exchange (ETDEWEB)

    Gibbs, S.G. (Nabla Corporation (US))

    1991-05-01

    Fiberglass sucker rods are assuming a place in artificial-lift technology. This paper briefly describes the manufacturing process and gives some design and operational hints for practical applications. It also describes some mathematical modeling modifications needed for fiberglass wave-equation design programs.

  3. Application of fiberglass sucker rods

    International Nuclear Information System (INIS)

    Fiberglass sucker rods are assuming a place in artificial-lift technology. This paper briefly describes the manufacturing process and gives some design and operational hints for practical applications. It also describes some mathematical modeling modifications needed for fiberglass wave-equation design programs

  4. Process development and fabrication for sphere-pac fuel rods. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Welty, R.K.; Campbell, M.H.

    1981-06-01

    Uranium fuel rods containing sphere-pac fuel have been fabricated for in-reactor tests and demonstrations. A process for the development, qualification, and fabrication of acceptable sphere-pac fuel rods is described. Special equipment to control fuel contamination with moisture or air and the equipment layout needed for rod fabrication is described and tests for assuring the uniformity of the fuel column are discussed. Fuel retainers required for sphere-pac fuel column stability and instrumentation to measure fuel column smear density are described. Results of sphere-pac fuel rod fabrication campaigns are reviewed and recommended improvements for high throughput production are noted.

  5. Site controlled red-yellow-green light emitting InGaN quantum discs on nano-tipped GaN rods

    Science.gov (United States)

    Conroy, M.; Li, H.; Kusch, G.; Zhao, C.; Ooi, B.; Edwards, P. R.; Martin, R. W.; Holmes, J. D.; Parbrook, P. J.

    2016-05-01

    We report a method of growing site controlled InGaN multiple quantum discs (QDs) at uniform wafer scale on coalescence free ultra-high density (>80%) nanorod templates by metal organic chemical vapour deposition (MOCVD). The dislocation and coalescence free nature of the GaN space filling nanorod arrays eliminates the well-known emission problems seen in InGaN based visible light sources that these types of crystallographic defects cause. Correlative scanning transmission electron microscopy (STEM), energy-dispersive X-ray (EDX) mapping and cathodoluminescence (CL) hyperspectral imaging illustrates the controlled site selection of the red, yellow and green (RYG) emission at these nano tips. This article reveals that the nanorod tips' broad emission in the RYG visible range is in fact achieved by manipulating the InGaN QD's confinement dimensions, rather than significantly increasing the In%. This article details the easily controlled method of manipulating the QDs dimensions producing high crystal quality InGaN without complicated growth conditions needed for strain relaxation and alloy compositional changes seen for bulk planar GaN templates.We report a method of growing site controlled InGaN multiple quantum discs (QDs) at uniform wafer scale on coalescence free ultra-high density (>80%) nanorod templates by metal organic chemical vapour deposition (MOCVD). The dislocation and coalescence free nature of the GaN space filling nanorod arrays eliminates the well-known emission problems seen in InGaN based visible light sources that these types of crystallographic defects cause. Correlative scanning transmission electron microscopy (STEM), energy-dispersive X-ray (EDX) mapping and cathodoluminescence (CL) hyperspectral imaging illustrates the controlled site selection of the red, yellow and green (RYG) emission at these nano tips. This article reveals that the nanorod tips' broad emission in the RYG visible range is in fact achieved by manipulating the InGaN QD

  6. Site controlled red-yellow-green light emitting InGaN quantum discs on nano-tipped GaN rods.

    Science.gov (United States)

    Conroy, M; Li, H; Kusch, G; Zhao, C; Ooi, B; Edwards, P R; Martin, R W; Holmes, J D; Parbrook, P J

    2016-06-01

    We report a method of growing site controlled InGaN multiple quantum discs (QDs) at uniform wafer scale on coalescence free ultra-high density (>80%) nanorod templates by metal organic chemical vapour deposition (MOCVD). The dislocation and coalescence free nature of the GaN space filling nanorod arrays eliminates the well-known emission problems seen in InGaN based visible light sources that these types of crystallographic defects cause. Correlative scanning transmission electron microscopy (STEM), energy-dispersive X-ray (EDX) mapping and cathodoluminescence (CL) hyperspectral imaging illustrates the controlled site selection of the red, yellow and green (RYG) emission at these nano tips. This article reveals that the nanorod tips' broad emission in the RYG visible range is in fact achieved by manipulating the InGaN QD's confinement dimensions, rather than significantly increasing the In%. This article details the easily controlled method of manipulating the QDs dimensions producing high crystal quality InGaN without complicated growth conditions needed for strain relaxation and alloy compositional changes seen for bulk planar GaN templates. PMID:27174084

  7. Site controlled Red-Yellow-Green light emitting InGaN Quantum Discs on nano-tipped GaN rods

    KAUST Repository

    Conroy, Michele Ann

    2016-03-10

    We report a method of growing site controlled InGaN multiple quantum discs (QDs) at uniform wafer scale on coalescence free ultra-high density (>80%) nanorod templates by metal organic chemical vapour deposition (MOCVD). The dislocation and coalescence free nature of the GaN space filling nanorod arrays eliminates the well-known emission problems seen in InGaN based visible light sources that these types of crystallographic defects cause. Correlative scanning transmission electron microscopy (STEM), energy-dispersive x-ray (EDX) mapping and cathodoluminescence (CL) hyperspectral imaging illustrates the controlled site selection of the red, yellow and green (RYG) emission at these nano tips. This article reveals that the nanorod tips’ broad emission in the RYG visible range is in fact achieved by manipulating the InGaN QD’s confinement dimensions, rather than significantly increasing the In%. This article details the easily controlled method of manipulating the QDs dimensions producing high crystal quality InGaN without complicated growth conditions needed for strain relaxation and alloy compositional changes seen for bulk planar GaN templates.

  8. Solitary waves on nonlinear elastic rods. II

    DEFF Research Database (Denmark)

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

    1987-01-01

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

  9. Rod and lamellar growth of eutectic

    Directory of Open Access Journals (Sweden)

    M. Trepczyńska-Łent

    2010-04-01

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

  10. Leaf spring puller for nuclear fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Fogg, J.L.

    1981-11-03

    A fuel rod puller in the form of a collet for pulling fuel rods from a storage area into grids of a nuclear reactor fuel assembly. The rod puller moves longitudinally through the grids to a storage area where projections on the end of leaf springs grasp onto an end plug in a fuel rod. Drive apparatus then pulls the rod puller and connected fuel rod from the storage area into the fuel assembly grids. The rod puller includes an outer tube having leaf springs on one end thereof in one modification, mounted within the outer tube is a movable plunger which acts to urge the leaf springs outwardly to a position to permit passing or with the end of a end plug. Upon withdrawal of the plunger, the leaf springs move into a groove formed in the end of a fuel rod end plug, and the fuel rod subsequently is pulled into the fuel assembly grids. In another modification, the leaf springs on the outer rod are biased in an outward direction and a longitudinally movable tube on the outer rod is moved in a direction to contract the leaf springs into a position where the projections thereof engage the groove formed in a fuel rod end plug.

  11. Rod and lamellar growth of eutectic

    OpenAIRE

    M. Trepczyńska - Łent

    2010-01-01

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

  12. 基于无量纲参数分析的驱动机构运行性能研究%Research on the Running Performance of Control Rod Drive Mechanism Based on Dimensionless Parameters Analysis

    Institute of Scientific and Technical Information of China (English)

    彭翠云; 彭宵微

    2015-01-01

    针对控制棒驱动机构的运行噪声信号,采用无量纲参数分析驱动机构运行性能。分析结果表明:峰态因数与裕度因子对驱动机构运行性能的变化非常敏感,反映了驱动机构的不同运行状态特性,可判别驱动机构早期运行故障,用于分析驱动机构运行故障程度,有效检测控制棒驱动机构的运行性能。%Dimensionless parameters analysis has been applied in control rod driving mechanism ( CRDM ) ai-ming at running noise signals of CRDM in this article.The analysis results indicated that Kurtosis factor and Margin factor were very sensitive to running state of CRDM and reflected the characteristics of different running state of CRDM, can distinguish the early fault of the driving mechanism and used to analyze the operation fail-ure of the driving mechanism.So Kurtosis factor and Margin factor could be applied in the running performance estimation of CRDM.

  13. 第三代核电站控制棒驱动机构冷却系统风机设计%Fan Design of Cooling System for Control Rod Drive Mechanism in Generation Ⅲ Nuclear Power Plant

    Institute of Scientific and Technical Information of China (English)

    翁娜; 沈秋平; 郝国锋

    2014-01-01

    为了满足堆顶组件整体吊装和冷却介质流道的要求,第三代核电站控制棒驱动机构(control rod drive mechanism,CRDM)冷却系统风机采用了有别于在传统压水堆核电站中此类风机的结构和布置形式,研究其性能和鉴定方案是设计该类设备的关键点.通过分析CRDM冷却系统和风机的设计要求,采用传统方法计算主要性能参数,确定了风机的类型和主要结构形式,计算结果表明,第三代核电站CRDM冷却系统风机为非标设计,国内暂无同类成熟产品,其设计需经试验和工程实践检验.针对风机设计中的关键考核点,结合工程经验和标准规范要求,拟定了用于质量鉴定的试验要求和型式试验方案,可为研制此类风机产品提供技术指导.

  14. Heating apparatus for single-fuel-rod experiments

    International Nuclear Information System (INIS)

    A single-fuel-rod heating apparatus was constructed for installation to Semi-Homogeneous Experimental Assembly (SHE); which is used to measure the reactivity temperature coefficient of a single fuel rod in verification of the precision of nuclear design of the VHTR (very high temperature reactor). The apparatus raises the temperature of a single fuel rod up to 7000C. A fuel rod is enclosed in a silica tube coiled with nichrome heater. The silica tube is confined in a zircalloy tube of which outer surfaces are cooled with air to remove the radiation heat. The zircalloy tube is then confined in a double-walled evacuated aluminum tube to prevent heat transfer to SHE core. The heating apparatus consists of evacuation, cooling, instrumentation, control and safety system, beside the heating tube. The heating tube is inserted in a space made by withdrawing a graphite matrix tube along the central axis of SHE core. The base of the heating tube is connected to an evacuation system, which is set on the table of the movable half of SHE. Performance of the heating apparatus shown by test operation with a single graphite rod are: 1. Single fuel rod: 24 mm in diameter, 2400 mm long. 2. Heating ability: The heating up of a single fuel rod to 7000C in 40 min with electric power 3 kW. 3. Cooling capacity: Blower in flow 1.4 m3/min at pressure 0.4 kg/cm2. 4. Heat leakage: lower than 20 W. (author)

  15. Delay stroke piston and rod for engine

    Energy Technology Data Exchange (ETDEWEB)

    Booher, B.V.

    1995-03-09

    A reciprocating piston internal combustion engine comprises a cylinder having opposed ends, a piston reciprocably mounted in the cylinder, a connecting rod having a crank journal end and a piston journal end, the connecting rod connected to the piston at the piston journal end by means for first and second wrist pins spaced longitudinally along the rod, the first wrist pin journaled in a bore in the piston and in a slot in the piston rod, and the second wrist pin journaled in a bore in the piston rod and a longitudinal slot in the piston. (author)

  16. Analysis of Fly Fishing Rod Casting Dynamics

    OpenAIRE

    Gang Wang; Norman Wereley

    2011-01-01

    An analysis of fly fishing rod casting dynamics was developed comprising of a nonlinear finite element representation of the composite fly rod and a lumped parameter model for the fly line. A nonlinear finite element model was used to analyze the transient response of the fly rod, in which fly rod responses were simulated for a forward casting stroke. The lumped parameter method was used to discretize the fly line system. Fly line motions were simulated during a cast based on fly rod tip resp...

  17. Guide for rotating sucker rods

    Energy Technology Data Exchange (ETDEWEB)

    Harrel, R.D.

    1986-11-04

    This patent describes an improved guide for use in a string of sucker rods rotated in a tubing string in a borehole, the sucker rods having threaded male ends, the guide comprising: an elongated upright solid cylindrical coupling body of external diameter less than the internal diameter of tubing in which it is to be used; a pair of spaced apart axle holders positioned in three recess; an axle received in each recess in the coupling body, the axis of each axle being parallel and spaced from the body longitudinal axis; a roller rotatably received on each axle, the periphery of each roller extending exteriorly of the external cylindrical surface of the coupling body; and means to retain each of the holders in the coupling body recess.

  18. Software design of the ATLAS Muon Cathode Strip Chamber ROD

    Science.gov (United States)

    Murillo, R.; Huffer, M.; Claus, R.; Herbst, R.; Lankford, A.; Schernau, M.; Panetta, J.; Sapozhnikov, L.; Eschrich, I.; Deng, J.

    2012-12-01

    The ATLAS Cathode Strip Chamber system consists of two end-caps with 16 chambers each. The CSC Readout Drivers (RODs) are purpose-built boards encapsulating 13 DSPs and around 40 FPGAs. The principal responsibility of each ROD is for the extraction of data from two chambers at a maximum trigger rate of 75 KHz. In addition, each ROD is in charge of the setup, control and monitoring of the on-detector electronics. This paper introduces the design of the CSC ROD software. The main features of this design include an event flow schema that decentralizes the different dataflow streams, which can thus operate asynchronously at its own natural rate; an event building mechanism that associates data transferred by the asynchronous streams belonging to the same event; and a sparcification algorithm that discards uninteresting events and thus reduces the data occupancy volume. The time constraints imposed by the trigger rate have made paramount the use of optimization techniques such as the curiously recurrent template pattern and the programming of critical code in assembly language. The behaviour of the CSC RODs has been characterized in order to validate its performance.

  19. Research on Designing Profiled Rod Warhead

    Institute of Scientific and Technical Information of China (English)

    Huijun Ning; Hao Wang; Cheng Zhang; Dongyang Chen; Wenjun Ruan

    2015-01-01

    A new Kinetic Energy Rod ( KER) warhead named profiled rod warhead is proposed in this paper. Based on the design of profiled rod warhead, a model of profiled rod driven by detonation is established. The detonation process is simulated by ANSYS/LS⁃DYNA, and the deployment velocity and initial flight attitude of rod are achieved. In addition, static rod deployment testing are performed to investigate the damage effect, the spatial flight attitude and deployment velocity. A satisfactory agreement is obtained by the comparison between numerical results and testing results. Meanwhile, the profiled rod studies are conducted to determine a higher penetrability compared with traditional cylindrical rods. Rigid body dynamics equations of profiled rod, which accounts for the influence of air resistance, are set up to predict the flight trajectory of long⁃distance. The results show that the profiled rod may provide a better penetration angle which still maintains a significant penetrability against projectiles when the rods move off long⁃distance range.

  20. Tests pinpoint sucker-rod failures

    Energy Technology Data Exchange (ETDEWEB)

    Elshawesh, F.; Elhoud, A.; Elagdel, E. [Petroleum Research Center, Tripoli (Libyan Arab Jamahiriya)

    1997-05-26

    A detailed metallurgical examination of a 7/8-inch and a 1-inch sucker rod revealed corrosion fatigue had caused their failure. The 7 to 8-inch rod had failed after a few months of service while the 1-inch rod failed after 1 year. Both rods had been used in a sweet-oil environment. Both rods failed by corrosion fatigue because of repeated loads during operations. Pitting because of the presence of chloride ions and carbon dioxide was initiated on the rod surface, which in turn acted as a crack origin from which the fatigue crack initiated and propagated during operations. The pitting was on the external surface. These pits were large and penetrated through the rod cross-section. Fatigue cracking is initiated at the bottom of the pit where high stress concentration is expected and propagated because the rods were subjected to the alternating stresses during operation. The extent of the fatigue crack varied in the two examined rods because of the difference in the rod heat treatment and microstructure. The paper discusses fatigue failure, the visual examination, macroscopic and microscopic examinations, rod properties, and future operations.

  1. Nuclear Fuel Test Rod Fabrication for Data Acquisition Test

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

  2. Actuator system history of safety rod lower latch problems review of latch inspection video tapes

    Energy Technology Data Exchange (ETDEWEB)

    Banks, J.J.

    1992-06-24

    During pre-restart testing the safety rod at position X26-YlO bound after being driven approximately two (2) feet out of the reactor. Subsequently, the rod was manually returned to it`s seated position. Inspection of the lower latch showed that the latch locking plunger button (screwed on to the bottom of the plunger shaft and retained by a pin through a hole drilled through the button and the plunger shaft) was missing. The shaft failed through the hole drilled for the retaining pin. The button, with the retaining pin intact, was found lodged between the safety rod upper adapter collar and the top of the safety rod thimble top fitting. Analysis of the safety rod latch and accompanying forest guide tube design provided assurance that this type of failure would not cause binding during the ``scramming`` of the safety rods. Inspection of all of the ``K`` safety rod lower latches revealed six other latches with missing plunger buttons, and nine with other non-conformances which required latch replacement. A history search conducted by Reactor Engineering Design, Components Handling Group, is included in this report. The history search shows that latch design modifications, as a part of initial development of the latch system and later to improve the delatching operation, were made from 1950 to 1960. These modifications created a condition where latch damage could occur. Video tapes were made during inspection of the safety rod latches in K area and control rod latches in L area. These tapes were reviewed by Reactor Engineering Design Components Handling engineers. The reviews were used for correlation of latch problems reported by the engineers/mechanics making the inspections. The K area tapes showed inspection of 65 of the 66 safety rod latches. The review of the tapes showed the plunger buttons to be missing from five latches. RED-CH reviewed the L Area video taped inspection of 35 control rod clusters (245 latches). No non-conformances were noted.

  3. Vibrational characteristics and wear of fuel rods

    International Nuclear Information System (INIS)

    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

  4. 控制棒驱动机构传热机理与隔热套性能研究%Heat Transfer Mechanism and Thermal Sleeve Performance of Control Rod Drive Mechanism

    Institute of Scientific and Technical Information of China (English)

    周肖佳; 王丰; 刘刚; 林绍萱; 毛飞; 詹阳烈; 张升; 顾汉洋

    2015-01-01

    The control rod drive mechanism (CRDM ) relies on forced cooling measure to maintain its operating temperature .For the complex issue of CRDM heat transfer mech-anism ,the thermo-siphoning natural heat convection analysis model was established based on the hypothesis w hich divides coolant inside CRDM into hot side and cold side , meanw hile ,the experiment to measure CRDM inside/outside temperature distribution and total heat dissipating capacity under different conditions was conducted .Through comparison of analysis results and experimental data , the analysis model based on hypothesis can correctly simulate CRDM heat transfer in real situation , the thermo-siphoning heat and mass transfer is the major way of CRDM axial heat transfer ,and the use of thermal sleeve can weaken thermo-siphoning and reduce heat dissipating capacity effectively .%控制棒驱动机构(CRDM )依靠强制冷却措施维持工作温度.本文针对CRDM 复杂的轴向传热机理 ,基于冷热侧流动的假设建立热虹吸自然对流分析模型 ,计算得到轴向温度分布及隔热套内径与热虹吸传热量之间的关系曲线 ;同时进行验证试验 ,测量不同情况下CRDM内外轴向温度分布和总散热量.通过分析和试验对比证明 :基于假设的分析模型能模拟实际情况 ,热虹吸传质传热是C RD M 轴向传热的主要途径 ,设置隔热套能有效抑制热虹吸、减少散热量.

  5. Study on segmental nonlinear dynamic properties for control rod drive mechanism%控制棒驱动机构的分段非线性动态特性

    Institute of Scientific and Technical Information of China (English)

    沈小要

    2011-01-01

    控制棒驱动机构是核电厂中的重要安全设备,其动态特性是研究和设计该机构的关键.首先建立控制棒驱动机构的磁路和电路方程.然后基于控制棒驱动机构动态提升过程分析,将运动过程分为3个阶段,并分别推导出各阶段的磁路-电路-机械运动耦合方程.采用解析解和数值仿真两种方法相结合,求解了控制棒驱动机构分段的非线性方程.最终得出动态过程中电流、电磁力、衔铁速度、衔铁位移和衔铁加速度等一系列反应控制棒驱动机构动态特性的参数随时间的变化曲线,并得到了实验验证.%The dynamic properties of control rod drive mechanism (CRDM) ,a key safety equipment for nuclear power plants,secure a crucial position in mechanism design. Firstly,relevant equations of electric and magnetic circuits are established. Then, the motional process is classified into three stages based on dynamic lifting process. Meanwhile, the magnetic-electric-mechanical coupling equations are deduced for each stage. By integrating the analytical solution with numerical simulation, the segmental nonlinear equations of CRDM are solved. Finally, such time-based parameters as the coil current,magnetic force,magnet displacement, velocity and acceleration are used for dynamic property attainment and experimental verification.

  6. 一种适用于十字形控制棒的超临界燃料组件设计%Supercritical Fuel Assembly Design Applicable for Cruciform Control Rod

    Institute of Scientific and Technical Information of China (English)

    朱发文; 雷涛; 程华旸; 庞华; 彭园; 茹俊

    2013-01-01

    The supercritical water-cooled reactor (SCWR) has been selected as one of the most promising reactors for Generation IV nuclear reactors due to its higher thermal efficiency and more simplified structure compared to state-of-the-art LWRs.However, its higher outlet temperature and higher temperature difference between inlet and outlet bring much challenge to the design of SCWR fuel assembly.In this paper, the present status of supercritical fuel assembly design at home and abroad is studied and a kind of fuel assembly with two-flow structure applying for cruciform control rod is proposed.The results show that, the design basically meets the requirements of fuel assemhly design, which has good performance.%超临界水冷堆(SCWR)是目前最有应用前景的第四代反应堆堆型之一,与现有轻水堆相比,具有热效率高、结构简单等诸多优势.但SCWR较高的出口温度以及进出口温差给SCWR燃料组件设计带来了很大的挑战.本文研究国内外超临界燃料组件设计的研究现状,提出一种适用于十字形控制棒的双流程燃料组件设计方案.结果表明,该方案基本满足超临界燃料组件的设计要求,具有较好的综合性能.

  7. Stability and failure analysis of steering tie-rod

    Science.gov (United States)

    Jiang, GongFeng; Zhang, YiLiang; Xu, XueDong; Ding, DaWei

    2008-11-01

    A new car in operation of only 8,000 km, because of malfunction, resulting in lost control and rammed into the edge of the road, and then the basic vehicle scrapped. According to the investigation of the site, it was found that the tie-rod of the car had been broken. For the subjective analysis of the accident and identifying the true causes of rupture of the tierod, a series of studies, from the angle of theory to experiment on the bended broken tie-rod, were conducted. The mechanical model was established; the stability of the defective tie-rod was simulated based on ANSYS software. Meanwhile, the process of the accident was simulated considering the effect of destabilization of different vehicle speed and direction of the impact. Simultaneously, macro graphic test, chemical composition analysis, microstructure analysis and SEM analysis of the fracture were implemented. The results showed that: 1) the toughness of the tie-rod is at a normal level, but there is some previous flaws. One quarter of the fracture surface has been cracked before the accident. However, there is no relationship between the flaw and this incident. The direct cause is the dynamic instability leading to the large deformation of impact loading. 2) The declining safety factor of the tie-rod greatly due to the previous flaws; the result of numerical simulation shows that previous flaw is the vital factor of structure instability, on the basis of the comparison of critical loads of the accident tie-rod and normal. The critical load can decrease by 51.3% when the initial defect increases 19.54% on the cross-sectional area, which meets the Theory of Koiter.

  8. Hybrid composite rods for concrete reinforcement

    OpenAIRE

    Fangueiro, Raúl; Pereira, Cristiana Gonilho; Jalali, Said; Araújo, Mário Duarte de; Marques, P.

    2010-01-01

    The current work is concerned with the development of braided composite rods for civil engineering applications, namely for concrete internal reinforcement, as a steel substitute. The research study aims at understanding the tensile behaviour of composite rods reinforced by a textile structure – braided structure with core reinforcement.Seven types of braided composite rods were produced, varying the type of fibres used as a core reinforcement of a polyester braided structure. ...

  9. 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.

  10. A new calculation method for the number of radial slots of a Terfenol rod

    Institute of Scientific and Technical Information of China (English)

    2009-01-01

    Terfenol is an ideal choice for medium to high power low frequency sonar. It can offer the transducer designer higher strain, higher power density, but the designer must be aware of the eddy current. To enhance efficiency of the barrel-stave transducer powered by a Terfenol rod, radial slots rather than laminations were used to control eddy currents in the Terfenol drive rod, and the effectiveness and the number of these slots were studied experimentally and calculated by finite element modeling. Based on the characteristic of vortex path, a new simple geometrical method to calculate the number of the radial slots of a Terfenol rod at the operating frequency is put forward in this paper. Moreover, the calculated results are in good agreement with those of using the finite element method (FEM) for the slotted Terfenol rod given by the literature. The method will save much cost to design Terfenol rod transducers.

  11. A new calculation method for the number of radial slots of a Terfenol rod

    Institute of Scientific and Technical Information of China (English)

    HE XiPing; ZHANG Pin

    2009-01-01

    Terfenol is an ideal choice for medium to high power low frequency sonar. It can offer the transducer designer higher strain, higher power density, but the designer must be aware of the eddy current. To enhance efficiency of the barrel-stave transducer powered by a Terfenol rod, radial slots rather than laminations were used to control eddy currents in the Terfenol drive rod, and the effectiveness and the number of these slots were studied experimentally and calculated by finite element modeling. Based on the characteristic of vortex path, a new simple geometrical method to calculate the number of the radial slots of a Terfenol rod at the operating frequency is put forward in this paper. Moreover, the calculated results are in good agreement with those of using the finite element method (FEM) for the slotted Ter-fenol rod given by the literature. The method will save much cost to design Terfenol rod transducers.

  12. 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

    -PAGE. Recombinant RodA 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 Rod...

  13. 控制棒驱动机构下部 Canopy 焊缝封堵组件设计与分析%Design and Analysis of Control Rod Drive Mechanism Lower Canopy Welding Plugging Assembly

    Institute of Scientific and Technical Information of China (English)

    孙振国; 李跃忠; 赵毛毛; 冉小兵; 戴长年; 刘森

    2015-01-01

    在某些因素的影响下,少量压水堆核电站控制棒驱动机构(CRDM )下部Canopy焊缝区域会发生破损泄漏。通过在焊缝外部安装封堵组件,能有效预防、封堵和遏制破损焊缝的进一步泄漏。本文针对该组件的结构特点,分析关键零件的受力情况,利用有限元软件验证该组件的封堵效果。结果表明,该组件能有效降低焊缝区域的应力水平。同时,通过对比分析预紧力、配合面角度和摩擦系数等因素对焊缝应力水平的影响趋势,得出了各因素的影响关系曲线和最佳参数。所分析数据可应用于结构设计、加工制造和装配工艺,使该组件达到最佳的封堵效果。%Control rod drive mechanism (CRDM ) lower Canopy welding zone could leak under the influence of certain factors in some pressurized water reactors .The plugging assembly mounted on the welding externally can prevent and seal the leakage .Accord‐ing to the structure characteristics of the components , the stress conditions of key components were analyzed and the plugging effect of the assembly was verified by using the finite element software .The results show that the component can effectively reduce the stress of welding zone .At the same time ,the stress level trends due to different pre‐tightening forces , contact surface angle and friction were analyzed , so that the relationship curves and the optimum parameters were educed .The analysis data can be applied to the structure design , manufacture and assembly process , for making the component to achieve the best effect of plugging .

  14. 控制棒驱动机构耐压壳下部密封环应力与疲劳分析%Stress and Fatigue Analysis for Lower Joint of Control Rod Drive Mechanisms Seal House

    Institute of Scientific and Technical Information of China (English)

    邵雪娇; 张丽屏; 杜娟; 谢海

    2013-01-01

    Two kinds of seal houses for control rod drive mechanisms which have different thickness of the lower seal ring was analyzed for its stress and fatigue by finite elemet method.In the fatigue compution,all the transitions were grouped into several groups,and then the elastoplastic strain correction factor was modified by analyzing thermal and mechanical load separately referring the rules of RCC-M 2002.The results show that the structure with thicker seal ring behaves more safely than the other one except in the second condition.Meanwhile,the amplify of the primary and secondary stress as well as fatigue usage factor can be reduced by regrouping the transients.The precision of fatigue usage factor can be elevated using modified Ke when the amplify of the primary and secondary stress is large to some extent produced by both thermal and mechanical loads.%采用有限元分析方法对某核电工程控制棒驱动机构耐压壳下部密封环的2种对接厚度进行了应力和疲劳分析对比,在疲劳分析中采用瞬态分组技术,同时参考RCC-M 2002规范对ANSYS程序中的弹塑性修正因子(Ke)进行解耦修正.结果表明,2种接头厚度的分析结果均满足RCC-M规范中的应力评定准则,其中,较薄密封环结构疲劳分析结果相对更安全,较厚密封环结构在其余工况相对更安全;在疲劳分析中对瞬态进行分组能明显降低使用系数和一次加二次应力之和幅值的保守性;在热和机械共同作用的一次加二次应力之和的幅值较高时,对Ke的修正能明显提高计算结果精度.

  15. 控制棒驱动机构管座法兰母材性能评估%Evaluation on Material Used for Control Rod Drive Mechanism Socket Flange

    Institute of Scientific and Technical Information of China (English)

    洪源平; 马新朝; 李世伟; 步伟东; 余伟炜

    2014-01-01

    针对控制棒驱动机构(CRDM)下部Ω环的热电偶(TC)管座母材在水压试验后出现缺陷的问题,参照 CRDM/TC 管座法兰母材的采购技术规范要求,对管座母材进行材质分析,比较了业主与制造厂提供的同批次材料性能差异。分析结果显示,业主提供的部分批次理化检验余料除晶粒度不符合采购技术规范外,内部还有较大尺寸夹杂,但材料的其他性能指标均满足采购规范技术要求。原材料内部存在混晶现象,材质性能分布不均匀。为此,结合材料失效机理研究了母材应力腐蚀裂纹扩展性能,试验结果表明,在一回路高温高压水环境下,超标夹杂的存在未对材料的应力腐蚀裂纹扩展速率产生明显影响。%Defects were introduced to the Ω seal hoop in bottom of the control rod drive mechanism (CRDM)after hydrostatic test.So the base material of the Ω hoop was analyzed and the differences of mechanical properties between the materials from proprietor and manufacturer are compared according to the purchase technical specification of CRDM/TC tube socket flange.Analysis results show that the grain size and the inclusion of the material from the proprietor are disqualification.In addition,there is mixed crystal in the material,thus causing an inhomogeneous performance for the material.Therefore,the stress corrosion crack propagation was investigated on the CRDM/TC tube socket flange base material.It is found that the stress corrosion crack propagation rate is not significantly affected by the inclusion in the material under the actual condition of the primary coolant circuit.

  16. Study on Electromagnetism Force of CARR Control Rod Drive Mechanism Experimental Machine%CARR控制棒驱动机构试验样机电磁力研究

    Institute of Scientific and Technical Information of China (English)

    朱学微; 甄建霄; 王玉林; 殷浩哲; 贾月光; 杨坤

    2015-01-01

    With the aim of acquiring electromagnetic force and electromagnetic field dis‐tributions of control rod drive mechanism (CRDM ) in China Advanced Research Reactor (CARR) ,the force analysis on the CRDM was taken .Manufacturing the experimental machine ,the electromagnetic force experiment was taken on it . The electromagnetic field and electromagnetic force simulation analyses of experimental machine were taken , working out distribution data of electromagnetic force and magnetic induction intensity distribution curve ,and the effects of permanent magnetic field on electromagnetic field and structure parameters on electromagnetic force .The simulation value is accord with experiment value ,the research results provide a reference to electromagnetic force study on CRDM in CARR ,and also provide a reference to design of the same type CRDM .%为获得中国先进研究堆(CARR)控制棒驱动机构电磁力及电磁场分布,对控制棒驱动机构进行了受力分析,设计制造CARR控制棒驱动机构试验样机。在试验样机上进行了电磁力试验,使用有限元分析软件Ansoft Maxwell对试验样机电磁场及电磁力分布进行仿真分析,计算得到了试验样机磁感应强度分布曲线和电磁力分布数据,以及永磁体磁场对电磁场的影响和结构参数对电磁力的影响。仿真结果与试验值符合较好,研究结果为CARR控制棒驱动机构电磁力研究和同类型控制棒驱动机构设计提供了参考。

  17. Hollow sucker rod for PCP systems

    Energy Technology Data Exchange (ETDEWEB)

    Villasante, J.A.; Ernst, H.A. [Tenaris Research and Development, Campana (Argentina)

    2008-07-01

    This paper described a new hollow sucker rod technology designed for use with progressive cavity pumps (PCPs). The technology provided a high torque load to yielding ratio, as well as high backspin resistance and pumping rates. The technology was also designed to allow for the injection of other fluids such as corrosion inhibitors or diluents via its hollow sucker rod. Torsion, axial, and bending load stress analyses were conducted to determine critical zones in a top hollow rod connection at the well head and a bottom hollow rod connection at the well bottom. The study showed that the ratio between the equivalent stress and ultimate tensile stress was a function of torsional load. Backspin analyses were conducted to determine the release of energy accumulated in the hollow rod and traditional pumping system. The evaluation showed that the elastic torsional deformation was lower in the hollow rod system, while backspin resistance was higher. Multiple make and break operations were conducted to determine torsional load values. Results from the study were used to optimize the hollow rod technology. It was concluded that the hollow sucker rod system is now being used in various configurations at sites around the world. 8 tabs., 14 figs.

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

    International Nuclear Information System (INIS)

    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)

  19. Film cooling of vertical fuel rods

    International Nuclear Information System (INIS)

    Spray cooling of vertical rods has been studied at low heat fluxes appropriate to the removal of fission product heating following a reactor shut down. A series of tests have been made at atmospheric pressure using electrically heated rods, both singly and in a seven rod cluster, cooled by a falling film of water. Four modes of film breakdown were observed; progressive evaporation of the film; dry-patch formation due to surface tension effects at high inlet subcooling; stripping of the film by the flooding action of counterflow steam; and the disruption of the film on a hot rod caused by sputtering. Each of these phenomena is described in relation to the application of film cooling to long vertical fuel rod clusters. (author)

  20. Attachment for sucker rod depth adjustment

    Energy Technology Data Exchange (ETDEWEB)

    Collins, N.D.

    1992-04-07

    This patent describes a surface unit of an oil well pumping system, having a walking beam, a suspended carrier bar and an interconnected sucker rod assembly for stroking a reciprocating down-hole pump. It comprises a cross bar having a centrally located passage therein for the sucker rod assembly and adapted to be transversely supported by the carrier bar; a depth adjusting bar, having a centrally located passage therein for the sucker rod assembly, positioned at a selected fixed dimension above and parallel to the cross bar and adapted to operatively support the sucker rod assembly; clamping means for fixing the sucker rod relative to the depth adjusting bar; and hydraulically extendable means supportively connecting the depth adjusting bar to the cross bar on at least each side of the carrier bar for adjusting the selected fixed dimension and maintaining the adjustment during operation.

  1. The sROD module for the ATLAS Tile Calorimeter upgrade demonstrator

    CERN Document Server

    Carrio Argos, Fernando; The ATLAS collaboration

    2014-01-01

    This work presents the first prototype of the super Read-Out Driver (sROD) demonstrator board for the Tile Calorimeter Demonstrator project. This project aims to test the new readout electronics architecture for the Phase 2 Upgrade of the ATLAS Tile Calorimeter, replacing the front-end electronics of one complete drawer with the new electronics during the Long Shutdown 1 (2013-2014), in order to evaluate its performance. The sROD demonstrator board will receive and process data from a complete module. Moreover the sROD demonstrator board will send preprocessed data to the present trigger system, and will transmit trigger control and timing information (TTC) and Detector Control System (DCS) commands to the front-end. A detailed description of the sROD board design, firmware and control and data acquisition software. We also will present the first results of this module during the commissioning of the upgraded TileCal module.

  2. Conceptual design of a passive, inherently safe emergency shutdown rod for high-temperature reactor applications

    International Nuclear Information System (INIS)

    The concept of a passive, inherently safe, and fail-safe design for an emergency control rod is presented. The functioning of the rod is based solely on inexorable physical laws. The operation of the rod in its emergency function does not require the intervention of a human operator, nor does it rely on any signal from a monitoring or safety system. Although the concept could be applicable to a variety of reactors (provided a normal temperature range is specified), in this paper, the concept is applied to the emergency shutdown of a pebble-bed reactor. The preliminary study presented here demonstrates that the proposed Electro-Magnetic Optimally Scramming Control Rod (EM-OSCR) naturally operates when needed. The rod is held out of the core region by the force of an electromagnet. The force is generated by a current carried by a conductor, a portion of which passes near or through the reactor core region. When the temperature in the conductor increases because of an increase in temperature in the reactor, the conductor resistivity increases. This, in turn, leads to a current decrease. When the current decreases below the level necessary to hold the rod up, the rod is released and it falls into the core under the effect of gravity. (author)

  3. A feasibility study for the application of enriched gadolinia burnable absorber rods in nuclear core design

    International Nuclear Information System (INIS)

    An analysis model using MICBURN-3/CASMO-3 is established for the enriched gadolinia burnable absorber rods. A homogenized cross section editing code, PROLOG, is modified so that it can handle such a fuel assembly that includes two different types of gadolinia rods. Study shows that Gd-155 and Gd-157 are almost same in suppressing the excess reactivity and it is recommended to enrich both odd number isotopes, Gd-155 and Gd-157. It is estimated that the cycle length increases by 2 days if enriched gadolinia rods are used in the commercial nuclear power plant such as YGN-3 of which the cycle length is assumed 2 years. For the advanced integral reactor SMART in which ultra long cycle length and soluble boron-free operation concept is applied, natural gadolinia burnable absorber rods fail to control the excess reactivity. On the other hand, enriched gadolinia rods are successful in controling the excess reactivity. To minimize power peakings, various placements of gadolinia rods are tested. Also initial reactivity holddown and gadolinia burnout time are parametrized with respect to the number of gadolinia rods and gadolinia weight fractions

  4. 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.

  5. Welding of zircaloy tubes for fuel rods fabrication

    International Nuclear Information System (INIS)

    The TIG process form welding zircaloy tube-end plug to fabricate UO2 pellet fuel rods is presented. The zircaloy tube is used as weld by welding zircaloy end-plugs inserted into ends. The technical aspects of this welding process, necessary equipment, and weld properties are discussed. The main problems involved with this type of weld and the nondestructive and destructive weld quality controls are emphasized. (Author)

  6. The sROD Demonstrator for the ATLAS Tile Calorimeter Upgrade

    CERN Document Server

    Carrió, F; The ATLAS collaboration; Ferrer, A; Fiorini, L; González, V; Hernández, Y; Higon, E; March, L; Moreno, P; Qin, G; Sanchis, E; Solans, C; Valero, A; Valls, J A

    2012-01-01

    This work presents the early design of the super Read-Out Driver (sROD) demonstrator board for the Tile Calorimeter Demonstrator project. This project aims to test the new readout electronics architecture for the Phase 2 Upgrade of the ATLAS Tile Calorimeter, replacing the front-end electronics of one complete drawer with the new electronics during the Long Shutdown 2013, in order to evaluate its performance. The sROD demonstrator board will receive and process data from 48 channels. Moreover sROD demonstrator board will send preprocessed data to the present trigger system, and will transmit trigger control and timing information (TTC) and Detector Control System (DCS) commands to the front-end. An overview on the implementation and electronics design of sROD demonstrator board for the Demonstrator project is presented here.

  7. Seal-welding detection device for fuel rod and detection method therefor

    International Nuclear Information System (INIS)

    The present invention provides a method of and a device for detecting presence or absence of abnormality of welded portions (nugget portions) for sealing an end plug sealing hole of a fuel rod. Namely, the end face of the fuel rod is photographed, and the nugget region is detected based on the photographed images by a nugget boundary recognition means. The region to be sealed is determined as a control-range mark on the image of the end face of the fuel rod. Whether the control-range mark is included in the nugget portion or not is compared and evaluated as a comparing and judging means. Then, the presence or absence of the abnormality for the seal in the nugget portion of the fuel rod end plug can be visually monitored automatically and continuously. In addition, since the remote-detection can be conducted and in non-contact manner, operator's exposure can be eliminated. (I.S.)

  8. The Mechanical Effect of Rod Contouring on Rod-Screw System Strength in Spine Fixation

    Science.gov (United States)

    Karakasli, Ahmet; Karaarslan, Ahmet A.; Ozcanhan, Mehmet Hilal; Ertem, Fatih; Erduran, Mehmet

    2016-01-01

    Objective Rod-screw fixation systems are widely used for spinal instrumentation. Although many biomechanical studies on rod-screw systems have been carried out, but the effects of rod contouring on the construct strength is still not very well defined in the literature. This work examines the mechanical impact of straight, 20° kyphotic, and 20° lordotic rod contouring on rod-screw fixation systems, by forming a corpectomy model. Methods The corpectomy groups were prepared using ultra-high molecular weight polyethylene samples. Non-destructive loads were applied during flexion/extension and torsion testing. Spine-loading conditions were simulated by load subjections of 100 N with a velocity of 5 mm min-1, to ensure 8.4-Nm moment. For torsional loading, the corpectomy models were subjected to rotational displacement of 0.5° s-1 to an end point of 5.0°, in a torsion testing machine. Results Under both flexion and extension loading conditions the stiffness values for the lordotic rod-screw system were the highest. Under torsional loading conditions, the lordotic rod-screw system exhibited the highest torsional rigidity. Conclusion We concluded that the lordotic rod-screw system was the most rigid among the systems tested and the risk of rod and screw failure is much higher in the kyphotic rod-screw systems. Further biomechanical studies should be attempted to compare between different rod kyphotic angles to minimize the kyphotic rod failure rate and to offer a more stable and rigid rod-screw construct models for surgical application in the kyphotic vertebrae. PMID:27651858

  9. Tipping time of a quantum rod

    Energy Technology Data Exchange (ETDEWEB)

    Parrikar, Onkar [Birla Institute of Technology and Science-Pilani, Goa campus, Zuarinagar, Goa 4032726 (India)], E-mail: onkarsp@gmail.com

    2010-03-15

    The behaviour of a quantum rod, pivoted at its lower end on an impenetrable floor and restricted to moving in the vertical plane under the gravitational potential, is studied analytically under the approximation that the rod is initially localized to a 'small-enough' neighbourhood around the point of classical unstable equilibrium. It is shown that the rod evolves out of this neighbourhood. The time required for this to happen, i.e. the tipping time, is calculated using the semi-classical path integral. It is shown that equilibrium is recovered in the classical limit, and that our calculations are consistent with the uncertainty principle.

  10. Morphological and biochemical studies of canine progressive rod-cone degeneration. 3H-fucose autoradiography

    International Nuclear Information System (INIS)

    Visual cell pathology and rod outer segment renewal were investigated in normal and PRCD-affected miniature poodles using 3H-fucose autoradiography. Twenty-four hours following the intravitreal injection of 3H-fucose, label accumulated diffusely over cone OS and in a banded pattern at the rod OS base. In normal rods, the band of 3H-label was displaced sclerad with time. PRCD-affected rods in the early stages of the disease (stages 0-1) also showed a similar 3H-label pattern but a significantly (P less than 0.001) reduced renewal rate (control = 2.35 +/- 0.43 mu/24 hr; affected = 0.99 +/- mu/24 hr). This abnormal renewal rate was present in central, equatorial, and peripheral visual cells and was not associated with the presence or density of pigment in the RPE cell layer. Biochemical studies indicated that the 3H-label was present as an integral membrane component in the rod OS and confirmed that canine rhodopsin is a fucosylated glycoprotein. The 3H-band in the rod OS layer disappeared in stage 2 of the disease; diffuse label now was present over rod OS that had decreased length and were reduced in number. At this stage of the disease, interphotoreceptor space was invaded by phagocytic cells, and photoreceptor nuclei were lost from the outer nuclear layer. These late degenerative changes were more extensive in the superior and inferior retinal meridians

  11. Increasing life times of sucker rod pumps; Verlaengerung des Behandlungsintervalles bei Tiefpumpsonden

    Energy Technology Data Exchange (ETDEWEB)

    Oberndorfer, M. [OMV AG, Vienna (Austria); Winter, U. [OMV AG, Vienna (Austria)

    1996-06-01

    As the oil-price and the US Dollar have been going down within the past years and production becomes more and more complicated it is necessary to reduce production costs. The cooperation between different departments led to a great progress in minimizing well-failures and therewith service life of sucker rod pumping increased. Starting from statistics on service life and failure some technical measurements are presented and discussed with regard to their efficacy. These include: - Sucker rod pumping (corrosion-resistant materials, appropriate use of piston and barrel, sand wiper, gaslock breaker), - sucker rods (use of high strength sucker rods -grade D instead of grade C-, use of sinker bars, exploration of wear of sucker rods caused by different sucker rod couplings and rod guides), - tubings (tubings made of J55 quality are normalized after the upsetting action by the producer according to an internal company specification, installation of glass-fiber-reinforced-tubings, test of tubing-rotators) and - reservoir fluid (change of corrosion inhibitor after careful tests in laboratory and field, more efficient control of parrafin through hot water and liquid hydrocarbons). (orig.) [Deutsch] Der Verfall des Oelpreises und US Dollars innerhalb der letzten Jahre sowie die stetig komplizierter werdenden Foerderbedingungen verlangen eine permanente Reduzierung der Foerderkosten. Durch das integrative Zusammenarbeiten verschiedenster Betriebsabteilungen sind in den letzten Jahren grosse Fortschritte bei der Schadensminimierung und der damit verbundenen Standzeiterhoehung der Gestaengetiefpumpinstallationen erzielt worden. Ausgehend von statistischen Uebersichten der Laufzeit- und Schadensentwicklungen werden verschiedene foerdertechnische Massnahmen hinsichtlich ihrer Wirksamkeit vorgestellt und diskutiert. (orig.)

  12. Anisotropy in CdSe quantum rods

    Energy Technology Data Exchange (ETDEWEB)

    Li, Liang-shi

    2003-09-01

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

  13. Computer simulation of rod-sphere mixtures

    CERN Document Server

    Antypov, D

    2003-01-01

    Results are presented from a series of simulations undertaken to investigate the effect of adding small spherical particles to a fluid of rods which would otherwise represent a liquid crystalline (LC) substance. Firstly, a bulk mixture of Hard Gaussian Overlap particles with an aspect ratio of 3:1 and hard spheres with diameters equal to the breadth of the rods is simulated at various sphere concentrations. Both mixing-demixing and isotropic-nematic transition are studied using Monte Carlo techniques. Secondly, the effect of adding Lennard-Jones particles to an LC system modelled using the well established Gay-Berne potential is investigated. These rod-sphere mixtures are simulated using both the original set of interaction parameters and a modified version of the rod-sphere potential proposed in this work. The subject of interest is the internal structure of the binary mixture and its dependence on density, temperature, concentration and various parameters characterising the intermolecular interactions. Both...

  14. Probabilistic assessment for nuclear fuel rods behavior

    International Nuclear Information System (INIS)

    BACO is a code for the simulation of the thermo-mechanical and fission gas behavior of a cylindrical fuel rod under operation conditions. Input parameters and, therefore, output ones may include statistical dispersion. In this paper, experimental CANDU fuel rods irradiated at the NRX reactor together with experimental MOX fuel rods and the IAEA'CRP FUMEX cases are used in order to determine the sensitivity of BACO code predictions. We analyze the CARA and CAREM fuel rods relation between predicted performance and statistical dispersion in order of enhanced their original designs. These exercises show the sensitivity of the predictions concerning such parameters and the extended features of the BACO code for a probability study. (author)

  15. Rod bundle burnout data and correlation comparisons

    International Nuclear Information System (INIS)

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

  16. High Power Performance of Rod Fiber Amplifiers

    DEFF Research Database (Denmark)

    Johansen, Mette Marie; Michieletto, Mattia; Kristensen, Torben;

    2015-01-01

    An improved version of the DMF rod fiber is tested in a high power setup delivering 360W of stable signal power. Multiple testing degrades the fiber and transverse modal instability threshold from >360W to ~290W.......An improved version of the DMF rod fiber is tested in a high power setup delivering 360W of stable signal power. Multiple testing degrades the fiber and transverse modal instability threshold from >360W to ~290W....

  17. Self-diagnosing braided composite rod

    OpenAIRE

    Fangueiro, Raúl; Zdraveva, E.; Pereira, Cristiana Gonilho; Ferreira, A.; Lanceros-Méndez, S.

    2010-01-01

    This paper presents the development of a braided reinforced composite rod (BCR) able to both reinforce and monitor the stress state of concrete structures. Carbon fibers have been used as sensing and reinforcing materials along with glass fiber. Various composites rods have been produced using an author patented technique based on a modified conventional braiding machine. The materials investigated were prepared with different carbon fiber content as follows: BCR2 (77% glass/23...

  18. A methodology for obtaining the control rod patterns in a BWR using genetic algorithms; Una metodologia para obtener los patrones de barras de control en un BWR usando algoritmos geneticos

    Energy Technology Data Exchange (ETDEWEB)

    Ortiz S, J.J.; Montes T, J.L. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico); Requena R, I. [Universidad de Granada, 18071 Granada (Spain)]. e-mail: jjortiz@nuclear.inin.mx

    2003-07-01

    In this work the GACRP system based on the genetic algorithms technique for the obtaining of the drivers of control bars in a BWR reactor is presented. This methodology was applied to a transition cycle and a one of balance of the Laguna Verde nuclear power station (CNLV). For each one of the studied cycles, it was executed the methodology with a fixed length of the cycle and it was compared the effective multiplication factor of neutrons at the end of the cycle that it is obtained with the proposed drivers of control bars and the multiplication factor of neutrons obtained by means of a Haling calculation. It was found that it is possible to extend several days the length of both cycles with regard to the one Haling calculation. (Author)

  19. Close packing of rods on spherical surfaces

    Science.gov (United States)

    Smallenburg, Frank; Löwen, Hartmut

    2016-04-01

    We study the optimal packing of short, hard spherocylinders confined to lie tangential to a spherical surface, using simulated annealing and molecular dynamics simulations. For clusters of up to twelve particles, we map out the changes in the geometry of the closest-packed configuration as a function of the aspect ratio L/D, where L is the cylinder length and D the diameter of the rods. We find a rich variety of cluster structures. For larger clusters, we find that the best-packed configurations up to around 100 particles are highly dependent on the exact number of particles and aspect ratio. For even larger clusters, we find largely disordered clusters for very short rods (L/D = 0.25), while slightly longer rods (L/D = 0.5 or 1) prefer a global baseball-like geometry of smectic-like domains, similar to the behavior of large-scale nematic shells. Intriguingly, we observe that when compared to their optimal flat-plane packing, short rods adapt to the spherical geometry more efficiently than both spheres and longer rods. Our results provide predictions for experimentally realizable systems of colloidal rods trapped at the interface of emulsion droplets.

  20. Close packing of rods on spherical surfaces.

    Science.gov (United States)

    Smallenburg, Frank; Löwen, Hartmut

    2016-04-28

    We study the optimal packing of short, hard spherocylinders confined to lie tangential to a spherical surface, using simulated annealing and molecular dynamics simulations. For clusters of up to twelve particles, we map out the changes in the geometry of the closest-packed configuration as a function of the aspect ratio L/D, where L is the cylinder length and D the diameter of the rods. We find a rich variety of cluster structures. For larger clusters, we find that the best-packed configurations up to around 100 particles are highly dependent on the exact number of particles and aspect ratio. For even larger clusters, we find largely disordered clusters for very short rods (L/D = 0.25), while slightly longer rods (L/D = 0.5 or 1) prefer a global baseball-like geometry of smectic-like domains, similar to the behavior of large-scale nematic shells. Intriguingly, we observe that when compared to their optimal flat-plane packing, short rods adapt to the spherical geometry more efficiently than both spheres and longer rods. Our results provide predictions for experimentally realizable systems of colloidal rods trapped at the interface of emulsion droplets. PMID:27131565

  1. Annular burnout data from rod bundle experiments

    International Nuclear Information System (INIS)

    Burnout data for annular flow in a rod bundle are presented for both transient and steady-state conditions. Tests were performed at the Oak Ridge National Laboratory in the Thermal Hydraulic Test Facility (THTF), a pressurized-water loop containing an electrically heated 64-rod bundle. The bundle configuration is typical of later generation pressurized-water reactors with 17 x 17 fuel arrays. Both axial and radial power profiles are flat. All experiments were carried out in upflow with subcooled inlet conditions, insuring accurate flow measurement. Conditions within the bundle were typical of those which could be encountered during a nuclear reactor loss-of-coolant accident. Level average fluid conditions within the test section were calculated using steady-state mass and energy conservation considerations for the steady-state tests and a transient, homogeneous, equilibrium computer code for the transient tests. Unlike tube dryout, burnout within a rod bundle does not necessarily occur at one distinct axial level. The location of individual rod dryout was determined by scanning rods axially and locating the position where rod superheat increased from approx. =0 to 30 K or greater. Thermocouple instrumentation within the bundle allows the location of dryout to be determined to within approximately +.5 cm for many of the tests

  2. Hydraulic Rod Drives for the CAREM Reactor

    International Nuclear Information System (INIS)

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

  3. 49 CFR 230.93 - Pistons and piston rods.

    Science.gov (United States)

    2010-10-01

    ... 49 Transportation 4 2010-10-01 2010-10-01 false Pistons and piston rods. 230.93 Section 230.93... Tenders Driving Gear § 230.93 Pistons and piston rods. (a) Maintenance and testing. Pistons and piston rods shall be maintained in safe and suitable condition for service. Piston rods shall be inspected...

  4. International symposium on fuel rod simulators: development and application

    Energy Technology Data Exchange (ETDEWEB)

    McCulloch, R.W. (comp.)

    1981-05-01

    Separate abstracts are included for each of the papers presented concerning fuel rod simulator operation and performance; simulator design and evaluation; clad heated fuel rod simulators and fuel rod simulators for cladding investigations; fuel rod simulator components and inspection; and simulator analytical modeling. Ten papers have previously been input to the Energy Data Base.

  5. Mechanical performance of fiberglass sucker-rod strings

    Energy Technology Data Exchange (ETDEWEB)

    Tripp, H.A.

    1988-08-01

    The natural frequencies of fiberglass sucker-rod strings can be calculated by treating the rod strings as modified spring/mass vibration systems. The ratio of the pumping-unit operating speed to the rod-string natural frequency can then be used as a basis for understanding fiberglass-rod performance and for predicting downhole pump stroke lengths.

  6. 秦山三期60Co调节棒提棒后控制棒驱动机构气空间氘气可燃性安全评价%Safety Assessment on Flammability of Deuterium in Air Space of CRDM After Withdrawal of 60Co Control Rod for Qinshan Phase Ⅲ

    Institute of Scientific and Technical Information of China (English)

    方立凯; 丁捷; 付亚茹

    2012-01-01

    由于控制棒驱动机构气空间内的氘气不能完成参与气体复合,而60Co调节棒提出重水液面后棒表面温度较高,因此,需分析气空间内氘气的可燃性.本文采用理论计算加试验的方法,分别确定了60Co调节棒提出重水液面后的温度及氘气混合气的可燃浓度和所需的最小点火温度.通过分析表明,在正常运行20个月内,60Co调节棒提棒不会造成氘气混合气的爆燃.%Since the deuterium (D2) in CRDM can't be recombined completely, the surface temperature of the 60Co control rod is high when it is withdrawn, and it is required to evaluate the flammability of D2 mixture gas. In this article, theoretical and test method were used to determine the surface temperature of the 80 Co control rod when it was withdrawn, the minimum flammability volume fraction and ignition temperature, respectively. The results show it is unlikely that the high surface temperature can be the potential for D2 ignition.

  7. Local Fuel Rod Crud Prediction Tool Applications

    Energy Technology Data Exchange (ETDEWEB)

    Krammen, Michael A.; Karoutas, Zeses E.; Wang, Guoqiang; Young, Michael Y

    2009-06-15

    A code system with attendant methods has been developed for modeling local fuel rod crud. This tool is used to perform the Crud Induced Localized Corrosion (CILC) risk assessment recommended by the EPRI crud and corrosion guidelines, which were developed in response to the INPO zero fuel failures by 2010 initiatives. The methodology is in production use. This paper will describe the range of problems the methodology has already been applied to and the especial pertinence to low duty fuel applications. The methodology begins with Computational Fluid Dynamics (CFD) computations over a fuel assembly grid span. The CFD results provide detailed relative variations in local heat transfer coefficient over the grid span. These very local relative variations are used to determine very local thermal hydraulic conditions over the entire axial length of every fuel rod in a reactor core over the life of the rod in reactor. The expansion using the local relative variations is currently accomplished with the HIDUTYDRV code. The very local thermal hydraulic conditions are combined with reactor coolant crud concentrations derived from EPRI BOA analysis as input to models for predicting very local fuel rod crud deposition. The reactor coolant crud concentrations are determined over each reactor cycle by reactor system wide crud mass balance calculations. The reactor coolant crud concentrations are used to calculate local crud thickness using mass transfer models which are a function of the local thermal conditions. The advanced crud deposition models also include models for calculating local crud dryout. Local crud deposition and crud dryout are strongly dependent on very local boiling or steaming, which are predicted through the translation of the CFD results. The local crud thickness and degree of local crud dryout are key factors in determining the margin or risk for local fuel rod cladding crud induced fuel failure. The development and first application of these methods was in

  8. Development of burnup dependent fuel rod model in COBRA-TF

    Science.gov (United States)

    Yilmaz, Mine Ozdemir

    predictions. After confirming that the new fuel thermal conductivity model in CTF worked and provided consistent results with FRAPTRAN predictions for a single fuel rod configuration, the same type of analysis was carried out for a bigger system which is the 4x4 PWR bundle consisting of 15 fuel pins and one control guide tube. Steady- state calculations at Hot Full Power (HFP) conditions for control guide tube out (unrodded) were performed using the 4x4 PWR array with CTF/TORT-TD coupled code system. Fuel centerline, surface and average temperatures predicted by CTF/TORT-TD with and without the new fuel thermal conductivity model were compared against CTF/TORT-TD/FRAPTRAN predictions to demonstrate the improvement in fuel centerline predictions when new model was used. In addition to that constant and CTF dynamic gap conductance model were used with the new thermal conductivity model to show the performance of the CTF dynamic gap conductance model and its impact on fuel centerline and surface temperatures. Finally, a Rod Ejection Accident (REA) scenario using the same 4x4 PWR array was run both at Hot Zero Power (HZP) and Hot Full Power (HFP) condition, starting at a position where half of the control rod is inserted. This scenario was run using CTF/TORT-TD coupled code system with and without the new fuel thermal conductivity model. The purpose of this transient analysis was to show the impact of thermal conductivity degradation (TCD) on feedback effects, specifically Doppler Reactivity Coefficient (DRC) and, eventually, total core reactivity.

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

    International Nuclear Information System (INIS)

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

  10. Neuronatin is a stress-responsive protein of rod photoreceptors.

    Science.gov (United States)

    Shinde, Vishal; Pitale, Priyamvada M; Howse, Wayne; Gorbatyuk, Oleg; Gorbatyuk, Marina

    2016-07-22

    Neuronatin (NNAT) is a small transmembrane proteolipid that is highly expressed in the embryonic developing brain and several other peripheral tissues. This study is the first to provide evidence that NNAT is detected in the adult retina of various adult rod-dominant mammals, including wild-type (WT) rodents, transgenic rodents expressing mutant S334ter, P23H, or T17M rhodopsin, non-human primates, humans, and cone-dominant tree shrews. Immunohistochemical and quantitative real time polymerase chain reaction (qRT-PCR) analyses were applied to detect NNAT. Confocal microscopy analysis revealed that NNAT immunofluorescence is restricted to the outer segments (OSs) of photoreceptors without evidence of staining in other retinal cell types across all mammalian species. Moreover, in tree shrew retinas, we found NNAT to be co-localized with rhodopsin, indicating its predominant expression in rods. The rod-derived expression of NNAT was further confirmed by qRT-PCR in isolated rod photoreceptor cells. We also used these cells to mimic cellular stress in transgenic retinas by treating them with the endoplasmic reticulum stress inducer, tunicamycin. Thus, our data revealed accumulation of NNAT around the nucleus as compared to dispersed localization of NNAT within control cells. This distribution coincided with the partial intracellular mislocalization of NNAT to the outer nuclear layer observed in transgenic retinas. In addition, stressed retinas demonstrated an increase of NNAT mRNA and protein levels. Therefore, our study demonstrated that NNAT is a novel stress-responsive protein with a potential structural and/or functional role in adult mammalian retinas. PMID:27109921

  11. Actuator system history of safety rod lower latch problems review of latch inspection video tapes

    Energy Technology Data Exchange (ETDEWEB)

    Banks, J.J.

    1992-06-24

    During pre-restart testing the safety rod at position X26-YlO bound after being driven approximately two (2) feet out of the reactor. Subsequently, the rod was manually returned to it's seated position. Inspection of the lower latch showed that the latch locking plunger button (screwed on to the bottom of the plunger shaft and retained by a pin through a hole drilled through the button and the plunger shaft) was missing. The shaft failed through the hole drilled for the retaining pin. The button, with the retaining pin intact, was found lodged between the safety rod upper adapter collar and the top of the safety rod thimble top fitting. Analysis of the safety rod latch and accompanying forest guide tube design provided assurance that this type of failure would not cause binding during the scramming'' of the safety rods. Inspection of all of the K'' safety rod lower latches revealed six other latches with missing plunger buttons, and nine with other non-conformances which required latch replacement. A history search conducted by Reactor Engineering Design, Components Handling Group, is included in this report. The history search shows that latch design modifications, as a part of initial development of the latch system and later to improve the delatching operation, were made from 1950 to 1960. These modifications created a condition where latch damage could occur. Video tapes were made during inspection of the safety rod latches in K area and control rod latches in L area. These tapes were reviewed by Reactor Engineering Design Components Handling engineers. The reviews were used for correlation of latch problems reported by the engineers/mechanics making the inspections. The K area tapes showed inspection of 65 of the 66 safety rod latches. The review of the tapes showed the plunger buttons to be missing from five latches. RED-CH reviewed the L Area video taped inspection of 35 control rod clusters (245 latches). No non-conformances were

  12. Improvement in J{sub c} performance below liquid nitrogen temperature for SmBa{sub 2}Cu{sub 3}O{sub y} superconducting films with BaHfO{sub 3} nano-rods controlled by low-temperature growth

    Energy Technology Data Exchange (ETDEWEB)

    Miura, S., E-mail: miura-syun12@ees.nagoya-u.ac.jp; Yoshida, Y.; Ichino, Y.; Xu, Q. [Department of Energy Engineering and Science, Nagoya University, Nagoya 464-8603 (Japan); Matsumoto, K. [Department of Materials Science and Engineering, Kyushu Institute of Technology, Kitakyushu 804-8550 (Japan); Ichinose, A. [Electric Power Engineering Research Laboratory, Central Research Institute of Electric Power Industry, Yokosuka, Kanagawa 240-0196 (Japan); Awaji, S. [Institute for Materials Research, Tohoku University, Sendai 980-8577 (Japan)

    2016-01-01

    For use in high-magnetic-field coil-based applications, the critical current density (J{sub c}) of REBa{sub 2}Cu{sub 3}O{sub y} (REBCO, where RE = rare earth) coated conductors must be isotropically improved, with respect to the direction of the magnetic field; these improvements must be realized at the operating conditions of these applications. In this study, improvement of the J{sub c} for various applied directions of magnetic field was achieved by controlling the morphology of the BaHfO{sub 3} (BHO) nano-rods in a SmBCO film. We fabricated the 3.0 vol. % BHO-doped SmBCO film at a low growth temperature of 720 °C, by using a seed layer technique (T{sub s} = 720 °C film). The low-temperature growth resulted in a morphological change in the BHO nano-rods. In fact, a high number density of (3.1 ± 0.1) × 10{sup 3} μm{sup −2} of small (diameter: 4 ± 1 nm), discontinuous nano-rods that grew in various directions, was obtained. In J{sub c} measurements, the J{sub c} of the T{sub s} = 720 °C film in all directions of the applied magnetic field was higher than that of the non-doped SmBCO film. The J{sub c}{sup min} (6.4 MA/cm{sup 2}) of the former was more than 6 times higher than that (1.0 MA/cm{sup 2}) of the latter at 40 K, under 3 T. The aforementioned results indicated that the discontinuous BHO nano-rods, which occurred with a high number density, exerted a 3D-like flux pinning at the measurement conditions considered. Moreover, at 4.2 K and under 17 T, a flux pinning force density of 1.6 TN/m{sup 3} was realized; this value was comparable to the highest value recorded, to date.

  13. Single Rod Vibration in Axial Flow

    Science.gov (United States)

    Weichselbaum, Noah; Wang, Shengfu; Bardet, Philippe

    2013-11-01

    Fluid structure interaction of a single rod in axial flow is a coupled dynamical system present in many application including nuclear reactors, steam generators, and towed antenna arrays. Fluid-structure response can be quantified thanks to detailed experimental data where both structure and fluid responses are recorded. Such datum deepen understanding of the physics inherent to the system and provide high-dimensionality quantitative measurements to validate coupled structural and CFD codes with various level of complexity. In this work, single rods fixed on both ends in a concentric pipe, are subjected to an axial flow with Reynolds number based on hydraulic diameter of Re =4000. Rods of varying material stiffness and diameter are utilized in the experiment resulting in a range of dimensionless U between 0.5 and 1, where U = (ρA/EI)1/2uL. Experimental measurements of the velocity field around the rod are taken with PIV from time-resolved Nd:YLF laser and a high speed CMOS camera. Three-dimensional and temporal vibration and deflection of the rod is recorded with shadowgraphy utilizing two sets of pulsed high power LED and dedicated CMOS camera. Through integration of these two diagnostics, it is possible to reconstruct the full FSI domain providing unique validation data.

  14. A feasibility study on the use of ultrasonic wave-guides for the diverse safety rod (DSR) drop time measurement

    International Nuclear Information System (INIS)

    In Prototype fast breeder reactor (PFBR), apart from control rods, safety rods are also used to effect safety action. It is required to determine the time taken by the safety rod to fall into the core after being de-energized by the electromagnet of the diverse safety rod drive mechanism. In order to measure the drop time, the feasibility of using an ultrasonic transducer on tope of a wave-guide, where near ambient temperature prevails, is being studied. Here, direct use of sodium immersible transducer is ruled out due to certain reliability aspects and space constrains. Different wave-guides like solid rods, bundle of wires, etc., made of stainless steel material were experimented to find their suitability for this application. The results of the above experiments are discussed in this paper. (author)

  15. Differential contribution of rod and cone circadian clocks in driving retinal melatonin rhythms in Xenopus.

    Directory of Open Access Journals (Sweden)

    Naoto Hayasaka

    Full Text Available BACKGROUND: Although an endogenous circadian clock located in the retinal photoreceptor layer governs various physiological events including melatonin rhythms in Xenopus laevis, it remains unknown which of the photoreceptors, rod and/or cone, is responsible for the circadian regulation of melatonin release. METHODOLOGY/PRINCIPAL FINDINGS: We selectively disrupted circadian clock function in either the rod or cone photoreceptor cells by generating transgenic Xenopus tadpoles expressing a dominant-negative CLOCK (XCLΔQ under the control of a rod or cone-specific promoter. Eyecup culture and continuous melatonin measurement revealed that circadian rhythms of melatonin release were abolished in a majority of the rod-specific XCLΔQ transgenic tadpoles, although the percentage of arrhythmia was lower than that of transgenic tadpole eyes expressing XCLΔQ in both rods and cones. In contrast, whereas a higher percentage of arrhythmia was observed in the eyes of the cone-specific XCLΔQ transgenic tadpoles compare to wild-type counterparts, the rate was significantly lower than in rod-specific transgenics. The levels of the transgene expression were comparable between these two different types of transgenics. In addition, the average overall melatonin levels were not changed in the arrhythmic eyes, suggesting that CLOCK does not affect absolute levels of melatonin, only its temporal expression pattern. CONCLUSIONS/SIGNIFICANCE: These results suggest that although the Xenopus retina is made up of approximately equal numbers of rods and cones, the circadian clocks in the rod cells play a dominant role in driving circadian melatonin rhythmicity in the Xenopus retina, although some contribution of the clock in cone cells cannot be excluded.

  16. Rod consolidation at the West Valley Demonstration Project

    International Nuclear Information System (INIS)

    A rod consolidation demonstration with irradiated pressurized water reactor fuel was recently conducted by personnel from Nuclear Assurance Corporation and West Valley Nuclear Services Company at the West Valley Demonstration Project in West Valley, New York. The rod consolidation demonstration involved pulling all of the fuel rods from six fuel Assemblies. In general, the rod pulling proceeded smoothly. The highest compaction ratio attained was 1:8:1. Among the total of 1074 fuel rods were some known degraded rods (they had collapsed cladding, a result of in-reactor fuel densification), but no rods were broken or dropped during the demonstration. One aim was to gather information on the effect of rod consolidation operations on the integrity of the fuel rods during subsequent handling and storage. Another goal was to collect information on the condition and handling of intact, damaged, and failed fuel that has been in storage for an extended period. 9 refs., 8 figs., 1 tab

  17. Fragmentation of an axially impacted slender rod

    Science.gov (United States)

    Ji, W.; Waas, A. M.

    2010-02-01

    Motivated by experimental results on the dynamic buckling and fragmentation of a vertical column impacted by a falling mass, results from an analytical model for dynamic buckling which considers the dynamic interaction between the axial column deformation and the out-of-plane buckling displacements are used to interpret the fragmentation process and the resulting fragment lengths. It is shown that a critical time exists for the rod to undergo fragmentation. At this critical time, the rod deforms in a modulated pattern of waves, setting up the stage for the ensuing fragmentation as a result of induced large curvatures that exceed the critical bending strain of the rod material. The resulting fragment length distributions, which show two characteristics peaks at \\frac{\\lambda}{2} and \\frac{\\lambda}{4} , where λ is a characteristic half-wavelength, are found to compare favorably with the experimental results.

  18. System analysis for sucker-rod pumping

    Energy Technology Data Exchange (ETDEWEB)

    Schmidt, Z.; Doty, D.R.

    1989-05-01

    Pumping free gas in an oil well can significantly decrease the efficiency of a sucker-rod-pumping installation. Pump placement depth and use of a downhole gas/liquid separator (gas anchor) were found to be significant variables in improving the overall efficiency. A procedure is presented that shows when and to what degree the use of a gas anchor improves the efficiency of a sucker-rod pumping system. It was found that at lower pump intake pressures, the gas anchor usually improves efficiency, but at higher pump intake pressures, use of a gas anchor produces no positive effect. Also, elevating the pump to the highest position that still allows proper pump loading was found to reduce the operating costs of a sucker-rod-pumping installation significantly. Finally, a procedure is presented to calculate directly the pump volumetric efficiency and required volumetric pump displacement rate.

  19. Interpretation of calculated forces on sucker rods

    Energy Technology Data Exchange (ETDEWEB)

    Lea, J.F.; Pattillo, P.D. (Amoco Production Research co., Tulsa, OK (United States)); Studenmund, W.R. (Stanford Univ., CA (United States))

    1995-02-01

    The analysis of working loads in a sucker rod string during a pumping cycle has received substantial coverage in the petroleum literature. These load predictions have tended to focus on mechanical design considerations such as excess load and fatigue prediction. In contrast, the current study addresses the issues of buckling associated with working axial/pressure loads in an attempt to clarify the means of both predicting buckling and minimizing its effects. The study begins with a review of the static loads acting near the pump, and proceeds to a discussion of how these loads relate to the tendency of a rod string to buckle on the downstroke. Critical to this discussion is the concept of effective tension. Definition of the effective tension leads to the application of this concept to sinker bar design as a means of mitigating the buckling tendency of a rod string. Key points are reinforced by illustrative examples.

  20. Self-Propelled Rods near Surfaces

    CERN Document Server

    Elgeti, Jens

    2009-01-01

    We study the behavior of self-propelled nano- and micro-rods in three dimensions, confined between two parallel walls, by simulations and scaling arguments. Our simulations include thermal fluctuations and hydrodynamic interactions, which are both relevant for the dynamical behavior at nano- to micrometer length scales. In order to investigate the importance hydrodynamic interactions, we also perform Brownian-dynamics-like simulations. In both cases, we find that self-propelled rods display a strong surface excess in confined geometries. An analogy with semi-flexible polymers is employed to derive scaling laws for the dependence on the wall distance, the rod length, and the propulsive force. The simulation data confirm the scaling predictions.

  1. Failure Analysis of A Fractured Connecting Rod

    OpenAIRE

    Mohammed, M. N.; Omar, M. Z.; Zainuddin Sajuri; A. Salah; M.A. Abdelgnei; Salleh, M. S.

    2012-01-01

    In many cases, the major reason behind or causing catastrophic engine failure is the occurrence of the connecting-rod failure and sometimes, such a failure can be attributed to the broken connecting rod’s shank especially when there is a probability of being pushed through the side of the crank-case, thereby making the engine irreparable. Thus, the major aim of the current work is to analyze the connecting rod failure. The study applied a finite element analysis and metallographic examination...

  2. Minor actinide transmutation on PWR burnable poison rods

    International Nuclear Information System (INIS)

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

  3. FEATURES OF HIGH CARBON WIRE ROD THERMOMECHANICAL WORKING IN A STREAM OF HIGH-SPEED WIRE MILL 150 OF «BSW»

    Directory of Open Access Journals (Sweden)

    V. A. Lutsenko

    2012-01-01

    Full Text Available Using a rolling line wire mill reducing-calibrating block the production of high-carbon wire rod subjected to combined thermomechanical treatment with controlled rolling and cooling, can reduce the spread of mechanical properties, reduce the depth of decarbonized layer with a uniform distribution in the surface of its rod on the perimeter, increasing the technological plasticity of rod in processing, virtually eliminating the formation of hardening structures on the surface during high- speed wire drawing.

  4. RESORBABLE HIGH-STRENGTH ROD FOR FRACTURE INTERNAL FIXATION

    Institute of Scientific and Technical Information of China (English)

    杨团民; 刘淼; 杨爱玲; 石宗利; 邱希江; 李毅; 同志超; 韩月

    2004-01-01

    Objective To find an ideal biomaterial for internal fixation. Methods Forty rabbits with fracture of the femur diaphysis (superiorcondyle) were treated by intramedullary nailing of femur with composites rod of resorbable DL-polylactic acid (PDLLA)-calcium metaphosphate (CMP), while steinmann's pin as control. The fracture healing, the material degradation and its mechanical properties were studied by X-ray films, macroscopic, microscopic and electron microscopic observations. Results No significant inflammatory reaction was found, and all the osteotomies were healed, while material was resorbed. Conclusion The PDLLA-CMP has excellent biocompatibility and mechanical properties, and it can be a promising implant material in orthopaedics surgery.

  5. Estimation of water-water energy reactor fuel rod failure in design basis accidents

    International Nuclear Information System (INIS)

    at the pulse reactors essentially differ from conditions for commercial WWER fuel rods in design RIA, first of all a power level, an initial characteristics of fuel rod, a level of outside pressure etcetera. The main purpose of these experiments consists in a conservative substantiation of fuel rod fragmentation criterion at fast reactivity increase. A preparation works for experimental researches of unirradiated and refabricated WWER-1000 type fuel rods behaviour at the NIR (the name of a research reactor) water loop now are carried out. The conditions of experiments on loading and on coolant parameters simulate design reactivity initiated accident (for WWER-1000 it is an accident caused by control rod ejection). The purposes of researches - study of mechanisms and processes influencing on thermal mechanical behaviour and high burnup (50-60 MWd/kgU) fuel rod failure, verification of the RAPTA-5 code. The test methodology and results of a priory accounts for the experimental assembly containing one unirradiated fuel rod and two refabricated fuel rods (burnup is about 50 MWd/kgU) are submitted. (author)

  6. Nuclear fuel rod straightness measuring system and method

    International Nuclear Information System (INIS)

    A method is described for measuring the straightness of a rod, comprising the following steps: (a) supporting the rod so that if the rod were straight, the rod would remain straight without transverse translational movement while supported and if rotated about its longitudinal axis, and so that if the rod were cambered, the rod would remain so cambered while supported and if rotated; (b) rotating the supported the rod so that if the rod were straight, the rod would be rotated about its longitudinal axis; (c) measuring the distances during the rotation between the supported and rotating the rod and rigidly-mounted, spaced-apart range finders, the range finders disposed apart from and directed towards the supported and rotating the rod and disposed so that if the rod were straight, the range finders would each be directed transverse to the longitudinal axis; and (d) calculating for each of the range finders the difference between the maximum and minimum of the distance measurements, the differences indicating the degree of straightness of the rod

  7. Modeling and simulation performance of sucker rod beam pump

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-30

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

  8. Modeling and simulation performance of sucker rod beam pump

    International Nuclear Information System (INIS)

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

  9. Modeling and simulation performance of sucker rod beam pump

    Science.gov (United States)

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

    2015-09-01

    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.

  10. Piston rod seal for a Stirling engine

    Energy Technology Data Exchange (ETDEWEB)

    Shapiro, W.

    1984-01-31

    In a piston rod seal for a Stirling engine, a hydrostatic bearing and differential pressure regulating valve are utilized to provide for a low pressure differential across a rubbing seal between the hydrogen and oil so as to reduce wear on the seal. 3 figs.

  11. Piston rod seal for a Stirling engine

    Energy Technology Data Exchange (ETDEWEB)

    Shapiro, Wilbur (Schenectady, NY)

    1984-01-01

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

  12. Technological improvements in wire rod mills

    Energy Technology Data Exchange (ETDEWEB)

    Lestani, M.

    1996-07-01

    The paper deals with the latest rolling technologies and hi-tech equipment developed by Danieli-mogardshammar to ensure top performance of modern wire rod mills. In particular, a high reduction sizing mill, a twin module fast finishing block and a high speed cropping shear are presented. (authors)

  13. Analysis of sucker rod and sinkerbar failures

    Energy Technology Data Exchange (ETDEWEB)

    Waggoner, J.R.; Buchheit, R.G.

    1993-03-01

    This report presents results of a study of performance and failures of the sucker rod/sinkerbar string used in beam-pumping operations through metallography, finite element analysis, and failure data collection. Metallography showed that the microstructure of the steel bar stock needs to be considered to improve the fatigue resistance of the sucker rod strings. The current specification based on tensile strength, or yield strength, may not be appropriate since failure occurs because of fatigue and not yielding, and tensile strength is not always a good measure of fatigue resistance. Finite element analysis of the threaded connection quantitatively assesses the coupling designs under various loading conditions. Subcritical fractures in metallography are also suggested by calculated stress distribution in threaded coupling. Failure data illustrates both magnitude and frequency of failures, as well as categorizing the suspected cause of failure. Application of the results in each of these project areas is expected to yield improved choice of metal bar stock, thread design, and make-up practices which can significantly reduce the frequency of sucker rod failures. Sucker rod failures today are not inherent in the process, but can be minimized through the application of new technology and observation of common-sense practices.

  14. Program optimizes sucker-rod pumping mode

    International Nuclear Information System (INIS)

    Direct energy costs for sucker-rod pumping can be optimized by selecting the right pump size, stroke length, and pumping speed for the required liquid production rate. Calculation procedures for a computer program are developed for optimizing the design of conventional pumping units

  15. Wear simulation of sucker rod couplings

    Energy Technology Data Exchange (ETDEWEB)

    Schumacher, W.J. (Armco, Inc., Middletown, OH (United States))

    1991-09-01

    This paper reports that sucker rod strings are devices used to actuate pumps located at the bottom of oil wells. The individual rods are connected together by threaded couplings. Since the couplings have a larger diameter than the rods, they sometimes contact the inside diameter of the tubing during the up and down pumping cycle. Usually, this is not problem unless buckling occurs in the downstroke; however, this can lead to accelerated wear of the coupling and tubing. In nonvertical wells (offset, deviated, or slanted), the contact is more severe and rapid wear takes place. Couplings are more easily replaced during shutdowns; it is very important to minimize wear to tubing since it is virtually impossible to replace. TRIBONIC 20, an iron-based alloy containing approximately 13% Mn, 5% Si, 5.5% Cr, and 5% Ni, was laboratory evaluated to determine whether or not it could solve the sucker rod coupling-production tubing wear problem. The alloy demonstrated outstanding wear resistance both to itself and in protecting type 1019 steel.

  16. Program optimizes sucker-rod pumping mode

    Energy Technology Data Exchange (ETDEWEB)

    Takacs, G. (Technical Univ. of Miskolc, Miskolc (HU))

    1990-10-01

    Direct energy costs for sucker-rod pumping can be optimized by selecting the right pump size, stroke length, and pumping speed for the required liquid production rate. Calculation procedures for a computer program are developed for optimizing the design of conventional pumping units.

  17. Sucker rod scraper method and device

    Energy Technology Data Exchange (ETDEWEB)

    Hickman, A.E.

    1988-05-03

    A plurality of sucker rod scrapers are securely attached in space apart relationship to a length of sucker rod for scraping paraffin deposits from the interior of a tubing string. This is done to vitiate obstruction to the flow of oil by maintaining a satisfactory effective flow area within the tubing string. The scrapers each have a spiraled scraping surface wound helically about the sucker rod and attached at each opposed end by the provision of a clamping member which includes a U-band and a heat shield. The U-band has confronting marginal edge portions which overlap the edge portions of the heat shield. The heat shield has a slot formed centrally therein for receiving a tab located at either of the opposed ends of the spiraled scraper. The confronting edges of the U-band are welded to one of the marginal opposed ends of the scraper, while the heat shield prevents the metal of the sucker rod being elevated to a temperature which changes its characteristics.

  18. Optimized design of the nuclear fuel rod transport container used for non-destructive testing with neutron radiography

    International Nuclear Information System (INIS)

    Background: Working under extreme conditions, nuclear fuel rods, the key component of nuclear plants and reactors, are easy to be broken. In order to be safe in operation, lots of testing methods on the fuel rods have to be carried out from fabrication to operation. Purpose: Neutron radiography is a unique non-destructive testing technique which can be used to test samples with radioactivity. As the essential equipment, the nuclear fuel rod transport container has to shield the radioactivity of fuel rod and control its movement during testing and transporting. Methods: The shielding simulation of the transport container was performed using the MCNP4C code, which is a general purpose Monte Carlo code for calculating the time dependent multi-group energy transport equation for neutrons, photons and electrons in three dimensional geometries. Results: The material and dimension of the transport container used for neutron radiography testing fuel rods at Chinese Advanced Research Reactor (CARR) were optimally designed by MCNP, and the mechanical devices used to control fuel rods' movement were also described. Conclusion: The 2-m long fuel rod can be tested at CARR's neutron radiography facility (under construction) with this transport container. (authors)

  19. Electrostatics and depletion determine competition between 2D nematic and 3D bundled phases of rod-like DNA nanotubes.

    Science.gov (United States)

    Park, Chang-Young; Fygenson, Deborah K; Saleh, Omar A

    2016-06-21

    Rod-like particles form solutions of technological and biological importance. In particular, biofilaments such as actin and microtubules are known to form a variety of phases, both in vivo and in vitro, whose appearance can be controlled by depletion, confinement, and electrostatic interactions. Here, we utilize DNA nanotubes to undertake a comprehensive study of the effects of those interactions on two particular rod-like phases: a 2D nematic phase consisting of aligned rods pressed against a glass surface, and a 3D bundled network phase. We experimentally measure the stability of these two phases over a range of depletant concentrations and ionic strengths, finding that the 2D phase is slightly more stable than the 3D phase. We formulate a quantitative model of phase stability based on consideration of pairwise rod-rod and rod-surface interactions; notably, we include a careful accounting of solution electrostatics interactions using an effective-charge strategy. The model is relatively simple and contains no free parameters, yet predicts phase boundaries in good agreement with the experiment. Our results indicate that electrostatic interactions, rather than depletion, are largely responsible for the enhanced stability of the 2D phase. This work provides insight into the polymorphism of rod-like solutions, indicating why certain phases appear, and providing a means (and a predictive model) for controlling those phases. PMID:27126684

  20. A study on the nuclear characteristics of enriched gadolinia burnable absorber rods; the first year (2000) report

    International Nuclear Information System (INIS)

    An analysis model using MICBURN-3/CASMO-3 is established for the enriched gadolinia burnable absorber rods. A homogenized cross section editing code, PROLOG, is modified so that it can handle such a fuel assembly that includes two different types of gadolinia rods. Study shows that Gd-155 and Gd-157 are almost same in suppressing the excess reactivity and it is recommended to enrich both odd number isotopes, Gd-155 and Gd-157. It is estimated that the cycle length increases by 2 days if enriched gadolinia rods are used in the commercial nuclear power plant such as YGN-3 of which the cycle length is assumed 2 years. For the advanced integral reactor SMART in which ultra long cycle length and soluble boron-free operation concept is applied, natural gadolinia burnable absorber rods fail to control the excess reactivity. On the other hand, enriched gadolinia rods are successful in controling the excess reactivity. To minimize power peakings, various placements of gadolinia rods are tested. Also initial reactivity holddown and gadolinia burnout time are parametrized with respect to the number of gadolinia rods and gadolinia weight fractions

  1. Validation Test of CARR Safety Rod Driving Mechanism

    Institute of Scientific and Technical Information of China (English)

    2008-01-01

    <正>CARR safety Rods are driven by hydraulic force. The safety rod driving mechanism is designed by Tsinghua University and manufactured by Shenyang LIMING factory. Two sets of the mechanism are used for the validation test.

  2. Improved designs reduce sucker-rod pumping costs

    Energy Technology Data Exchange (ETDEWEB)

    Takacs, G. [Univ. of Miskolc (Hungary)

    1996-10-07

    Pumping mode selection, optimum counterbalance determination, and rod string design are factors that can reduce operational costs and improve sucker-rod pumping operations. To maximize profits from sucker-rod pumped wells, designs must aim at technically and economically optimum conditions. Assessment of surface and downhole energy losses are basic considerations for improving system efficiency. It is important to properly select the pumping mode, such as the combination of plunger size, pumping speed, stroke length, and rod taper design. The best pumping mode maximizes lifting efficiency and, at the same time, reduces prime-mover power requirements and electrical costs. Surface equipment operational efficiency can be improved with optimum counterbalancing of the pumping unit, and top achieve an ideal sucker-rod pumping system, a tapered rod string must have a proper mechanical design. The paper discusses rod pumping, downhole energy losses, surface losses, optimum efficiency, mode selection, counterbalancing, minimizing the cyclic load factor, and rod string design.

  3. Longitudinal Vibrations of Rheological Rod With Variable Cross Section

    Institute of Scientific and Technical Information of China (English)

    Katica(Stevanovic)HEDRIH; AleksandarFILIPOVSKI

    1999-01-01

    Longitudinal vibrations of rheological rod with variable cross section are examined.Particular solutions and eigenfunction are accomplished for natural vibrations of the rod with hereditary material of standard hereditary body.Some examples are given.

  4. Composite models for combined rod and fluid dynamics in sucker-rod pumping well systems

    Energy Technology Data Exchange (ETDEWEB)

    Lekia, S.D.L.

    1989-01-01

    This study presents the derivation and the numerical solution of composite models in which both the rod string and the fluid dynamics are coupled so as to accurately account for the effects of viscous friction in sucker-rod pumped wells. A viscous damped hyperbolic first order partial differential equation is coupled to the time derivative of Hooke's law to model the rod string motion and Navier Stokes equations are used to model the fluid dynamics in the rod-tubing annulus. A set of four equations comprise the composite model from which four sub-models for different flow scenarios are considered. The equations are solved numerically by a shock capturing algorithm known as the MacCormack Explicit Scheme which is a two-step predictor-corrector scheme and is second order accuracy in time and space. Five example problems covering various pump setting depths, fluid properties and surface pumping unit kinematics are presented to study the effects of certain important variables. From the analyses of the results of these example problems it is concluded that (1) while the effects of fluid dynamics may appear masked in shallow to medium depth sucker-rod pumped wells, they can not be ignored in deeper wells where large discrepancies occur in the prediction of system parameters, (2) the load range decreases moderately as viscosity increases and the predicted polished rod horsepower does not change significantly over the range of viscosities studied in shallow to medium depth sucker-rod pumped wells, (3) the presence of small quantities of the gas phase in the fluid column reduces system peak torque and precipitate the need for smaller counterbalance weights and (4) the influence of two-phase gas-liquid flow in the rod-tubing annulus on system design parameters declines with increasing pump setting depth. The results are compared against other design models appearing in the literature.

  5. Fluorescent colloidal silica rods - synthesis and phase behavior

    OpenAIRE

    Kuijk, A.

    2012-01-01

    Although the experimental study of spherical colloids has been extensive, similar studies on rod-like particles are rare because suitable model systems are scarce. To fulfill this need, we present the synthesis of monodisperse rod-like silica colloids with tunable dimensions. Rods were produced with diameters of 200 nm and larger and lengths up to 10 µm, which resulted in aspect ratios ranging from 1 to 25. The growth mechanism of these rods involves emulsion droplets of water in pentanol, in...

  6. Distributed Mode Filtering Rod Fiber Amplifier With Improved Mode Stability

    DEFF Research Database (Denmark)

    Laurila, Marko; Alkeskjold, Thomas Tanggaard; Broeng, Jes;

    2012-01-01

    We report 216W of average output power from a photonic crystal rod fiber amplifier. We demonstrate 44% power improvement before onset of the mode instability by operating the rod fiber in a leaky guiding regime.......We report 216W of average output power from a photonic crystal rod fiber amplifier. We demonstrate 44% power improvement before onset of the mode instability by operating the rod fiber in a leaky guiding regime....

  7. Sucker rod string design of the pumping systems

    Directory of Open Access Journals (Sweden)

    Chun Hua Liu

    2015-08-01

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

  8. Improvement to the pattern of control rods of the equilibrium cycle of 18 months for the CLV using bio-inspired algorithms; Mejora del patron de barras de control del ciclo de equilibrio de 18 meses para la CLV empleando algoritmos bio-inspirados

    Energy Technology Data Exchange (ETDEWEB)

    Perusquia, R.; Ortiz, J.J.; Montes, J.L. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)]. e-mail: rpc@nuclear.inin.mx

    2003-07-01

    Nowadays in the National Institute of Nuclear Research are carried out studies with some bio-inspired optimization techniques to improve the performance of the fuel cycles of the boiling water reactors of the Laguna Verde power plant (CLV). In the present work two bio-inspired techniques were applied with the purpose of improving the performance of a balance cycle of 18 months developed for the CLV: genetic algorithms (AG) and systems based on ants colonies (SCH). The design of the reference cycle it represents in several aspects an optimal cycle proposed starting from the experience of several operation decades with the boiling water reactors (BWR initials for Boiling Water Reactor) in the world. To try to improve their performance is beforehand a difficult challenge and it puts on test the feasibility of the optimization methods in the reloads design. The study of the bio-inspired techniques was centered exclusively on the obtaining of the control rod patterns (PBC) trying to overcome the capacity factor reached in the design of the reference cycle. It was fixed the cycle length such that the decrease of the coast down period would represent an increase of the capacity factor of the cycle; so that, it diminishes the annual cost associated with the capital cost of the plant. As consequence of the study, was found that the algorithm based on the ants colonies reaches to diminish the coast down period in five and half days respect to the original balance cycle, what represents an annual saving of $US 74,000. Since the original cycle was optimized, the above-mentioned, shows the ability of the SCH for the optimization of the cycle design. With the AG it was reach to approach to the original balance cycle with a coast down period greater in seven days estimating an annual penalization of $US 130,000. (Author)

  9. Shield mining frame piston rod. Schildausbaugestell-Kolbenstange

    Energy Technology Data Exchange (ETDEWEB)

    Schuett, F.

    1981-05-02

    A piston rod for a shield mining frame for coal mining is described. This has radial outward connecting openings at the free end for hydraulic pipes. The plug-in connections are pushed in here and held with clamps. The piston rod part, in which these openings are situated, is made as a bar. The piston rod and bar form one part.

  10. Measurement of the Speed of Sound in a Metal Rod.

    Science.gov (United States)

    Mak, Se-yuen; Ng, Yee-kong; Wu, Kam-wah

    2000-01-01

    Suggests two improved methods to measure the speed of sound in a metal rod. One employs a fast timer to measure the time required for a compression pulse to travel along the rod from end to end, and a second uses a microphone to measure the frequency of the fundamental mode of a freely suspending singing rod. (Author/ASK)

  11. ROD INTERNAL PRESSURE QUANTIFICATION AND DISTRIBUTION ANALYSIS USING FRAPCON

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-01-01

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

  12. Carbon Inverse Opal Rods for Nonenzymatic Cholesterol Detection.

    Science.gov (United States)

    Zhong, Qifeng; Xie, Zhuoying; Ding, Haibo; Zhu, Cun; Yang, Zixue; Gu, Zhongze

    2015-11-18

    Carbon inverse opal rods made from silica photonic crystal rods are used for nonenzymatic cholesterol sensing. The characteristic reflection peak originating from the physical periodic structure works as sensing signals for quantitatively estimating cholesterol concentrations. Carbon inverse opal rods work both in cholesterol standard solutions and human serum. They are suitable for practical use in clinical diagnose.

  13. Fluorescent colloidal silica rods - synthesis and phase behavior

    NARCIS (Netherlands)

    Kuijk, A.

    2012-01-01

    Although the experimental study of spherical colloids has been extensive, similar studies on rod-like particles are rare because suitable model systems are scarce. To fulfill this need, we present the synthesis of monodisperse rod-like silica colloids with tunable dimensions. Rods were produced with

  14. Sealing system for piston rod of hot gas engine

    Energy Technology Data Exchange (ETDEWEB)

    Lundholm, S.G.; Ringqvist, S.A.

    1980-11-25

    An improvement to a sealing system for restricting fluid flow around a piston rod between a piston cylinder and crankshaft space in a hot gas engine where a seal element is secured around the piston rod in an intermediate chamber, the improvement including a link in the crankshaft space connecting, and permitting relative radial motion between, the piston rod and the crosshead and an o-ring having a diameter substantially greater than that of the piston rod and being secured between a lower ring securing the seal element in place around the piston rod and a wall of the intermediate chamber for frictionally restricting radial movement of the lower ring.

  15. STUDY ON A HYDROPHOBIC-HYDROPHILIC GRADIENT ROD

    Institute of Scientific and Technical Information of China (English)

    Jun Ma; Bai-yu Li; Hai-yun Liu; Zhi-min Zheng; Jian Xu

    2004-01-01

    A hydrophobic-hydrophilic gradient rod with a length of 40 mm and a diameter of 3 mm was prepared by heating a polymethylsilsesquioxane rod in a cylindrical stove with temperature gradient. The rod was thus pyrolyzed under a temperature gradient condition. The organic end of the gradient rod appears hydrophobic with a contact angle of 109.9° while the other end is hydrophilic with a contact angle of 62.4°. The gradient chemical structure and the gradient microstructure along the rod were characterized by FTIR and SEM, respectively.

  16. Test Research on Special Sucker Rod for Screw Pump

    Institute of Scientific and Technical Information of China (English)

    Zhang Mingyi; Chen Mingzhan; Li Zhi

    2006-01-01

    @@ According to the statistics of straight thread sucker rods' application in screw pump in Daqing Oilfield before2000, the proportion of sucker rods' yearly breakaway reached to 41.6%, taking up 70% of the total wells that were checked. Thus it can be seen that the rods breakaway problem was becoming the main barrier restricting screw pump large-scale population and application. Since then,the development work on the special sucker rods for screw pump had been carried on. Through the analysis on the failure position and failure form of the sucker rods',the following conclusions arepresented:

  17. Whole-rod testing of intact and defective LWR rods under expected dry-storage conditions

    International Nuclear Information System (INIS)

    The objective of this project is to provide the Nuclear Regulatory Commission with information to confirm or establish spent fuel dry storage licensing positions relative to: (1) the long-term, low-temperature (less than 2500C) behavior of spent fuel rods in dry storage; and (2) the radioactive contamination potential of crud from cladding in dry storage. The basic need for this data is to: confirm long-term, low temperature (less than 2500C) spent fuel dry storage performance predictions based on theoretical analyses and on results from high-temperature, short-term laboratory tests; determine the nature and behavior of crud layers as a function of dry storage time; and determine the potential radioactive crud contamination (e.g., spalling characteristics) for dry storage. An eight-rod test matrix of PWR and BWR rods was chosen which consisted of all combinations of intact or breached cladding in an oxidizing or inert atmosphere. The PWR rods (30.5 GWD/MTU) were discharged from H.B. Robinson in May 1974, and the BWR rods (12.9 GWD/MTU) were discharged from Peach Bottom in March 1976. The eight test rods were visually inspected for crud and defects with the results recorded on video tape. Cladding penetration was confirmed. All the rods were put in test capsules with the appropriate atmosphere and leak checked. The test capsules were loaded into a test train and the train was placed in the furnace cavity. The test was started on September 15, 1982 and is presently at 2300C. After the first 10-month run is completed, an interim examination, consisting of visual inspection, gamma scanning, and crud sampling, will be conducted

  18. Photonic mesophases from cut rod rotators

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-01-14

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

  19. Oligo(naphthylene–ethynylene) Molecular Rods

    DEFF Research Database (Denmark)

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

    2013-01-01

    Molecular rods designed for surface chirality studies have been synthesized in high yields. The molecules are composed of oligo(naphthylene–ethynylene) skeletons and functionalized at their two termini with carboxylic acids and hydrophobic groups. The molecular skeletons were constructed by means...... of palladium-catalyzed Sonogashira reactions between naphthyl halides and acetylenes. The triazene functionality was used as a protected iodine precursor to allow linear extension of the molecular rods during the synthe-ses. The carboxylic acid groups in the target molecules were protected as esters during...... the synthesis to keep the large aromatic molecules soluble during their syntheses. These rigid oligomers were designed to form lamella-like structures when adsorbed on a surface, through which multiple distinguishable surface conformations should be obtainable. Preliminary scanning tunneling microscopy imaging...

  20. Absorber rod drive for nuclear reactors

    International Nuclear Information System (INIS)

    The invention concerns an absorber rod drive for Boiling Water Reactors, in which a mechanical drive is combined with a hydraulic drive working separately from it, so that both drives are situated concentric within an overall length. The driving torque of a motor is transmitted to a threaded spindle, which moves a free adjacent hollow piston vertically via a fixed nut. The same means are used for the hydraulic liquid which is used as coolant or moderator and there are nozzles, annular gaps and/or bores between the hydraulic system and the reactor pressure vessel for the purpose of pressure compensation. All the components of the absorber rod drive except the sealing housing and the setting drive are situated in one casing tube taking the differential pressure. (orig./HP)