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Sample records for agincd control rod

  1. Control rod

    International Nuclear Information System (INIS)

    In a control rod of a nuclear reactor, B-10 is distributed such that the concentration of B-10 in boron carbide is lowered from the outer side to the inner side of the control rod. Alternatively, the inside of a blade is radially divided into a plurality of regions, and the amount of boron carbide loaded in the regions is reduced from the outer side to the inner side of the control rod. Alternatively, a plurality of sintered products of boron carbide are disposed radially in the blade, and the sintered product is divided to a first region where the B-10 content is relatively low and a second region having a higher B-10 content than the first region, and the sintered product disposed on the inner side is constituted such that the position of the first region in the sintered product is localized to the inner side of the control rod. Then, improvement of reliability and reduction of cost can be attained while maintaining an effective neutron absorbing performance of control rods taking neutron flux distribution into consideration. (N.H.)

  2. 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.

  3. 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.)

  4. Control rod operation device

    International Nuclear Information System (INIS)

    Purpose: To reduce operator's operation burdens in the low power state, while moderate his mental burdens upon high power state for the operation of control rod operation device. Constitution: Coordinate of main control rods to be operated, aimed insertion and withdrawal positions and velocity are calculated by the control rod operation sequence and the control rod worth table, to output a control rod selection signal and a control rod operation signal. The control rod operation device conducts extraction or insertion of control rods by these signals. In this way, the operator can automatically operate the control rods by merely manipulating the control rod operation device, by which the operator's operation burden can be reduced in low power state. Further, since the selection of the control rods, the operation speed, etc. are judged by an electronic computer also upon high reactor power state, operator's metal burdens can be moderated. (Kamimura, M.)

  5. Visual inspections of the neutron absorber control rods of the IEA-R1 reactor

    International Nuclear Information System (INIS)

    The Fuel Engineering Division at IPEN/CNEN-SP developed facilities for visual inspection of the IEA-R1 fuel elements and neutron absorbing control rod assemblies inside the research reactor pool. This work presents the method of visual inspection performed at IEA-R1 research reactor. These inspections were adopted to evaluate and to follow the state of the Ag-In-Cd control assemblies fabricated at CERCA in 1972 that remain in use at the reactor core. In 1998, 2000 and 20001, visual inspections were performed in these control rod assemblies, which the general conditions were evaluated. (author)

  6. Control rod drives

    International Nuclear Information System (INIS)

    Purpose: To prevent damages in control rod drives upon connection and disconnection with control rods by providing, to an extension rod, a closure and opening guide mechanism which is adapted to open and close depending upon connection and disconnection with the control rod. Constitution: In control rod drives having a driving section and a lower mechanism, the lower mechanism has a guide tube engaged into the upper cover of a reactor container and suspended therefrom into the reactor container. An opening and closure mechanism with guide blades capable of contacting the inner wall of the guide tube is secured to an extension rod for fastening the gripper which supports the control rod. Such a mechanism cap prevent damages in the control rod drives and the control rods due to the connection after disconnection of them by the buffering action between the extension rod and the control rod, as well as damages in both of them caused by rolling such as in earthquakes by the buffering action between the extension rod and the guide tube. (Moriyama, K.)

  7. 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.)

  8. Control rod assembly

    International Nuclear Information System (INIS)

    In a control rod assembly comprising an extension rod extended upwardly from the upper end of a control rod main body disposed in a reactor core and an extension tube engaging a grip portion disposed to an upper portion of the extension rod for suspending the control rod main body, a shrinkable portion is disposed to a part of the extension tube or extension rod, or a grip portion shrinkable in the axial direction is disposed to the extension rod. Further, a spring is interposed to a portion of the extension tube and bellows are disposed to the inner side or the outer side of the spring. A double-cylindrical temperature sensing member is disposed surrounding the outer side of the bellows or the spring. Liquid metals are sealed in the temperature sensing member or the bellows. This can improve the response of the coolants to the temperature elevation and can suppress the change of the reactor core insertion amount relative to temperature change during usual operation. (T.M.)

  9. Control rod drives

    International Nuclear Information System (INIS)

    Purpose: To fix a magnetic rotor to a drive shaft and at the time of non-driving, to restrain the rotor by permanent magnets thereby to hold the position of the control rod safely and accurately. Constitution: A control rod position holding device is provided in a motor or a drive shaft of a control rod drive. This device consists of a rotor and a stator, the rotor being provided on its circumference with salient-poles arranged equidistantly, and the position of the rotor being determined depending upon the transfer distance of the control rod and the conversion ratio of the converter. On the other hand, the stator has salient-poles (any of them is a permanent magnet) having the number of poles and the positional relationship equivalent to those of the rotor, and provided in the inner periphery of a cylinder using the drive shaft as a central shaft and wound with a winding. When the control rod is not driven, the poles of the rotor are attracted by the magnetic force of the confronting poles of the stator, thereby to prevent the inverse rotation of the motor shaft due to the dead weight of the control rod. When a current is caused to flow through the winding, the magnetic force of the permanent magnet, and the stator release the rotor. (Yoshino, Y.)

  10. Control rod blocking device

    International Nuclear Information System (INIS)

    Purpose: To increase the degree of freedom for the reactor operation by control rod blocking by monitoring the critical power ratio (CPR) with real time. Constitution: There has been a problem that the withdrawal of control rods may occasionally be inhibited with all the margin in view of CPR. The present invention dissolves this problem. That is, the control rod withdrawal device periodically calculates CPR, and calculated CPR upon generation of a control rod withdrawing signal by conpensating the result of calculation with a LPRM signal and a reactor core flow rate signal. The CPR at real time is compared with a predetermined setting value to output a control rod withdrawing inhibition signal depending on the result of the comparison. In the device as described above, since CPR is monitored at real time, the control rod can be withdrawn without causing fuel damages, as well as the inhibition of withdrawal irrespective of the presence of margin in view of CPR can be avoided. Accordingly, degree of freedom in the reactor operation can be increased. (Kamimura, M.)

  11. 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.)

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

    International Nuclear Information System (INIS)

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

  13. Control rod drive

    International Nuclear Information System (INIS)

    A control device for long time stopping is disposed for stopping by applying the same number of normal rotation and reverse rotation to a ball spindle at a predetermined time interval. Even in a case where a control rod is not operated for a long period of time, sticking between a sealing material and a ball spindle is prevented, rotational torque is not increased excessively, and the control rod can always be operated normally. Further, a stopping control device is disposed for applying rotation after the stop of the rotation of the ball spindle in the reverse direction within one turn. Lubricants and obstacles are introduced between the surfaces purified by the rotation, to prevent the direct contact of the purified surfaces to each other and correct the deformation of sealing members. Therefore, the rotational torque is not increased excessively. (N.H.)

  14. Control rod position detector

    International Nuclear Information System (INIS)

    The device of the present invention can save blowers for compulsory cooling. That is, the control rod position detector comprises (1) a control rod driving shaft made of a ferromagnetic material moving in a pressure vessel of a nuclear reactor and (2) detector coils arranged to the outside of the pressure vessel each at an identical distance over the moving stroke of the driving shaft for detecting the position of the driving shaft by the change of inductance. In addition, heat insulation materials are disposed between the detector coils and the reactor pressure vessel. Then, heat from the reactor pressure vessel can be insulated. Accordingly, temperature of the detector coils can be reduced by natural cooling. As a result, since it is no more necessary to dispose compulsory cooling fans as required in a conventional case, the entire device can be constituted economically, and the reliability of the device is improved. (I.S.)

  15. Control rod drive mechanisms

    International Nuclear Information System (INIS)

    Purpose: To accurately measure the loads generated upon scram and judge the absence or presence of deceleration in control rod drive mechanisms. Constitution: Control rod drive mechanisms for use in a BWR type reactor includes an index tube vertically movably, connected at the upper end to the control rod and having a drive piston at the lower end. A piezoelectric member for detecting the load generated upon uprise of the index tube is disposed and signals from the piezoelectric member is connected to a calculation processing device. A load exerted when the index tube uprises is measured by way of the piezoelectric member upon scram thereby judging the absence or presence of the decelerating operation. Therefore, the nuclear reactor can be shutdown only when it is required with no excess safety operation than required. As a result, the reactor availability can be improved and, in addition, it is also possible to mitigate the burden of in-service inspection and reduce the operators' exposure. (Kamimura, M.)

  16. 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.)

  17. 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.)

  18. Nuclear reactor with control rods

    International Nuclear Information System (INIS)

    The invention relates to liquid cooled nuclear reactors. In particular, it concerns reactors with mobile control rods in a straight line and guide tubes to guide these control rods through the internal upper components of the reactor vessel and in the aligned fuel assemblies of the core

  19. Control rod drives

    International Nuclear Information System (INIS)

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

  20. 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.)

  1. 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

  2. Criticality calculation and control rods for the Westinghouse Reactor Evaluation Center facility

    International Nuclear Information System (INIS)

    This work evaluates two clean critical cores by WIMS-TRACA/CITATION codes calculation, 4 energy groups and bi dimension geometry. The first core is composed of U O2 with a clad of stainless steel and 20 absorbers Ag-In-Cd absorbers rods, the second is composed of U O2 with a clad of Zircaloy and 12 B4 C absorbers rods. (author)

  3. 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

  4. Control rod calibration including the rod coupling effect

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  6. Advanced gray rod control assembly

    Science.gov (United States)

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

    2013-09-17

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

  7. Digitization on control rod position indication system

    International Nuclear Information System (INIS)

    This paper introduces the design mechanism, system structure of Control Rod Position Indication System and the application of Programmable Logic Controller (PLC) on Control Rod Position Indication System. (authors)

  8. Flaw detecting device for control rod cluster

    International Nuclear Information System (INIS)

    The device of the present invention measures the reduction of a wall thickness of each cladding tube of control rod caused by abrasion or flaws at a high accuracy in a short period of time in a control rod cluster for a PWR type reactor. Namely, a control rod cluster for a PWR type reactor is formed by bundling a plurality of control rods at one end. The reduction of the wall thickness of the control rod cladding tubes is measured by a flaw detecting method using eddy current. A group of flaw detecting probes is moved in the axial direction relative to the control rod cluster to scan all the control rods in the axial direction of the cladding tubes. The group of the flaw detecting probes has circular flaw detecting probes at least by the total number of the control rods corresponding to the position of each of the arranged control rods of the control rod cluster. Each of the circular flaw detecting probes has a measuring hole through which one control rod can pass. Accordingly, all of the control rods are scanned from one end to the other end for the control rod cluster thereby capable of measuring the entire surface simultaneously. (I.S.)

  9. LWR control assembly designs: A historical perspective

    International Nuclear Information System (INIS)

    Control rod designs and materials have evolved in response to performance problems in both PWRs and BWRs. Irradiation-assisted stress corrosion cracking (IASCC) due to absorber swelling has primarily affected BWR control rods with B4C absorbers, but has also occurred in PWRs with Ag-In-Cd absorbers. The primary problems for some designs of PWR control rods have been wear of the rodlets against upper internal components and swelling with tip wear and cracking. Competition amongst vendors for supplying control rod reloads has also resulted in design improvements. This paper provides an historical review of PWR and BWR control rod designs, their problems and remedies. (author)

  10. Snubber assembly for a control rod drive

    International Nuclear Information System (INIS)

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

  11. Fuel followed control rod installation at AFRRI

    International Nuclear Information System (INIS)

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

  12. Control rods in LMFBRs: a physics assessment

    International Nuclear Information System (INIS)

    This physics assessment is based on roughly 300 control rod worth measurements in ZPPR from 1972 to 1981. All ZPPR assemblies simulated mixed-oxide LMFBRs, representing sizes of 350, 700, and 900 MWe. Control rod worth measurements included single rods, various combinations of rods, and Ta and Eu rods. Additional measurements studied variations in B4C enrichment, rod interaction effects, variations in rod geometry, neutron streaming in sodium-filled channels, and axial worth profiles. Analyses were done with design-equivalent methods, using ENDF/B Version IV data. Some computations for the sensitivities to approximations in the methods have been included. Comparisons of these analyses with the experiments have allowed the status of control rod physics in the US to be clearly defined

  13. Control rods in LMFBRs: a physics assessment

    Energy Technology Data Exchange (ETDEWEB)

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

    1982-08-01

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

  14. Control rod guide tube assembly

    International Nuclear Information System (INIS)

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

  15. Control rod studies for enigma configurations

    International Nuclear Information System (INIS)

    Collaboration is underway between Argonne and the CEA-Cadarache on the preparation of experiments for the ENIGMA program dedicated to the reactor physics experiments supporting the development of gas-cooled fast reactors. Specifications have been defined for the study of control rods in the central void zone of ENIGMA configurations. Deterministic calculations of the rodded configurations have been performed using the ERANOS code system. The various core criticality states for the different phases of the control rod experiments have been determined by specifying the number of additional fuel assemblies required to restore criticality. Control rod worths, flux distributions, and reactions rate distributions for a few nuclides have been analyzed. The study revealed the significant impact of spatial heterogeneity in the rod configurations used for the experiments, indicating flexibility for the control rod experiments

  16. Control rod drive of nuclear reactor

    International Nuclear Information System (INIS)

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

  17. 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.

  18. 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

  19. TVO'S Experiences with Fuel and Control Rods

    International Nuclear Information System (INIS)

    The Finnish regulatory guides require that the nuclear power plants must maintain a supervision programme for the fuel and the control rods. The aim of the supervision is to evaluate whether the performance is in accordance with the design bases and the expectation. At each unit at least 8 fuel assemblies and fuel channels are annually inspected visually. Many dimensional measurements and oxide thickness measurements have also been performed. During the 1980's TVO reused fuel channels. The dimensions of the channels were measured before reuse, which means that TVO has measured the dimension of more than 1000 fuel channels. The rate of fission gas release from the fuel pellets to the free volume of the rods has been determined, based on the Kr-85 activity of the plenum gas, on 22 fuel assemblies. Eddy current inspection has been applied to find leakers. During shutdowns the high exposure control rods of control cells are rather exchanged than inspected and reinserted, the inspection of these control rods is made later. Some medium exposure control rods are selected for inspection during the shut down. Normally only visual inspection is performed. However, some rods exhibiting cracks in the visual inspection have been neutron radiographed. A summary of the inspection results is given in this paper. The future development foreseen for the design and operation strategy is also presented, concerning fuel assemblies, control rods, operating strategies and power up-rating

  20. 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

  1. French LMFBR's control rods experience and development

    International Nuclear Information System (INIS)

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

  2. 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

  3. Estimation of irradiated control rod worth

    Energy Technology Data Exchange (ETDEWEB)

    Varvayanni, M., 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); Antonopoulos-Domis, M. [School of Electrical and Computer Engineering, Aristotle University of Thessaloniki, Thessaloniki (Greece)

    2009-11-15

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

  4. 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

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

    International Nuclear Information System (INIS)

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

  6. Guiding system for a control rod

    International Nuclear Information System (INIS)

    This invention concerns a system for guiding and maintaining in a vertical position a control rod for a fast neutron nuclear reactor cooled by a liquid metal. These rods are fitted to slide in fixed sleeves, located at given places in the reactor core, and taking the shape of boxes with a regular polygonal straight cross section, open at their upper end and fitted with a positioning foot at their lower end. Efficient cooling of the absorbing part of the rod is assured through a proper flow of the liquid metal (sodium) cooling the core and flowing inside the control rod sleeve from the bottom upwards as from the bottom foot to the open top end. Most of this flow is forced to pass inside the rod itself and not between the rod and the sleeve in which it slides. The system, whilst ensuring that the rod slides in its sleeve without any risk of jamming, leaves the upper part of the sleeve free

  7. Control rod drives for HTGR type reactor

    International Nuclear Information System (INIS)

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

  8. Nuclear reactor remote disconnect control rod coupling indicator

    International Nuclear Information System (INIS)

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

  9. 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.

  10. 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

  11. Control rod driving device for nuclear reactor

    International Nuclear Information System (INIS)

    In a magnetic jack type control rod drives, superconducting wires having current limiting effect are connected to a line for supplying current to coils. Ceramics type superconducting material is used as the superconducting wires. If abnormality should occur in the control rod drives, superconducting wire reaches a critical temperature due to the temperature elevation of the coolants. Then, the superconducting wires are instantaneously changed from the superconducting material to insulating material which is the inherent nature of the ceramic material to rapidly interrupt the current flowing through the coils. Thus, scram can be conducted simply. (S.T.)

  12. 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)

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

    International Nuclear Information System (INIS)

    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)

  14. 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.)

  15. 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.

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

    International Nuclear Information System (INIS)

    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 10B 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% 10B. It was also found that NEOL results for different fuel types and different fuel enrichments have a variation of about 2-3% in 10B 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

  17. Replacement means for control rod drive mechanism

    International Nuclear Information System (INIS)

    Object: To permit assembling and removal operation of a control rod drive mechanism to be carried out speedily and properly irrespective of the degree of skill of the operating personnel. Structure: When removing a control rod drive mechanism (CRD) a service platform and a frame body are operated for bringing a CRD lift guide frame to a position below the CRD to be removed. Then, a CRD receptacle is placed at the lower end of the CRD, and water is drained from the CRD. Subsequently, a chain is driven by a drive means in a direction which lowers the receptacle, and only the CRD is lowered along the CRD lift guide frame. Thereafter, the CRD is secured at its upper portion by a support means, and the CRD lift guide frame is lowered by a lift jack to thereby permit revolution of the CRD, The CRD lift frame after revolution is lifted and then removed to the outside. (Kamimura, M.)

  18. The lifetime of the control rod drives

    International Nuclear Information System (INIS)

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

  19. Control rod drive mechanisms seismic analysis

    International Nuclear Information System (INIS)

    In the Taishan joint-design, in order to finish Control Rod Drive Mechanism (CRDM) seismic analysis, a response spectrum analysis in combination with quasi-static analysis and a time history analysis are performed according to the different design stages and inputs. The simulated and simplified model of nonlinear structure is studied in the basic design stage; the translation of seismic input data and the use of nonlinear elements are studied in the detailed design stage. (authors)

  20. 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)

  1. 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.)

  2. 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

  3. Safety analysis of control rod drive computers

    International Nuclear Information System (INIS)

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

  4. 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

  5. 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

  6. 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

  7. Abrasion measuring method for rod of control rod assembly of reactor

    International Nuclear Information System (INIS)

    The present invention provides a method of easily measuring abrasion caused on the outer surface of control rods of a control assembly to be used in a PWR type reactor. Namely, the control rod assembly comprise a plurality of control rods assembled in a cluster-like manner. Light is irradiated to a control rod to be measured from an optical measuring device for measuring the extent of abrasion on the surface of the control rods. The distance is measured by receiving the reflected light. The depth of abrasion is determined by comparing the thus measured distance to the abraded portion and the distance to an integral portion. Then, the depth of the abrasion is adjusted based on the control rod position and the angle to determine final depth of abrasion. The abrasion of control rods can be measured by remote control using one kind of light sensor. The device can be reduced in the size and the time for the measuring operation can also be shortened. (I.S.)

  8. 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

  9. 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.

  10. The integration of control rod calculation and VSOP

    International Nuclear Information System (INIS)

    Accurate calculation of the control rod outside the active zone in pebble bed HTR requires special treatment, integration of this detailed control rod calculation with whole core calculation like VSOP package requires more efforts, which is not realized before. Accurate calculation of control rod is proposed to use the discontinuity factor-corrected diffusion method accompanied with transport calculation in advance. Appropriate coupling between the control rod calculation and whole core calculation must be developed to take into account of the buckling feedback, spectrum change, data transform and so on. In this paper, internal spectrum calculation module inside VSOP is used to model the control rod region, and the neutron leakage obtained from the whole core calculation is used as the boundary condition of the control rod region. The numerical calculation demonstrated the feasibility and accuracy of this coupling and integration method. (author)

  11. Control rod for a pebble bed nuclear reactor

    International Nuclear Information System (INIS)

    In order to leave the density of the pebble bed unchanged when the control rod is driven in and out, the tip of the control rod is provided with moving parts in the form of conical hemispheres or spoons. These parts move close to the control rod when it is driven in and spread out due to the effect of gravity when it is driven out. This loosens the heap of pebble shaped operating elements again. (DG)

  12. 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

  13. Detection of defects in control rods by eddy current examination

    International Nuclear Information System (INIS)

    To detect the defects of control rods in a reactor, a standard specimen including external defects, internal defects, and through-hole defects is fabricated. The eddy current signals of these defects are stored, analyzed on a PC by a program developed to acquire data of eddy currents. The optimum frequency for detecting defects of the control rods is 200 kHz. The defect location, and defect shape of the cladding in the control rods are detected by analyzing the impedance phase of the eddy current. It is confirmed that the defects in hafnium of control rods can be detected. (author)

  14. Control Rods in high-Flux Swimming-Pool Reactors

    International Nuclear Information System (INIS)

    Control-rod problems in open swimming-pool high-flux and high specific power research reactors are examined in the light of the calibrations and experiments made during the construction of the SILOE reactor. Control-rod operating experience for this reactor at 13 MW is also described. 2. The following are considered in turn: (a) Reactivity balances and reactivity values for the different types of rod tested (cadmium, B4C , rare earths and combinations of these different elements). (b) Flux peaks set up in the core by the presence of the control rods, their incidence on the specific power, the fast fluxes that can be obtained and means of increasing them. (c ) The technological problems involved in constructing the rods. (d) In-pile cooling, vibration, deformation and scram-time problems. 3. In conclusion, current studies on control rods in open swimming-pool reactors operating in the 10 - 30 1W range are briefly summarized. (author)

  15. Electrical device for controlling control rods

    International Nuclear Information System (INIS)

    Purpose: To simplify the structure of a pulse generator and improve the controllability by the use of binary signals representing the generation-time interval of pulses as the designation signals for the number of pulse generation to be applied from a computing circuit to a pulse generator. Constitution: A driving speed signal from a computing circuit is inputted as a pulse-interval-input signal into a pulse generator. An oscillator outputs clock pulses and a counter counts up the clock pulses. A comparator compares the value in a register with the value on the counter and issues pulse-train signals on every coincidence between them. The pulse train signal is converted into pulses with a predetermined pulse width in a one-shot circuit and then outputted as a pulse-train output signal. (Kawakami, Y.)

  16. Digitization of the rod control and rod position systems at Qinshan NPP

    International Nuclear Information System (INIS)

    The rod control and rod position systems in No.1 reactor of Qinshan nuclear power plant have been remodeled into digitized systems. The digital systems were achieved by PLC logic control functions, and a human-machine interface was used for online monitoring of system status and parameter changes. Redundant configuration was adopted for PLC controller to improve the system flexibility and reliability. The original analog circuits were modified into double-hold circuits against failures. Redundant communication between the rod control and rod position system was used to reduce the numbers of field cable. A data collection cabinet was designed to sample the coil waveform data on the drive mechanism for real-time data acquisition and facilitate post-accident analysis. (authors)

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

    International Nuclear Information System (INIS)

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

  18. Optimized Control Rods of the BR2 Reactor

    International Nuclear Information System (INIS)

    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.

  19. 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.

  20. 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.)

  1. Guide-tube and control rod for nuclear reactor

    International Nuclear Information System (INIS)

    The inside of the nuclear reactor guide tubes is of square cross section. A control rod drives a sliding cage and control fingers. Said sliding cage carries a system of interconnected mobile spacers for guiding control fingers

  2. 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.

  3. 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)

  4. Control rod studies in small and medium sized fast reactors

    International Nuclear Information System (INIS)

    Control rods are the primary safety mechanism in the operation of fast reactors. Neutronic parameters associated with the control rods have to be evaluated precisely for studying the behaviour of the reactor under various operating conditions. Control rods are strong neutron absorbers discretely distributed in the reactor core. Accurate estimation of control rod parameters demand, in principle transport theory solutions in exact geometry. But computer codes for such evaluations usually consume exorbitantly large computer time and memory for even a single parameter evaluation. During the design of reactors, evaluation of these parameters will be required for many configurations of control rods. In this paper, the method used at Indira Gandhi Centre for Atomic Research for estimating the parameters associated with control rods is presented. Diffusion theory solutions were used for computations. A scheme using three dimensional geometry represented by triangular meshes and diffusion theory solutions in few energy groups for control rod parameter evaluation is presented. This scheme was employed in estimating the control rod parameters in a 500 Mw(e) fast reactor. Error due to group collapsing is estimated by comparing with 25 group calculations in three dimensions for typical cases. (author). 5 refs, 4 figs, 3 tabs

  5. 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.

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

    International Nuclear Information System (INIS)

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

  7. BWR control rod design using tabu search

    International Nuclear Information System (INIS)

    An optimization system to get control rod patterns (CRP) has been generated. This system is based on the tabu search technique (TS) and the control cell core heuristic rules. The system uses the 3-D simulator code CM-PRESTO and it has as objective function to get a specific axial power profile while satisfying the operational and safety thermal limits. The CRP design system is tested on a fixed fuel loading pattern (LP) to yield a feasible CRP that removes the thermal margin and satisfies the power constraints. Its performance in facilitating a power operation for two different axial power profiles is also demonstrated. Our CRP system is combined with a previous LP optimization system also based on the TS to solve the combined LP-CRP optimization problem. Effectiveness of the combined system is shown, by analyzing an actual BWR operating cycle. The results presented clearly indicate the successful implementation of the combined LP-CRP system and it demonstrates its optimization features

  8. Operation of Control Rod Driving Mechanism controller at HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Gyu, Doo Seung; Woo, Lee Min; San, Choe Yeong; Kyoo, Kim Hyung [KAERI, Daejeon (Korea, Republic of)

    2012-10-15

    HANARO (High flux Advanced Neutron Application Reactor) achieved its first critical operation in 1995. Recently, there has been fast developments in the field of electronics. Many manufacturers of I and C components have disappeared or merged with the other companies. The suppliers of the control systems of the CRDM (Control Rod Driving Mechanism) at HANARO have disappeared. Therefore, we needed to change the control system of the CRDM since we cannot be provided with maintenance any longer. In this paper, we investigated the operation of the control system of the CRDM when the controller and motor driver are changed.

  9. ETRR-2 control rod withdrawal accident

    International Nuclear Information System (INIS)

    A safety evaluation of Egypt new reactor, Egypt Test and Research Reactor number 2 (ETRR-2), has been completed successfully. Intensive efforts have been made for design basis accidents (DBA) analysis. The present work presents analysis of one of these accidents, i.e. the uncontrolled reactivity insertion accident (RIA) due to erroneous withdrawal of control rod (CR) during normal operating conditions. The reactivity insertion may be fast or slow, depending on the speed of CR withdrawal. The availability of the scam system is considered. The reactivity insertion function (RIF) is modeled by: 1 - an approximate ramp functions (RF), and 2 - actual S(cosine) function (CF). A computer code TRANSP19 is developed for the analysis. It models RIA of material test and research reactors (MTR). The code was verified against results obtained from RETRAN 02 and PARET codes and showed a good agreement. The study shows that ETRR-2 core, MTR type, withstands this type of perturbation, fast or slow, in condition that the shutdown system (SHDS) is available. Otherwise the coolant, clad, and fuel temperatures may exceed their design goal values [safety limits (SL)] and the clad may ruptures, coolant may boil, and the fuel may damage. The core reciprocal period never exceeds 140 s-1 for all accident cases (minimum period is 21.98 s) which demonstrates that no explosion other than sonic blockage could occur

  10. Control rod driving hydraulic pressure device

    International Nuclear Information System (INIS)

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

  11. 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.)

  12. Calculations of control rods interference for tajoura research reactor

    International Nuclear Information System (INIS)

    Measurement or experimental determination of the total reactivity worth of the safety and all shim control rods of any research reactor is very essential for the safe operation of the reactor. Integral worth curves are usually used to pre-estimate the criticality positions of the reactor before carrying out any new experiment. These integral worth curves themselves depend very much on the criticality state at which they have been determined. In this paper the computer code CITVAP is used to investigate the shadowing or control rods interference by analytically calibrating the safety, regulating, and shim rods of the Tajoura research reactor. Because experimental integral reactivity curves are usually obtained with control rods positions corresponding to a critical state of the reactor. In our calculations the values of three cases are compared. The case at which the position of the control rods correspond to a critical or close to critical state is taken as a reference case, the values in this case are very close to the experimental values. The obtained results show that shadowing effect is very significant in research reactors, which have small cores with control rods placed close to each other and they also show that for each rod there is an upper limit value and a lower limit value for each rod integral worth and this is very important for the safe operation of the reactor. For the regulating rod which has a small integral reactivity worth, less than 0.5 in our reactor, the upper and lower limits are 0.54 βeff and 0.27 βeff respectively. For a central shim rod they are 4.4 βeff and 3.6 βeff for the other rods the values are tabulated among the results. Due to shadowing effect, the total worth calculated for all the rods together, that is the change in reactivity between the case where all rods are fully raised and the case when they are fully inserted, is higher than the sum of the worth of the individual rods each is being calibrated with the other rods are

  13. Nuclear safety enhancement by structural changes of the control rods

    Energy Technology Data Exchange (ETDEWEB)

    Ciocanescu, M.; Preda, M.; Iorgulis, C.; Truta, C. (Institute for Nuclear Research, Pitesti (Romania))

    1999-12-15

    This paper presents the modification of structure of the control rods which are intended to be done in order to improve the nuclear safety of the TRIGA-SSR-14 MW Pitesti, Romania. In 1996 and 1997 a number of 2 control rod got inoperable due to high corrosion of the welds, water penetration into the absorbent followed by the swelling generated by internal pressure. Although the safe operation of the reactor is not yet affected, it was decided that a more reliable control rod has to be designed and manufactured. Basically, the new control rod will use, as the old ones, boron carbide as absorbent, but boron carbide will be assembled in a different way. There will be manufactured a set of 16 incoloy tubes for each control rod filled with boron carbide pellets and leak tight sealed. These tubes will be mounted in a square array, on the perimeter of the control rod, 5 tubes on each side of the square. The preliminary design on which neutronic analysis on this structure has been performed using specific computation codes. The result of this analyze shows that new control rod reactivity is slightly lower (with about 10%) compared to the older one, but assures a full compensation capacity. Helium release analysis done here reveals that the Helium pressure in the tube at estimated end-of -life would not exceed 57 atm. The advantages of this new concept: it is avoided the problem of helium or radiolysis gases generated into the control rod, and which causes, as we shown, control rod swelling. (orig.)

  14. 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)

  15. 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 spatial effects which have not needed consideration in earlier reactors but which in CAGR would need very careful evaluation. (author)

  16. Experience with General Electric's control rods for boiling water reactors

    International Nuclear Information System (INIS)

    Boron carbide undergoes significant volume increase with irradiation. The results are high stresses in the absorber containment. General Electric's Duralife control rods have used high purity stainless steel since 1983 to mitigate stress corrosion failure resulting from these high stresses. GE's Marathon control rod uses both stress corrosion resistant material and an absorber containment system which reduces the local stress resulting from boron carbide swelling. Hot cell examinations have demonstrated the stress corrosion resistance of high purity stainless steel and the importance of eliminating crevices from high neutron fluence and high stress regions. Material selection was also based on reducing control rod end of life activity. (orig.)

  17. VVER-1000 fresh fuel and CPS control rods entry control

    International Nuclear Information System (INIS)

    Procedures and documentation used for the entry control of fresh fuel and CPS control rods in the Kozloduy NPP are described. Entry control includes: tests of the control equipment; inspection of the containers with the fuel casks; the Nuclear fuel group instructions , supervised by the controlling physicist. There are two types of fresh fuel control - entry control and technical certification of the fuel after the storage term has inspired. The results from the tests show that mechanical damages of the casks in case of violation of the transportation conditions show as changes in the form and disposition of the casks and is easily found by external observation of the casks and inspection of the passability of the assembly. In some cases additional inspections are made. The entry control is completed by the issuance of a permission for operation

  18. Research on control rod drive mechanism seismic test acceptance criteria

    International Nuclear Information System (INIS)

    Background: There is no clear requirement on the rod drop performance of Control Rod Drive Mechanism (CRDM) in seismic condition. Purpose: Acceptance criteria of AP1OOO CRDM seismic test need to be determined. Methods: Related regulations and the safety function of AP1000 CRDM are investigated, as well as the conclusions drawn from the CRDM seismic tests worldwide. Results: Acceptance criteria of this test should be in accordance with the limit is in AP1OOO Nuclear Plant Safety Analysis Report. Conclusions: Drop time of control rods in AP1000 CRDM seismic test at the room temperature without flow is 2.7 s before and after Safe Shutdown Earthquake (SSE). (authors)

  19. Appearance detection device for control rod cluster assembly

    International Nuclear Information System (INIS)

    A reactor control rod cluster assembly (RCCA) is suspended from above, and the appearance is monitored by TV cameras while inserting rods to guide holes of a base stand. The appearance of the entire circumference of the rods can be observed and monitored by mirrors disposed at the periphery of the rods, and images from four directions are taken by two TV cameras to make the monitoring device compact and inexpensive. If the base stand has upper and lower two stages, and the guide holes having mirrors and guide holes having no mirrors are arranged in a point of symmetry, the appearance of all the rods can be observed and monitored only by rotating the RCCA by 180deg and inserting it to the base stand twice. (N.H.)

  20. Controllable single photon stimulation of retinal rod cells

    CERN Document Server

    Phan, Nam Mai; Bessarab, Dmitri A; Krivitsky, Leonid A

    2013-01-01

    Retinal rod cells are commonly assumed to be sensitive to single photons [1, 2, 3]. Light sources used in prior experiments exhibit unavoidable fluctuations in the number of emitted photons [4]. This leaves doubt about the exact number of photons used to stimulate the rod cell. In this letter, we interface rod cells of Xenopus laevis with a light source based on Spontaneous Parametric Down Conversion (SPDC) [5], which provides one photon at a time. Precise control of generation of single photons and directional delivery enables us to provide unambiguous proof of single photon sensitivity of rod cells without relying on the statistical assumptions. Quantum correlations between single photons in the SPDC enable us to determine quantum efficiency of the rod cell without pre-calibrated reference detectors [6, 7, 8]. These results provide the path for exploiting resources offered by quantum optics in generation and manipulation of light in visual studies. From a more general perspective, this method offers the ult...

  1. 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

  2. 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

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

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

  5. Knowledge based system for control rod programming of BWRs

    International Nuclear Information System (INIS)

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

  6. 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

  7. Device for positioning the control rods of a nuclear reactor

    International Nuclear Information System (INIS)

    The description is given of a device for detecting, without any mechanical contact, the position of a mobile component, particularly the control rod of a nuclear reactor, along a given path determined by means of an electromagnetic transducer comprising: (a) a primary winding fed by a current source or alternating voltage to generate an induction flow, (b) a secondary winding inductively connected to the primary winding and fitted with a number of secondary coils spaced out at regular intervals along the path and connected together two by two, the coils of each couple being in phase opposition to form an elemental differential detector, (c) a mobile rod comprising at least one magnetic part for modifying the inductive coupling between the primary and secondary windings when the rod is being moved and so generate a secondary output voltage representing the position of the 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. Replacement of power supply for reactor control rod magnet

    International Nuclear Information System (INIS)

    The magnet power source supply panel for the control rod driving mechanism (CRDM) of JRR-3 adopted series system as a stable DC source required for the electromagnetic coil to hang the control rods by generating magnetic force. Since it had passed 25 years since its installation, it was updated by employing a switching method as countermeasure for aging and for eliminating high heat generation. This paper explains JRR-3 facilities, control rod drive mechanism, and overview of the magnet power supply panel for CRDM. Next, it takes up the replacement of the magnet power supply panel for CRDM, and explains the contents of changes in the power supply method, the contents of changes from existing current amplifier part to updated current amplifier, and resulting high efficiency. Moreover, as the improved points, it explains reduction in switching, and redundant configuration of cooling fan. The single testing and function testing of the apparatus were conducted to confirm the performance. (A.O.)

  14. Evaluation of control rod motion simulator research reactors

    International Nuclear Information System (INIS)

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

  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. Automatic control rod programming for boiling water reactors

    International Nuclear Information System (INIS)

    The objective of long-term control rod programming is to develop a sequence of exposure-dependent control rod patterns that assure the safe and efficient depletion of the nuclear fuel for the duration of the cycle. A two step method was effected in the code OCTOPUS to perform this task automatically for the Pennsylvania and Power Light Co.' BWRs. Although the execution of OCTOPUS provides good or satisfactory results, its input and execution mode has been improved by making it more user friendly and automatic. (authors)

  17. 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

  18. 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.

  19. A flow model for a control rod drop analysis

    International Nuclear Information System (INIS)

    In pressurized water reactor (PWR), the core must be shut down quickly to prevent damage of the reactor internals if any operating limits are exceeded. Therefore, the scram time of control rod is one of the most important design parameters for the safety of the nuclear plant. The control rod drop time is affected by various types of fluid resistance force, so the accurate calculation of the flow field will play a significant role in the drop time prediction. In this paper, a new flow model for a control rod drop analysis of PWR is presented. The calculation of flow between rod and thimble tube is described in detail. According to the drop progress of a control rod, the flow analysis is divided into three phases, 1) above the ventilation holes, 2) between the holes and the entrance of dashpot, 3) after the entrance of dashpot. In each phase, several non-linear equations which describe the flow field are established based on the conservation of energy and flow. Then, the Newton iterative method is used to solve these non-linear equations. A computer code is developed based on this model and several sensitivity analyses are carried out by using this code. The effects of structural design parameters changes, namely, the diameter and the length of dashpot, the diameter of the ventilation holes, on the scram time and the fluid resistance force are discussed and presented. These results show that the new model is useful and accurate in the analysis of control rod drop and the code can be an effective tool for the fuel assembly design of PWR. (author)

  20. 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

  1. 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

  2. Age-related degradation of BWR control rod drives

    International Nuclear Information System (INIS)

    This paper reviews the major age-related degradation mechanisms for U. S. boiling water reactor (BWR) control rod drives (CRDs). Component aging caused by various types of stress corrosion cracking, fatigue, general corrosion, wear, and rubber degradation are discussed. (author)

  3. Inspection of PUSPATI TRIGA Reactor (RTP) core and control rod

    International Nuclear Information System (INIS)

    The 1 MW PUSPATI TRIGA Reactor (RTP), located at Malaysian Nuclear Agency has been operated since its first criticality on 28 June 1982. The RTP uses uranium zirconium hydride fuel enriched to about 20% of U-235. The RTP has four control rods made up of boron carbide where three are fuel-followers and one is an air-follower. The aluminium cylindrical core can accommodate up to 127 fuel elements while the reflector surrounding it is made from high purity graphite. Since, the reactor power is relatively small, natural convection is used for cooling. Light water is used both as a coolant and as well as a moderator. Visual inspection of the core, fuel and control rods are carried out routinely to ascertain their integrity. An underwater camera and boroscope was used to visually inspect the top grid plate of the core as well as the control rods. No visible defect was detected at the top grid plate however, two of the fuel-follower control rods had blemishes on its surface. This paper will describe the findings of the visual inspection as well as corrective actions taken. (author)

  4. Control rod drive mechanism test program. Revision 3

    International Nuclear Information System (INIS)

    A description is given of the testing and development of three control rod drive mechanisms for use on commercial PWR plants designed by B and W. The test results indicate that all three drives are reliable and ensure safe, dependable reactor operation

  5. 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.)

  6. 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)

  7. 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)

  8. 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

  9. Hafnium control rod for nuclear reactors

    International Nuclear Information System (INIS)

    This patent describes an improved control device for nuclear fission reactor having a core fissionable fuel is an assembly composed of fuel units grouped into spaced apart bundles which are immersed in liquid coolant in operating service, and wherein the control device is provided with means for reciprocal movement into and out from the core of fuel intermediate the spaced apart bundles of the assembly. The control device having a frame including an upper and lower support member connected by an elongated central spine support with the upper and lower support members each of transverse cruciform configuration having four radially extending arms projecting outward from the central spine support and a sheath of bladelike configuration extending from each of the arms of the upper support to each of the arms of the lower support and each of the sheaths longitudinally adjoining the spine support. Each sheath containing therein a neutron absorbing component consisting essentially of a plurality of parallel flattened hollow tubes of hafnium metal with their axis aligned with the spine support, and each sheath. Each of the plurality of parallel flattened hollow tubes therein in their flattened portions are provided with a multiplicity of openings along their length for entry and the presence of liquid coolant through the length of the flattened hollow tubes of hafnium

  10. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Hursin, M.; Kochunas, B.; Downar, T. J. [Univ. of California at Berkeley, Berkeley (Canada)

    2008-10-15

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

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

    International Nuclear Information System (INIS)

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

  13. Visual inspections of N Reactor horizontal control rod channels

    International Nuclear Information System (INIS)

    Safety surveillance is performed in horizontal control rod (HCR) channels to locate conditions which could slow or block rod travel. The findings guide the application of preventive measures to assure eventual rod motion impairment will not occur. Borescopes and, more recently, miniaturized closed circuit television (CCTV) cameras have been used for these examinations. Inspections and measurement results are documented in annual surveillance reports, however reported CCTV observations have been limited to highlights. The objective of this report is to catalogue the CCTV recordings in a format suitable for analysis and interpretation and to ease the access to any desired location by noting tape counter readings corresponding with each tube block in view. Searching file tapes for conditions in a specific areas in the past required counting blocks as they passed the camera to determine the distance from a feature like the edge of the reflector or a steam vent gap. This report adds the observations from recent rod channel inspections (1987 and 1988) to a comprehensive survey of graphite conditions in the moderator and reflector regions of the N Reactor core. When completed, the stand-by status of graphite components will be available for use in restart or decommissioning deliberations

  14. Failure of control rod support at magnet assembly

    International Nuclear Information System (INIS)

    On October 2, 1975, the Cornell University TRIGA MARK II Pulsing Reactor experienced a failure of the Safety Control Rod Support Magnet Assembly. During routine steady-state operation at 100 kw the reactor was suddenly shut down by what was indicated as a linear power recorder scram. This was later verified to be caused by an electrical transient. After performing appropriate channel checks, attempts to restore magnet current were unsuccessful, even though all scram circuits were reset. The results of circuit tracing revealed a low-resistance-short across the magnet leads of the Safety Control Rod Drive System. Further investigation located the short circuit inside the magnet assembly. A brief account of locating the trouble and method of repair will be discussed. (author)

  15. 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)

  16. Control rod drives for use in FBR reactors

    International Nuclear Information System (INIS)

    Purpose: To improve thermal fatigue-resistance and corrosion resistance thereby increasing the working life of control rod drives. Constitution: Control rod drives for use in FBR type reactors are made of invar alloy substrate and ceramic layers composed of cordierite composition coated on the surface of the substrates. Since the thermal expansion coefficients for the invar alloy and the ceramic layer are of about 10-6 - 10-7 respectively, thermal stresses due to the transient thermal changes in the surface are small to thereby prevent the generation of fatigue cracks, particularly, those fatigue cracks in the circumferential direction at the surface. Accordingly, blocking of coolants due to the stripping of the cracked portions can be prevented. (Moriyama, K.)

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

    International Nuclear Information System (INIS)

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

  18. Calculation model of the control rod efficiency with mutual interaction

    International Nuclear Information System (INIS)

    This work presents a simplified model derived from the exact perturbation theory and intended for evaluating the total worth of a set of N mutually interacting control rods. The model is based on three basic assumptions validated in previous works. A fourth assumption is made in this work further simplifing the calculations. The results obtained from the proposed model closely agree with those from the standard calculational approach (with relative errors smaller than 2%) rendering the method approapriate for project purposes. (author)

  19. 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...

  20. 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...

  1. 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.

  2. 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

  3. Linear displacement system for nuclear reactor control rods and its operating procedure

    International Nuclear Information System (INIS)

    The linear displacement system described for nuclear reactor control rods is characterized in that two withholding mechanisms of different types and possibly an operating mode are associated with each displacement system. Each withholding mechanism can hold the control rods in the high position. These rods carry out an emergency shutdown of the reactor. Detectors are used to locate the position of the control rods and their holding by the withholding mechanism

  4. 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.

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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)

  7. 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)

  8. 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)

  9. 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.

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

    International Nuclear Information System (INIS)

    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.

  11. A practical method to determine background signals in dynamic control rod reactivity measurement

    International Nuclear Information System (INIS)

    A practical method to eliminate background signals from measured detector signals was developed for the Dynamic Control rod Reactivity Measurement (DCRM). The developed method is applied to determine the rod worth of Nuclear Power Plants in KOREA. (author)

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

    International Nuclear Information System (INIS)

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

  13. Fabrication of Control Rod System of RSG-GAS

    International Nuclear Information System (INIS)

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

  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. Clinch River Breeder Reactor secondary control rod system

    International Nuclear Information System (INIS)

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

  16. Safety and position control system for a nuclear reactor control rod

    International Nuclear Information System (INIS)

    This device comprises a vertically mobile tube, terminating at its bottom end with an electromagnet maintaining the control rod, and of which the upper end is maintained by a second electromagnet, so that when the current to the two electromagnets is cut simultaneously the tube drops under the effect of gravity, thereby helping with its weight to push the control rod into its sleeve, even if the latter has accidental distortions. Application is for nuclear reactors

  17. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-05-15

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

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

    International Nuclear Information System (INIS)

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

  20. Control Rod Worth Measurement in Monju Restart Core

    International Nuclear Information System (INIS)

    The Japanese prototype fast breeder reactor (FBR) Monju resumed the system startup test (SST) in May 2010 after fourteen year and five month suspension since the sodium leakage of the secondary heat transport system in December 1995. Core confirmation test (CCT) is the first step of SST which consists of three steps, and finished in July 2010. Valuable basic data for FBR development was obtained in CCT, such as reactor physics data of the core which contained 1.5wt%/HM in average of Am-241 accumulated due to the Pu-241 decay during the long-term suspension. Control rod reactivity worth measurement was carried out to calibrate the reactivity worth of control rods and to confirm the core characteristics such as excess reactivity and reactivity shutdown margin to be satisfied with safety criteria. The high prediction accuracy of the core management code system was demonstrated based on the measured data. Furthermore, the examination was conducted to shorten the measurement period. (author)

  1. Determination of PWR control rods reactivity worth using the analysis of the power signal during a rod drop

    International Nuclear Information System (INIS)

    The present work is related to the establishment of a validation file concerning a method for determining control rod efficiency. This method is based on the analysis of the power signal obtained during a rod drop in pressurized water reactor (PWR). The main purpose of our investigations was, in particular, to analyse the various conditions which would permit large scale application in PWRS, of the above method. Five different series of measurements were first interpreted to elucidate a theoretical model. This model, in turn, was used to establish the criteria for precision and good representation. As a result of our analysis, it may now be concluded that the power signal method can estimate the total control rod efficiency (normalized to the mean power) within 5%, provided that value of total β is known. This is achieved by taking as a representative signal, the weighted sum of 8 excore detectors and five suitabily placed in core detectors

  2. Detailed B-10 depletion in control rods operatingin a Nuclear Boiling Water Reactor

    OpenAIRE

    Johnsson, John

    2011-01-01

    In a nuclear power plant, control rods play a central role to control the reactivity ofthe core. In an inspection campaign of three control rods (CR 99) operated in theKKL reactor in Leibstadt, Switzerland, during 6 respectively 7 consecutive cycles,defects were detected in the top part of the control rods due to swelling caused bydepletion of the neutron-absorbing 10B isotope (Boron-10). In order to correlatethese defects to control rod depletion, the 10B depletion has in this study beencalc...

  3. Identification of damping mechanism of TRR-II reactor control rod during tree fall insertion

    International Nuclear Information System (INIS)

    In light water reactors, control rods are in general inserted into reactors by gravity. In order to achieve a rapid shutdown, it is required to insert control rods as fast as possible. On the other hand, a control rod with a fast falling velocity would impose a substantial impact to reactor structure as well as to the rod itself. Therefore, a damping force must come into effect, especially during the final stage of the free fall of the control rod. The purpose of this study is to develop a mathematical model and a numerical simulation to describe and identify the damping mechanism; and apply this model to the design of the control rod used in TRR-II reactor of the Institute of Nuclear Energy Research (INER) of Taiwan. The damping effect of a falling control rod comes from two factors: the viscous shear stress occurred in a narrow gap between the rod and an outer tube which confines the lateral movement of the rod, and the pressure force exerted on the rod by the compressed water under the rod. The viscous shear stress can be analyzed by assuming a couette flow between the rod and the outer tube similar to the viscous force occurred in rheology. In doing this, the flow rate in each flow path is closely related to the pressure gradient in the flow path and can be evaluated using an electrical circuit analogy. The results of the code prediction were compared to the experimental results as carried out by the INER. Finally, a parametric study was applied to estimate the effects of the various factors including gap thickness, size of the flow holes, and other geometric considerations on the rod falling velocity. The results of this study can serve some technical support during the stage of rod design and manufacture

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

    International Nuclear Information System (INIS)

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

  5. 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)

  6. Experimental evaluation of the seismic capacity of VVER 440-213 type reactor control rod system

    International Nuclear Information System (INIS)

    The experimental evaluation of the WWER-440/213 control rod drive seismic capacity was carried out on the CKTI-VIBROSEISM Horizontal Shaking Table, specifically designed for seismic testing of the full-scale control rod drive of WWER-440 and WWER-1000 reactors. A detailed description is given of the experimental conditions and the methodology used. The results are compared with technical demands on control rods in emergency situations. (Z.S.) 4 refs

  7. 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

  8. A procedure of measuring the position of the control rods of a nuclear reactor

    International Nuclear Information System (INIS)

    This invention relates to a process for determining the position of all the control rods of a nuclear reactor. A control rod is joined to an operating rod in a magnetic material actuated by electromagnets. The free end of the operating rod moves along the longitudinal centre of a reactance coil that is powered through a high impedance by a constant amperage alternating current. The impedance of the coil varies with the penetration of the operating rod in this coil, therefore the determination of the voltage across the coil terminals indicates the position of the control rod. As there is a large number of control rods in a nuclear reactor, the magnetic coupling between reactance coils close to each other would introduce errors in the determinations. The control rods are therefore distributed into a certain number of groups so that none of these groups includes control rods located close to each other. The groups of rods thus formed are powered sequentially and the voltage across the terminals of the reactance coils is determined only during the times when the power is supplied

  9. 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)

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

    International Nuclear Information System (INIS)

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

  11. Heat resistant driving coil and control rod drive mechanism

    International Nuclear Information System (INIS)

    Ceramic materials are used for each part of driving coils and used as the driving coils for a driving shaft. That is, a cylindrical bobbin having outwardly protruding flanges on the entire circumference at the upper and the lower portions is made of stainless steels. Ceramics sheets are appended as necessary to the outer circumferential surface of the bobbin. Then, ceramic electric wires are wound around the outer circumference of the bobbin by a required number of turns to constitute coils. The electric wire is prepared by coating the conductor of nickel-plated copper with ceramic coating material, disposing an insulation material to the outer circumference thereof the further coating the outside with ceramic coating material. This can improve the heat resistance and, since the control rod drives using such heat resistant driving coils can operate at a high temperature. It requires no cooling device and can simplify the reactor and its peripheral structures. (T.M.)

  12. 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

  13. Establishment of analysis procedure for control rod reactivity worth

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-03-01

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

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

    International Nuclear Information System (INIS)

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

  15. Development for the control device of control rods in the KMRR using a personal computer

    International Nuclear Information System (INIS)

    This report presents that an experimental facility for the control of rods in the KMRR(Korea Multipurpose Research Reactor) is designed and fabricated using four stepping motors and an IBM-PC/XT personal computer. The main structures f the facility are pulse generator, motor driver, rod position and present power indicator, and trip status display and alarm actuator. The rod driving mechanism is composed of four stepping motors, gear boxes, cable sheaves and encoders. Hybrid two phase stepping motors of 1.8 degree step angle are used , and 1-2 phase excitation mode is used. Pulse generator is designed using an INTEL 8085A microprocessor. Two types of shaft encoders are used to control the rod's movement and to detect any error. The encoders used for input side (stepping motor shaft) are 200 pulses per revolution and the other ones for output side(sheave shaft) are 1000 pulses per revolution. The control program is written in Turbo-Basic language, and the control rods can be controlled automatically and manually. (Author)

  16. 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

  17. 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)

  18. Study of friction and wear conditions of control rod drive and control cluster in PWR

    International Nuclear Information System (INIS)

    A good wear resistance is needed for control rod drive, control cluster and fixed guiding systems in internal structure of reactor primary circuit. In an experimental programme life-size material is verified with pressurized water in a loop called SUPERBEC at Cadarache and a research programme is set to find pairs of materials to reduce wear of fixed and moving parts. This programme is divided in three phases: first examination of control rod drive and guide wear, then reproducing wear in a laboratory test and study of material pairs and finally testing of materials in the pressurized water loop

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

    International Nuclear Information System (INIS)

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

  20. Non-linearity correction of control rods worth for critical extrapolation during start-up

    International Nuclear Information System (INIS)

    Distant extrapolation is usually used during the startup of the research reactor, by lifting the control rods step by step to reach the critical state. Due to the non-linearity of the integral worth of the control rods, this process was risky or conservative, especially when the rods were positioned in the non-linear region. A formula could be derived from the point reactor model. in which the reciprocal of the count rate was proportional to Δkeff. Together with the integral worth curve of the control rods, the effect of the non-linearity could be corrected. This method was validated by critical extrapolation data. (authors)

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

    International Nuclear Information System (INIS)

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

  2. An installation for detecting the position of the control rods in a nuclear reactor

    International Nuclear Information System (INIS)

    Description is given of a digitally controlled installation for detecting the position of the control rods of a nuclear reactor. The installation is characterized in that it comprises a magnetic element providing a portion of a driving rod, a plurality of spaced Hall effect transducers distributed in pairs along an axis parallel with the axis of the casing of the driving rods, in the vicinity of the casing outer periphery, and means adapted to receive the output signals of the various transducers for providing an indication of the position of each control rod in the reactor vessel. That installation can be applied to pressurized water nuclear reactors

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

    International Nuclear Information System (INIS)

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

  4. 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)

  5. Control rod calibration methods for fast breeder reactors applied to Phenix

    International Nuclear Information System (INIS)

    The control and the emergency shutdown of a fast breeder reactor depends essentially on control rods. For this reason, it is imperative to know exactly how much anti reactivity is introduced with the rods in the reactor core. Different methods have been compared in order to see if they are compatible with Phenix reactor. Their limits have been studied. The shadow and anti shadow effects that can the rods make one to the other and then their effective weight of the rods screen have been clarified. (N.C.)

  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. 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...

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

    International Nuclear Information System (INIS)

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

  10. Monitoring device for operation of reactor control rod driving mechanism

    International Nuclear Information System (INIS)

    The device of the present invention detects occurrence of abnormality of control rod driving mechanisms in an early stage by extracting changes of a controlling current for the CRDM of a PWR type reactor. Namely, the device of the present invention comprises an abnormality detection and processing device which performs wavelet conversion of signals of the current flowing in a lift coil, signals of the current flowing in a movable griper coil and signals of the current flowing in a stationary griper coil in the CRDM. The device compares the effective value of the wavelet conversion with a previously set reference value. The abnormality of CRDM is analyzed based on the comparative results showing that the effective value of the WAVELET conversion exceeds a predetermined relationship with the reference value. With such procedures, slight change of waveforms can be recognized accurately based on the information represented by three axes, namely, a time axis, the extent of extension/contraction of a base function and a corelationship of the base functions, without using an expensive accelerometer. (I.S.)

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

    International Nuclear Information System (INIS)

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

  12. 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

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

    International Nuclear Information System (INIS)

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

  14. 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)

  15. Control rod driving mechanism, and control device and operation method therefor

    International Nuclear Information System (INIS)

    The upper portion of a housing of control rod driving mechanisms is secured to a reactor pressure vessel, and the lower portion thereof is sealed by a closing plug. Gears are formed on the outer circumference of a driving shaft vertically moving with the linkage of a control rod in a pressure vessel, and a linear reluctance motor comprising a stator iron core having gears on the inner circumference of a stator and a stator coil for driving the driving shaft. There are disposed a latch mechanism for holding the control rod by engaging with the gears of the driving shaft and a position detector for detecting the position of the inserted control rod by the gears of the driving shaft or magnets mounted to the gears. Since the inner structure can be simplified with no shaft-sealing portion, the frequency for the maintenance and inspection can be reduced to improve the reliability of sealing portions of the pressure vessel. The space for maintenance and inspection of the lower portion of the pressure vessel can be reduced thereby making the height of a reactor building low and strengthen the earthquake proof structure. (N.H.)

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

    International Nuclear Information System (INIS)

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

  17. 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.

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

    International Nuclear Information System (INIS)

    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

  19. Aging assessment of BWR control rod drive systems

    Energy Technology Data Exchange (ETDEWEB)

    Greene, R.H.

    1991-01-01

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

  20. Aging assessment of BWR control rod drive systems

    International Nuclear Information System (INIS)

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

  1. 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.

  2. Control rod drive housing made of recrystallised centrifugal casting

    International Nuclear Information System (INIS)

    The possibility of fabricating the Control Rod Drive (CRD) Housings for Boiling Water Reactor (BWR) pressure vessels by recrystallised centrifugal casting without welding was investigated. The material used was to the specification ASME SA-351 CF3M, and after centrifugal casting it was cold formed by a process involving 15% to 30% plastic deformation before the final solution heat treatment, so that recrystallisation can be obtained. Various mechanical, metallurgical and nondestructive tests were performed, and the following superior characteristics were confirmed. Metallurgical tests showed that small austenite grain and 11 to 12% isolated ferrite are uniformly distributed throughout the whole parts of the CRD housing. Tests at room temperature, and up to 3000C on tensile specimens taken from various portions of the CRD housing were satisfactory. The weldability of the recrystallised casting CRD housing to Inconel stub tube was satisfactory. The attenuation of ultrasonic waves of the recrystallised casting was compared with those of cast and wrought material, and it was confirmed that ultrasonic tests can be performed with almost the same sensitivity as on wrought materials. When such nondestructive ultrasonic tests were made on the whole length of a typical sample, supported by liquid penetrant tests, no defective indications were found. (U.K.)

  3. Indirect air cooling techniques for control rod drives in the high temperature engineering test reactor

    International Nuclear Information System (INIS)

    The high temperature engineering test reactor (HTTR) is the first high-temperature gas-cooled reactor in Japan with reactor outlet gas temperature of 950 deg. C and thermal power of 30 MW. Sixteen pairs of control rods are employed for controlling the reactivity change of the HTTR. Each standpipe for a pair of the control rods, which is placed on the top head dome of the reactor pressure vessel, contains one control rod drive mechanism. The control rod drive mechanism may malfunction because of reduction of the electrical insulation of the electromagnetic clutch when the temperature exceeds 180 deg. C. Because 31 standpipes stand close together in the standpipe room, 16 standpipes for the control rods, which are located at the center, should be cooled effectively. Therefore, the control rod drives are cooled indirectly by forced air circulation through a pair of ring-ducts with proper air outlet nozzles and inlets. Based on analytical results, a pair of the ring-ducts was installed as one of structures in the standpipe room. Evaluation results through the rise-to-power test of the HTTR showed that temperatures of the electromagnetic clutch and the ambient helium gas inside the control rod standpipe should be below the limits of 180 and 75 deg. C, respectively, at full power operation and at the scram from the operation.

  4. 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)

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

    International Nuclear Information System (INIS)

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

  6. Development of control rod driving mechanism for high neutron flux reactor in Kyoto University (KUHFR)

    International Nuclear Information System (INIS)

    KUHFR is a coupling type reactor of 30 MW power output, which have two light-water-moderated and cooled cores inside the heavy water reflector. There are six sets of control rod driving mechanism (CRDM) in each core, each set driving one control rod. The newly developed driving system for CRDM is a unique one not employed in any other reactor. The main specifications required are as follows: Drive length 650 mm, driving speed 100 mm/min; control rod magnet deenergizing time 0.3 sec or less, control rod falling time to 90% stroke 1 sec or less, finished O.D. 190 mm or less. There were difficulties in selecting the driving system, because various control rod driving systems adopted in power and research reactors have both merits and demerits. As a result of investigation, three systems have been produced for trial, experimented and compared, and the moving coil type CRDM has been employed because it is suitable in many points, e.g. it allows continuous motion of control rods. The construction of moving coil type CRDM is explained. In the progress of development from No. 1 to No. 3 system is described, starting at the magnetic circuit calculation. As the running performance of the CRDM, the relationship between the plunger shift in a coil and upward force, and the differential linear running performance, following properties and stopping characteristics of control rods for coil movement are described. (Wakatsuki, Y.)

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

    International Nuclear Information System (INIS)

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

  8. 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)

  9. 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.

  10. 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)

  11. 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.

  12. Seismic appraisal test of control rod drive mechanism of China experiment fast reactor

    International Nuclear Information System (INIS)

    The structure of the control rod drive mechanism in pool type sodium-cooled fast reactor is the characterized by long, thin, and geometric nonlinearity, and the seismic load is multiple activation. The anti-seismic evaluation is always paid great attention by the countries developing the technology worldwide. This article introduces the seismic appraisal test of the control rod drive mechanism of China Experimental Fast Reactor (CEFR) performed on a seismic platform which is vertical shaft style and multiple activation. The result of the test shows the structural integrity and the function of the control rod drive mechanism could meet the design requirements of the earthquake intensity. (authors)

  13. 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.

  14. 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

  15. 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

  16. 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.)

  17. Control rod upper pin and roller removal and replacement

    International Nuclear Information System (INIS)

    There has been increasing utility interest in removing cobalt sources as a means to minimize dose rates and shipping costs associated with operations, maintenance, shipping and radioactive waste disposal. This activation product accounts for approximately 80% of the dose rates at a plant. Estimates suggest that older (original equipment) control rod blades (CRBs) using stellite rollers and Haynes Alloy 25 pins can contribute as much as 43% of the cobalt in the plant. Since CRBs are the most concentrated source of cobalt in the reactor vessel, pin and roller (P ampersand R) replacement will provide significant reductions in cobalt-60 levels. The reductions will be measurable within 2 years, and should plateau within 5 to 6 years. This paper described a new, simple, cost effective, field proven in-situ method to replace these high source term contributing components and return the CRB to the reactor core for continued service. The paper also briefly describes a methodology that has been developed to estimate the labor and material costs along with a man-rem dollar savings for removal of the upper P ampersand Rs from in-service CRBs purchased prior to 1982. This method permits comparison to the cost of early replacement of CRBs. The result of this study is an EXCEL spreadsheet cost/benefit analysis that can be made plant specific by incorporation of plant data. The Analysis shows that the removal of P ampersand Rs can be beneficial for a Co-60 contribution as low as 5-10%. Recent interest by non-utility radioactive material users in obtaining the removed P ampersand Rs may further increase the benefit and reduce the cost associated with in-situ removal

  18. 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)

  19. 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

  20. 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)

  1. Core reactor with a reactor core which can be controlled by control rods

    International Nuclear Information System (INIS)

    In order to move and position the control rods, the piston-cylinder units of the drives, where there are stepped sections with narrow and wide gaps opposite one another, are connected via a single control pipe each to a valve block. The valves for controlling several units are combined there and are connected to a common pump for the coolant. The valve block with all the control pipes carrying the coolant is accommodated in a sealed metal component in the reactor pressure vessel, so there can be no leaks to the outside. The invention is particularly suitable for reactors of small power based on the boiling water principle. (orig./HP)

  2. 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)

  3. The research on control rod insertion of a boiling water reactor with water hydraulic drive

    International Nuclear Information System (INIS)

    This thesis reports on the hydraulic driving system, powered by an accumulator. This drive system is mainly used for the drive of control rods of nuclear reactors. In case of strong earthquakes, control rods are set in gaps between fuel assemblies to scram nuclear reactors. Characteristics of the system have not been analyzed. The analysis of this system is necessary in order to present the designs that are intended to be a variety of situations. So we developed the model of the hydraulic control rod driving system. The model that we have created is able to reproduce the actual driving. Also, there is a load on the system by an earthquake. This load is caused by the contact of the deformed fuel assembly and control rod. This load model is obtained by solving the equation of motion of the beam. (author)

  4. 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

  5. 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)

  6. Leaked water detection device for control rod drive and BWR type reactor

    International Nuclear Information System (INIS)

    The device of the present invention can specify a control rod drive causing great amount of water leakage among a large number of control rod drives. Namely, water leaked from the control rod drives is introduced to each of leaked water pipelines. Further, it is introduced from the leaked water pipelines to flow glasses at which leaked water can visually be recognized individually, and then discharged through a drain pipeline. With such procedures, the amount of leaked water from the leaked water pipelines can visually be recognized at the flow glasses. As a result, the control rod drives which cause a great amount of leakage can be specified among large number of control rod drives. Accordingly, an accurate inspection schedule for a shaft-sealing portion of the control rod drives can be formed. The shaft-sealing portion degradated in the sealing property can reliably be inspected and repaired. Purge water can be ensured to improve reliability of the operation of equipments. (I.S.)

  7. 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

  8. A hybrid attitude controller consisting of electromagnetic torque rods and an active fluid ring

    Science.gov (United States)

    Nobari, Nona A.; Misra, Arun K.

    2014-01-01

    In this paper, a novel hybrid actuation system for satellite attitude stabilization is proposed along with its feasibility analysis. The system considered consists of two magnetic torque rods and one fluid ring to produce the control torque required in the direction in which magnetic torque rods cannot produce torque. A mathematical model of the system dynamics is derived first. Then a controller is developed to stabilize the attitude angles of a satellite equipped with the abovementioned set of actuators. The effect of failure of the fluid ring or a magnetic torque rod is examined as well. It is noted that the case of failure of the magnetic torque rod whose torque is along the pitch axis is the most critical, since the coupling between the roll or yaw motion and the pitch motion is quite weak. The simulation results show that the control system proposed is quite fault tolerant.

  9. 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.

  10. 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

  11. Atucha-2 obliquely inserted control rods RELAP5-3D/NESTLE model

    International Nuclear Information System (INIS)

    Atucha-2 is a Siemens-designed PHWR reactor in phase of commissioning in the Republic of Argentina. Its geometrical complexity and peculiarity (e.g., oblique control rods, positive void coefficient) required a developed and validated complex three dimensional (3D) neutron kinetics (NK) model. In the framework of the agreement between NASA and University of Pisa a detailed NESTLE (three-dimensional neutron kinetics code) model of the Atucha-2 NPP was developed. This document summarizes the procedures for the implementation of the oblique control rods into the RELAP5/NESTLE model: a particular arrangement of RELAP5/NESTLE control rods insertion mode for such kind of oblique control rods and an implementation into the homogenized two group cross sections of ad-hoc calculated correction factors (these parameters were obtained by previously executed Monte Carlo calculations) was developed. Some applications, among the scenarios selected to perform safety analysis of the Atucha-2 NPP (CNA-II), are also reported: preliminary Scram Rod Worth, analysis of a Control Rod Ejection Accidents and a CR faulty withdrawal. (author)

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

    International Nuclear Information System (INIS)

    An extensive series of control rod measurements was made in the Aagesta reactor during the low power experimental period following the first criticality. This report describes the part of these investigations made with the pulsed neutron method, comprising nearly 300 measurements. The main objective was the determination of control rod reactivity worths for different rods and groups of rods, but some supplementary measurements were also made, e.g. a determination of the prompt neutron decay constant for the delayed critical condition and four different cores. The cores consisted of 20, 32, 68, and 140 fuel elements respectively, and measurements were made at room temperature and with the moderator level close to critical for each core, and for the 140-element core also with full moderator height and at the temperatures 140 deg C and 215 deg C. Both fully and partly inserted control rod groups were investigated. The measurements at critical water level give directly the control rod reactivity worths, whereas those with full water height give the shut-down reactivity. A comparison was made between measured reactivity worths for a number of rod groups and those calculated with the HETERO code. The prompt neutron decay constant at delayed criticality α0=β/l, for the full core at 215 deg C was found to be 9.60 ± 0.30/sec, corresponding to l = 0.76 ± 0.02 msec. The shut-down reactivity with 16 coarse control rods in pos. A-D 22, 40-04, 44, 26 is -5% at 25 deg C and -13% at 215 deg C. The relative error is usually around 8% in the reactivity worths, originating mainly from the higher harmonics content in the measured curves

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

    International Nuclear Information System (INIS)

    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. Activation calculation of steel of the control rods of TRIGA Mark III reactor

    International Nuclear Information System (INIS)

    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)

  15. 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)

  16. Aerosol behavior during SIC control rod failure in QUENCH-13 test

    International Nuclear Information System (INIS)

    In a nuclear reactor severe accident, radioactive fission products as well as structural materials are released from the core by evaporation, and the released gases form particles by nucleation and condensation. In addition, aerosol particles may be generated by droplet formation and fragmentation of the core. In pressurized water reactors (PWR), a commonly used control rod material is silver-indium-cadmium (SIC) covered with stainless steel cladding. The control rod elements, Cd, In and Ag, have relatively low melting temperatures, and especially Cd has also a very low boiling point. Control rods are likely to fail early on in the accident due to melting of the stainless steel cladding which can be accelerated by eutectic interaction between stainless steel and the surrounding Zircaloy guide tube. The release of the control rod materials would follow the cladding failure thus affecting aerosol source term as well as fuel rod degradation. The QUENCH experimental program at Forschungszentrum Karlsruhe investigates phenomena associated with reflood of a degrading core under postulated severe accident conditions. QUENCH-13 test was the first in this program to include a silver-indium-cadmium control rod of prototypic PWR design. To characterize the extent of aerosol release during the control rod failure, aerosol particle size distribution and concentration measurements in the off-gas pipe of the QUENCH facility were carried out. For the first time, it was possible to determine on-line the aerosol concentration and size distribution released from the core. These results are of prime importance for model development for the proper calculation of the source term resulting from control rod failure. The on-line measurement showed that the main aerosol release started at the bundle temperature maximum of T ∼ 1570 K at hottest bundle elevation. A very large burst of aerosols was detected 660 s later at the bundle temperature maximum of T ∼ 1650 K, followed by a

  17. 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).

  18. 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)

  19. Design, Fabrication and Testing of the Control Rods for the Experimental Gas-Cooled Reactor

    International Nuclear Information System (INIS)

    The criteria and methods used for the design of the control rods for the Experimental Gas-Cooled Reactor are described. The final mechanical design was derived from extensive thermal and mechanical calculations and actual experience obtained by fabrication of a prototype rod. The nuclear design of the rod was based on detailed calculations, the accuracy of which was checked by comparison with a measurement of rod worth made with the Physical Constants Test Reactor; By means of a meticulous application of basic principles the calculation agreed with the measurement within the experimental uncertainty. The most important nuclear aspect of the design is the large amount of epithermal absorption, which approximately doubles the worth over that of a purely thermal absorber. The rod is of an articulated type and consists of hot-pressed B4C-bushings clad in stainless-steel. The unique design of the load-supporting members allows operation at cladding temperatures up to 1600°F. Comparisons are made with control-rod designs for other gas-cooled reactors, and justifications for the choice of design features and material selection are discussed. The fabrication procedures and the final test programme for verification of the adequacy of the design are described. (author)

  20. Evaluation of MONJU core damage risk due to control rod function failure

    International Nuclear Information System (INIS)

    MONJU is a sodium-cooled, loop-type prototype fast breeder reactor with three primary cooling loops that can supply 280 MW of electricity. The limiting conditions of operation defined in the safety regulations for MONJU given the allowed outage time were evaluated by a probabilistic safety assessment technique in our previous study. If a function failure is found in a control rod, certain measures are required by the safety regulations. In this case, if it is confirmed within 24 h that no other control rods are stuck, reactor operation is allowed to continue. To assess the validity of the 24 h allowable time in view of core damage risk, it is necessary to analyze the conditions to be changed when a stuck rod is discovered. Furthermore, to develop a method for evaluating the probability of a control rod insertion failure, it is necessary to re-estimate the frequency of core damage under control rod insertion failure conditions. This paper describes a method for this re-estimation. The probability of an insertion failure of one control rod has been calculated by applying Jeffrey's noninformative prior distribution by considering insertion times based on the results of a mock-up test. The necessary rod insertion numbers for the main reactor shutdown and backup shutdown systems were considered. The results showed that a completion time of 24 h gives a safety margin comparable to that of the Incremental Conditioned Core Damage Probability, that is, an acceptable risk threshold represented by the U.S. Nuclear Regulatory Commission RG 1.177. Thus, the timeframe defined in the present safety regulations was concluded to be appropriate. (author)

  1. Study of the effect of heterogeneity of the control rods in the PHÉNIX reactor

    International Nuclear Information System (INIS)

    Neutron cross-section processing for fast reactor sub-regions containing control rods has to take into account heterogeneity effects in order to get a reliable assessment of the control rod reactivity worth. Several numerical methods have been developed in the past with a support of experimental campaigns and are employed nowadays to reasonably treat such effects by using sophisticated neutronics codes. The SIMMER code is employed at KIT for severe accident analyses of metal-cooled fast reactors and other reactor systems. Neutron cross-sections are processed in the original SIMMER version by approximating each reactor region as a homogeneous medium. This simplified treatment results in the overestimation of the control rod worth. Efforts are therefore going on to extend the code in order to take into account heterogeneity effects in the control rod subassemblies. In this paper a technique proposed in the past was applied to take into account these effects as basis for further SIMMER extensions. With this aim a 3D (HEX-Z) ERANOS model of the PHÉNIX reactor has been assessed. Further, the corresponding 3D (XYZ) PARTISN model has been employed, the latter code being now introduced in SIMMER as a new neutronics solver. Results show that the employed technique improves the capability of SIMMER and ERANOS codes to predict the control rod reactivity worths. Results also show that the HEX-Z ERANOS and XYZ PARTISN models for PHÉNIX reasonably agree. (author)

  2. Experiment and analysis of B4C simulating control rod on FCA V-3 Assembly

    International Nuclear Information System (INIS)

    Reactivity worths of the B4C simulating control rods have been measured and analysed on FCA V-3 Assembly which was constructed as the engineering mock-up for Experimental Fast Reactor ''JOYO''. Assembly V-3 differs from JOYO in the blanket composition and the simulating control rod is 1/2 of that of JOYO in size. We have made efforts to check the adequacy of the nuclear design method and to improve the design accuracy for JOYO by determining the range of ratios of the theoretical to experimental values (C/E). The reactivity worth of the B4C control rod is obtained by the measurement of the sub-criticality of the system containing the control rod. In the present work the neutron source multiplication method was employed. In the calculation we employed the multigroup diffusion approximation for the core and blanket, and the collision probability method for the effective cross sections of the B4C simulating control rod region. The cross section set used in the calculation is JAERI-FAST Version II. The lowest limit of the sub-criticality of the system is -6% delta k/k and the C/E ranges from 1.00 to 1.03 in the present work. (author)

  3. 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.)

  4. Method of controlling moving-coil type control rod driving mechanisms

    International Nuclear Information System (INIS)

    Purpose: To enable solenoid plungers to sufficiently follow after abrupt changes of moving speed of moving-coils in nuclear reactors. Method: In a control circuit for moving-coil type control rod driving mechanisms of nuclear reactors, the velocity of a driving device for the moving-coils is detected by a velocity detector to control the velocity change of exciting currents in the coils depending on a velocity instruction signal. Since the velocity change of the coil exciting current varies depending on the change in the velocity instruction signal, the solenoid plunger can smoothly follow after the moving coils electromagnetically coupled therewith, and the deviation between the moving-coils and the solenoid plunger, that is, the driving axis can be minimized. Accordingly, smooth reactor control can be attained. (Takahashi, M.)

  5. Arrangement for operation of a boiling water reactor where after an operating period some control rods are replaced with control rods of a higher control value

    International Nuclear Information System (INIS)

    The invention is concerned with a method for obtaining longer reactor operating cycles and/or higher power outputs from a BWR reactor. At a partial refueling of the reactor, only some of the regulating rods - in the centre of the core - are exchanged for rods of a higher regulating worth than the original regulating rods (10-20% higher). 50-80% of the regulating rods in the centre of the core are replaced, while the other positions are occupied by regulating rods that have already been used during the proceeding operating period. (L.E.)

  6. 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.

  7. 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.

  8. Development and the results for the control rods in MKII core of experimental fast reactor Joyo

    Energy Technology Data Exchange (ETDEWEB)

    Miyakawa, Shunichi; Soga, Tomonori; Takatsuto, Hiroshi [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1997-07-01

    Since the first control rod design for the Joyo MKII core (about twenty years ago), there have been several challenging improvements; for example, a helium venting mechanism and a flow induced vibration prevention mechanism. Forty-four control rods with these various modifications have been fabricated. To date, thirty-four have been irradiated and the sixteen have been examined. This experience and effort has produced fruitful results: (1) Efficiency and reliability of the diving-bell type Helium venting mechanism. (2) Efficiency of the flow induced vibration prevention mechanism. (3) Efficiency of the improvement for scram damping mechanism. (4) Clarification of absorber-pellet-cladding-mechanical-interaction (ACMI) phenomena and preventive methods. The fourth result listed above has been a subject of investigation for fifteen years in several countries, that is a main phenomena to dominate control rod life time. The results of this investigation of ACMI in absorber elements are discussed. (J.P.N.)

  9. 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.

  10. 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)

  11. On-line critical control rod pattern prediction algorithm for BWR plant startup

    International Nuclear Information System (INIS)

    This paper describes an on-line algorithm for predicting the critical control rod pattern, which has been developed to reduce the mental strain on operators while withdrawing control rods in the BWR plant startup operation. The proposed algorithm estimates a target eigenvalue (eigenvalue bias) for a three-dimensional neutron kinetics model with a neutron source incorporating actual neutron detector readings. The critical control rod pattern is then predicted based on the estimated eigenvalue bias. The algorithm has been verified using data obtained from an actual startup operation on a BWR model-5 plant, and the estimated eigenvalue bias agreed well with the effective multiplication factor at the criticality actually determined from the operator's judgement. (author)

  12. A survey of control rod measurements in ZPPR and their analysis

    International Nuclear Information System (INIS)

    A large number of measurements of control rod worths have been made at ZPPR over the past 16 years, covering a wide range of fast reactor core designs. Both experimental techniques and analytical methods have improved over this period. The results of analysis using ENDF/B-IV and ENDF/B-V.2 nuclear data are reviewed and the calculation methods employed are described in some detail. Special experiments of control rod heterogeneity and boron enrichment effects have been made to aid extrapolation from critical experiments to power reactors. The analysis of parameters related to insertion of control rods into the core, such as fission rate distributions, is also summarized. (author). 15 refs, 3 figs, 10 tabs

  13. 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

  14. 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.)

  15. 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)

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

    Directory of Open Access Journals (Sweden)

    Nusrat Jahan

    2011-09-01

    Full Text Available 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 was implemented with a dedicated circuit assembly and a conventional personal computer. A high-level Visual Basic real-time programming has been developed for data acquisition, reactivity calculation, online display (numerically as well as graphically, saving data, etc. To measure reactivity worth of TRIGA reactor control rods the rod drop experimental technique has been adopted. The results of tests experiments, carried out with the rod drop method for measuring various reactivity worth of control rods have been presented in the paper. A comparison between this results with the results using period method and that of computation method, demonstrated that the response of this reactivity measurement system is fast enough to monitor and measure the safety-related reactivity and power excursions in the reactor.

  17. 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

  18. Position indicator for movable coil type reactor control rod driving mechanism

    International Nuclear Information System (INIS)

    Purpose: To enable the accurate and continuous indication of the position of a movable coil type reactor control rod driving mechanism. Constitution: The position of an electromagnet magnetically coupled to a plunger connected to a reactor core control rod is detected by an electromagnet position detector, and the displacement of the positions of the electromagnet and the plunger is detected by a relative position detector connected to the electromagnet. The detected values of both the detectors are used to calculate the position of the driving mechanism. (Aizawa, K.)

  19. 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)

  20. 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)

  1. 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)

  2. 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)

  3. Investigating The Integral Control Rod Worth Of The Miniature Neutron Source Reactor MNSR

    International Nuclear Information System (INIS)

    Determining control rod characteristics is an essential problem of nuclear reactor analysis. In this research, the integral control rod worth of the miniature neutron source reactor MNSR is investigated. Some other parameters of the nuclear reactor, such as core excess reactivity, shut down margin, are also calculated. Group constants for all reactor components are generated by the WIMSD code and then are used in the CITATION code to solve the neutron diffusion equations. The maximum relative error of the calculated results compared with the measurement data is about 3.5%. (author)

  4. Mechanical analysis of a boron carbide control rod for pressurized water reactor application

    International Nuclear Information System (INIS)

    A control rod using boron carbide as the neutron poison was analyzed for use in a pressurized water reactor. The motivation for this study stems from the increasing use of boron carbide in control rod elements for large commercial nuclear power stations and the potential safety hazard related to excessive inelastic cladding deformation due to pellet-cladding interaction. As such, the radiation induced dilatation of the boron carbide pellets and the ensuing pellet-cladding interaction phenomena were the dominant concerns of this investigation. Based on the small deformation theory of linear elasticity, the method developed herein can be used to predict that rod burnup which initiates the yielding of the cladding. The details of the physical properties of boron carbide and related design data required for the analysis are included

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

    International Nuclear Information System (INIS)

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

  6. Comparison of transport and diffusion theory for control rod worths in the CONRAD-ST cores

    International Nuclear Information System (INIS)

    In this paper control rod worth calculations using the MARC/PN code are presented, and used to highlight any differences between diffusion and transport theory for the CONRAD configuration ST-C1R. Also the calculations can be used to examine any interaction effects between separate rings of rods, and asymmetric insertion modes. The CONRAD experiments using MASURCA have been devised to examine advanced fast reactor designs such as heterogeneous cores, and large cores. The core studied in this paper constitute a single phase of the CONRAD programme and are devoted to studying control rod reactivity and power distribution sensitivity to basic data. Two cores are used to examine different enrichment zone configurations in a large conventional homogeneous core. 3 refs, 5 figs, 9 tabs

  7. Analysis of control rod withdrawal end-of-life tests in the PHENIX reactor at IGCAR

    International Nuclear Information System (INIS)

    The post-analysis of control rod withdrawal end-of-life experiments, performed in the PHENIX reactor during June 2009, were carried out at IGCAR as a part of IAEA Coordinated Research Project (CRP) on “Control rod withdrawal and sodium natural circulation tests performed during the PHENIX end-of-life experiments”. The main objective of this CRP was to assess the prediction capability of distorted radial power distribution due to absorber rod movement (insertion and/or withdrawal) at nominal power. In addition, core reactivity, absorber rod worth, S-curve, maximum neutron flux, SA-wise power and sodium heating deviation with respect to the reference core for various critical core configurations were analyzed. 3-D diffusion theory calculations using FARCOB (IGCAR) and ERANOS-2.1 (European) code systems were made for this analysis. IGCAR prediction of radial power distribution due to absorber rod movement is very close to the measured values. No significant deviation is observed between the results of FARCOB and ERANOS-2.1 and also between the results of other CRP participants. This benchmark exercise has provided a wide international forum to verify the method, computer codes and cross section data employed at IGCAR for the physics design calculations of sodium cooled fast reactors, by keeping the experimental data as the reference. (author)

  8. 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)

  9. Demonstration of EBR-II power maneuvers without control rod movement

    International Nuclear Information System (INIS)

    A group of five plant inherent control tests was successfully conducted in November 1987 in the Experimental Breeder Reactor II. These tests demonstrated that the plant power of a metal-fueled reactor can be passively controlled over a large power range by slowly changing the primary flow and the reactor inlet temperature. These variables are, in turn, regulated by the primary pump speed, the secondary flow, and the turbine inlet pressure. In all tests, control rods were not used to regulate power. It was demonstrated that the plant power can be controlled with reasonable accuracy without using control rods when the reactivity feedback characteristics of the reactor are well understood and the plant controllers are adequately designed

  10. 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.

  11. Parallel Magnetic Flow Electromagnet for Movable Coil Control-rod Driving Mechanism

    International Nuclear Information System (INIS)

    The parallel magnetic flow electromagnet can effectively relax the saturation, which easily takes place in the single magnetic flow electromagnet, and accordingly can improve the drive capacity of the movable coil electromagnet drive mechanism for a mobile reactor control rod. (authors)

  12. Fuel loading and control rod patterns optimization in a BWR using tabu search

    International Nuclear Information System (INIS)

    This paper presents the QuinalliBT system, a new approach to solve fuel loading and control rod patterns optimization problem in a coupled way. This system involves three different optimization stages; in the first one, a seed fuel loading using the Haling principle is designed. In the second stage, the corresponding control rod pattern for the previous fuel loading is obtained. Finally, in the last stage, a new fuel loading is created, starting from the previous fuel loading and using the corresponding set of optimized control rod patterns. For each stage, a different objective function is considered. In order to obtain the decision parameters used in those functions, the CM-PRESTO 3D steady-state reactor core simulator was used. Second and third stages are repeated until an appropriate fuel loading and its control rod pattern are obtained, or a stop criterion is achieved. In all stages, the tabu search optimization technique was used. The QuinalliBT system was tested and applied to a real BWR operation cycle. It was found that the value for k eff obtained by QuinalliBT was 0.0024 Δk/k greater than that of the reference cycle

  13. Experimentation with the prototype of the PEC control rod operating mechanism (PCROM): washing activities and results

    International Nuclear Information System (INIS)

    Experimentation on prototypes of Pec components is presently being carried out at Casaccia Cre. This report shows the results of the first cycle of experimentation of the control rods operating mechanism prototype (Pcrom), concerning the aspects of sodium removal and the checks after experimentation. The activities carried out for the finalization of the washing procedure are also reported

  14. Study of structural condition of WWR-K reactor automatic control rod material

    International Nuclear Information System (INIS)

    In this paper the results of structural investigation at WWR-K thermal research reactor automatic control (AC) rod materials of SAV-1 aluminium alloy and Kh18N10T austenitic steel. The structural investigation was performed with help of scanning and transmission electronic microscopy methods. (author)

  15. MCNP evaluation of top node control rod depletion below the core in KKL

    International Nuclear Information System (INIS)

    In previous studies, there has been identified a significant discrepancy in the BWR control rod top node depletion between the two core simulator nodal codes POLCA7 and PRESTO-2, which indicates that there is a large general uncertainty in nodal codes in calculating the top node depletion of fully withdrawn control rods. In this study, the stochastic Monte Carlo code MCNP has been used to calculate the top node control rod depletion for benchmarking the nodal codes. By using the TIP signal obtained from an extended TIP campaign below the core performed in the KKL reactor, the MCNP model has been verified by comparing the axial profile between the TIP data and the gamma flux calculated by MCNP. The MCNP results have also been compared with calculations from POLCA7, which was found to yield slightly higher depletion rates than MCNP. It was also found that the 10B depletion in the top node is very sensitive to the exact axial location of the control rod top when it is fully withdrawn. By using the MCNP results, the neutron flux model below the core in the nodal codes can be improved by implementing an exponential function for the neutron flux. (author)

  16. UK studies of the performance of boron carbide control rod pins for the fast reactor

    International Nuclear Information System (INIS)

    The preferred neutron absorbing material in control rods for modern fast reactors is boron carbide. This report presents the current status of the UK programme on the development of boron carbide control rod pins. The objective of the programme is to maximise the life of the pins, initially for the UK Prototype Fast Reactor (PFR) and, more recently, for the European Fast Reactor (EFR). The pin life is currently assessed against three criteria, the onset of pellet-cladding mechanical interaction at power, the boron carbide pellet centre temperature, and cladding embrittlement due to the combined effects of irradiation damage and pellet-cladding chemical interaction. Results are presented from the post-irradiation examination of static pins exposed in demountable sub-assemblies in PFR and pins from PFR control rods. The variables include stainless steel [M316 (CW)] and nimonic [PE16] cladding, sodium and helium pin filling, top and bottom pin gas venting and boron carbide with two levels of 10B enrichment from different sources. The results obtained are compared with the 'BORCON' computer model of fast reactor control rod pin performance. (author)

  17. Mechanical tests of the bolt of the gripper latch on the control rod cluster

    International Nuclear Information System (INIS)

    Failure analysis and mechanical testing indicate that control rod drive mechanisms malfunctioning by 1995-96 is due to rupture by fatigue of a bolt inside the stationary gripper assembly. Fatigue is enhanced by free working following surface adaptation and unscrewing of the assembly. These results are taken into account for the choice of a new anti-rotation device. (authors)

  18. Aging assessment of the Combustion Engineering and Babcock and Wilcox control rod drives

    International Nuclear Information System (INIS)

    The effects of aging upon the Babcock ampersand Wilcox (B ampersand W) and Combustion Engineering (CE) control rod drive systems have been evaluated. For this study, the CRD system boundary included the control rod assemblies, guide tubes, control rod drive mechanism, control system components, rod position indication components, and cooling system. Detailed operation experience data for 1980 to 1990 was evaluated to identify the predominant failure modes, causes, and effects. The results of this evaluation, along with an assessment of component material and operating environment, lead to the conclusion that both the B ampersand W and CE CRD systems are susceptible to age degradation. Failures of the CRD system have resulted in significant plant effects including power reductions, plant shutdowns, scrams, and ESF actuations. Information on current plant system inspection and maintenance practices were obtained from two B ampersand W plants, and four CE plants through an industry survey. The results of this survey indicate that some plants have modified the system, replaced components, and established preventive maintenance programs, some of which effectively address the aging issue, while others do not. The potential application of some advanced monitoring inspection techniques are discussed

  19. 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.; Hourigan, K.

    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...

  20. 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).

  1. Quantitative uncertainty and sensitivity analysis of a PWR control rod ejection accident

    Energy Technology Data Exchange (ETDEWEB)

    Pasichnyk, I.; Perin, Y.; Velkov, K. [Gesellschaft flier Anlagen- und Reaktorsicherheit - GRS mbH, Boltzmannstasse 14, 85748 Garching bei Muenchen (Germany)

    2013-07-01

    The paper describes the results of the quantitative Uncertainty and Sensitivity (U/S) Analysis of a Rod Ejection Accident (REA) which is simulated by the coupled system code ATHLET-QUABOX/CUBBOX applying the GRS tool for U/S analysis SUSA/XSUSA. For the present study, a UOX/MOX mixed core loading based on a generic PWR is modeled. A control rod ejection is calculated for two reactor states: Hot Zero Power (HZP) and 30% of nominal power. The worst cases for the rod ejection are determined by steady-state neutronic simulations taking into account the maximum reactivity insertion in the system and the power peaking factor. For the U/S analysis 378 uncertain parameters are identified and quantified (thermal-hydraulic initial and boundary conditions, input parameters and variations of the two-group cross sections). Results for uncertainty and sensitivity analysis are presented for safety important global and local parameters. (authors)

  2. Continuous non contacting control of the degree of admission of filler rods

    International Nuclear Information System (INIS)

    In laboratory tests a method was found to control continuously and non-contacting the degree of admission of filler rods. Behind the filling station the absorption of the ionizing radiation of a 90Sr beta source is measured. After successful tests with the laboratory equipment on the manufacturing machine of filler rods a prototype plant was constructed. The calibration is made by setting the measuring value of the empty filler rod equal to 0% and the measuring value of the optimum degree of admission equal to 100%. Between these two joints a scale is calculated so that to each measuring value a degree of admission can be assigned. The measuring time is 1 s. The limits of the allowable degrees of admission are freely adjustable. The construction of the plant is described. (authors)

  3. 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

  4. Summary of dynamic analyses of the advanced neutron source reactor inner control rods

    International Nuclear Information System (INIS)

    A summary of the structural dynamic analyses that were instrumental in providing design guidance to the Advanced Neutron source (ANS) inner control element system is presented in this report. The structural analyses and the functional constraints that required certain performance parameters were combined to shape and guide the design effort toward a prediction of successful and reliable control and scram operation to be provided by these inner control rods

  5. 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...

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

    International Nuclear Information System (INIS)

    Operating experience has proved so far that BWR control rods cannot be used for the total reactor life time as originally presumed, but instead has to be considered as a consumable article. After only few operating cycles, the mechanism of absorber failure has been shown to be neutron induced boron carbide swelling and stress cracking of the absorber tubes, followed by erosion of the absorber material. In the case that operation of such a control rod is continued in control cells, this can lead to an increase of the local power density distribution in the core and, under certain conditions, can even cause fuel rod damage. A non destructive testing method has been developed called 'UNDERWATER NEUTRON RADIOGRAPHY' applicable for any BWR control rod. 'Lead-control rods' being radiographed are used to evaluate their actual nuclear worth by the help of a special analytical procedure developed and verified by the author. Nuclear worth data plotted against bum up history data will allow to create an 'EMPIRIC MODEL'. This model includes the basic idea of operating control rods of a certain design first in a control position up to a target fluence limited to an amount just below the appearance of control rod washout. Afterwards they have to be moved in a shut down position to work therefor the total remaining holding period. The initial model is applicable to any CR-design as long as sufficient measuring-data and thus data about the nuclear worth are available. The results of these experiences are extrapolated to the whole reactor holding period. After modelling no further measurements of this particular control rod type are necessary in any reactor. The second focal point is to provide an APPROXIMATION EQUATION. By knowing the absorber radius, B4C density and absorber enclosure data an engineer will calculate reliably the working life of any control rod design on control position. indicated as maximum allowable neutron fluence margin until absorber wash-out starts. This

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

    International Nuclear Information System (INIS)

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

  8. 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)

  9. 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

  10. 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.

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    As proposed by the SPX-1 analysis task force, MSM (modified source multiplication) correction factors have been produced for a series of control rod configurations established in the first critical core C1D with minimum fissile loading and in the fully loaded core CMP. The report gives a complete description of the method used at KfK to produce these correction factors and summarises the evaluated experimental results obtained. The KfK method is characterized by a 'two-step-adjustment': A basic reactivity scale adjustment and a subsequent rod worth adjustment. The first adjustment was achieved by 'tuning' either the axialbuckling in the leakage term D B2 or the average number of neutrons per fission in the production term so that the excess reactivity of the so-called 'Follower-core' with all control rods fully raised was properly reproduced. As this excess reactivity could not be directly determined by an experiment, it had to be assessed from the shut-down worth of the main control system in combination with measured fractions of the S-curve of this system. In the second adjustment, the absorber cross sections were tuned to reproduce experimental rod worths. While for the analysis of the C1D experiments, MSM correction factor calculations were performed in 2D centre-plane geometry only, the analysis of the CMP measurements employed both 2D and 3D calculations. (orig./HP)

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

    Energy Technology Data Exchange (ETDEWEB)

    Eyler, J.H.

    1981-01-01

    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.

  14. 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

  15. 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.

  16. 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.

  17. Numerical calculation for flow field of servo-tube guided hydraulic control rod driving system

    International Nuclear Information System (INIS)

    A new-style hydraulic control rod driving mechanism was put forward by using servo-tube control elements for the design of control rod driving mechanism. The results of numerical simulation by CFD program Fluent for flow field of hydraulic driving cylinder indicate that the bigger the outer diameter of servo-tube, the smaller the resistance coefficient of variable throttle orifice. The zero position gap of variable throttle orifice could be determined on 0.2 mm in the design. The pressure difference between the upper and nether surfaces of piston was mainly created by the throttle function of fixed throttle orifice. It can be effectively controlled by changing the gap of variable throttle orifice. And the lift force of driving cylinder is able to meet the requirement on the design load. (authors)

  18. An improved method for calculating control rod reactivity worths in fast sodium cooled reactor cores

    International Nuclear Information System (INIS)

    An improved method is presented to determine the reactivities of strongly inhomogeneous control rod arrangements in fast sodium cooled reactor cores. The method is based on a detailed evaluation of the multiplication constants for the rods embedded in a large surrounding of fuel material. These calculations are performed using two-dimensional transport theory, with an accurate representation of the actual geometry in RΘ coordinates and with fine discretizations in coordinate space and energy. Three-dimensional whole core calculations are carried out in diffusion approximation, with a coarse spatial hexagonal-Z mesh and few energy groups, replacing the individual reactor cells by homogeneous arrangements. The homogenized macroscopic group cross sections are generated with standard methods, however using reduced boron contents of the absorber pins as compared with their actual values. The appropriate boron concentrations are found by comparing the control rod reactivity worths resulting from the two-dimensional transport calculations with those determined from corresponding diffusion calculations with homogenized compositions for the corresponding regions, which possess as many features of the final whole core calculations as possible. In this way, the corrections necessitated by the heterogeneity, transport, mesh, and condensation effects are incorporated in the macroscopic cross sections. With these as input, the computed rod worths of the secondary shutdown system of the SUPERPHENIX-1 (SPX-1) power production core are essentially improved as compared with results of earlier calculations. This progress of the calculational method is clearly demonstrated by a comparison with measured reactivity worths. (orig.)

  19. Study of two control rods of a district heating nuclear plant

    International Nuclear Information System (INIS)

    This paper broaches the study of the control rods to ensure a convenient working during load following of the nuclear reactor THERMOS. The mathematical model is descriptive of the whole of the nuclear plant (point model for the core and the heat balances). Two power control are studied. The first, like PWR, is a program for the mean temperature of primary water. The second takes into account the structure of the plant and is described by a schedule of powers

  20. 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)

  1. Numerical simulation of 900 MW control rods impact friction vibration and wear

    International Nuclear Information System (INIS)

    Impact-friction vibrations and wear have motivated a great research and development program aiming at understanding the impact and vibration behaviour of these components through experimental and numerical works. This report presents a numerical simulation of the vibrations of a single control rod and of a whole control cluster. Excitation sources for this component are due to hydraulic forces and are situated in the lower part of the rods and in the part of the cluster. Some parametric computations have been carried out on a single rod, to evaluate the effect of the lower excitation source. Different excitation levels, different eccentricities or static forces have been computed and compared to measurements on the MAGALY mock-up representing a complete rod cluster. A numerical model for the complete cluster allowed the evaluation of the upper excitation source effects. This source appears to be less powerful than the lower one. These results have been validated by comparison with MAGALY measurements. At last, some computations were performed with a model of the complete cluster, taking into account the both excitation sources. A parametric study on eccentricity and static forces has been carried out. A comparison with MAGALY measurements seems to be fairly fitting, showing that the numerical results are of the right order of magnitude. Through this numerical study, we have shown that numerical simulation of a complete control rod cluster could be lead, and we have obtained some new informations about impact forces and wear rates that need to be confirmed by more computational or experimental works or in-situ measurements. (author). 10 annexes, 11 refs

  2. 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.

  3. 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

  4. Flow mixing inside a control-rod guide tube – Experimental tests and CFD simulations

    International Nuclear Information System (INIS)

    This paper covers a combined experimental and computational effort carried out at Vattenfall Research and Development AB in order to study the thermal mixing in the annular region between a top tube and a control-rod stem. The low frequency thermal fluctuations in this region can result in problems with thermal fatigue and have caused cracks in the control-rod stems of several nuclear reactors (). The flow in the vertical annular region formed by the top tube and the control-rod stem is characterized by the mixing of hot bypass flow with cold crud-removal flow. The crud-removal flow is flowing upwards along the control-rod stem, and the warmer bypass flow is entering through eight horizontal holes positioned in the lower part of the guide tube and four holes in the upper part of the top tube, forming jets. Two full-scale models of a control rod, including the control-rod stem and the guide tube, were constructed. The first model, designed to work at atmospheric conditions, was made of Plexiglass, in order to be able to visualize the mixing process, whereas the second one was made of steel to allow for a higher temperature difference between the two flows, and the heating of the top tube. CFD simulations of the case at atmospheric conditions were also carried out. Both the experiments and the simulations showed that the mixing region between the cold crud-removal flow and the warm bypass flow is dominated by large flow structures coming from above. The process is characterized by low frequency, high amplitude temperature fluctuations. The process is basically hydrodynamic, caused by the downward transport of flow structures originated at the upper bypass inlets. The damping thermal effects through buoyancy is of secondary importance, as also the scaling analysis shows, however a slight damping of the temperature fluctuations can be seen due to natural convection due to a pre-heating of the cold crud-removal flow. The comparison between numerical and experimental

  5. 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.

  6. 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

  7. Application of pattern recognition techniques to the detection of the Phenix reactor control rods vibrations

    International Nuclear Information System (INIS)

    The incipient detection of control rods vibrations is very important for the safety of the operating plants. This detection can be achieved by an analysis of the peaks of the power spectrum density of the neutron noise. Pattern Recognition techniques were applied to detect the rod vibrations which occured at the fast breeder Phenix (250MWe). In the first part we give a description of the basic pattern which is used to characterize the behavior of the plant. The pattern is considered as column vector in n dimensional Euclidian space where the components are the samples of the power spectral density of the neutron noise. In the second part, a recursive learning procedure of the normal patterns which provides the mean and the variance of the estimates is described. In the third part the classification problem has been framed in terms of a partitioning procedure in n dimensional space which encloses regions corresponding to normal operations. This pattern recognition scheme was applied to the detection of rod vibrations with neutron data collected at the Phenix site before and after occurence of the vibrations. The analysis was carried out with a 42-dimensional measurement space. The learned pattern was estimated with 150 measurement vectors which correspond to the period without vibrations. The efficiency of the surveillance scheme is then demonstrated by processing separately 119 measurement vectors recorded during the rod vibration period

  8. 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

  9. 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.)

  10. 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.

  11. Postirradiation examination of JOYO MK-II control rods. Irradiation performance of absorber pins

    International Nuclear Information System (INIS)

    Postirradiation examinations of JOYO MK-II control rods have been carried out since 1983, where 16 subassemblies with total 110 absorber pins of initial load to the fifth reload control rods have been subjected to a number of both non-destructive and destructive examinations. In the course of postirradiation examinations, a cracking of cladding tube was found in the total 15 absorber pins in five control assemblies. This paper indicates the results of postirradiation examinations and analysis of absorber pin performance using CORAL code to elucidate the cause of absorber pin cracking in JOYO MK-II control rods. The cause of cladding failure was attributed to the ACMI where the gap closure due to relocation of B4C pellet took place from early times of irradiation. The code analysis by CORAL indicated that the cladding strain due to ACMI was not fully absorbed by the irradiation creep and that the plastic strain became large enough to make a fracture of absorber pins with an increasing burnup. (J.P.N.)

  12. Control rod position and temperature coefficients in HTTR power-rise tests. Interim report

    International Nuclear Information System (INIS)

    Power-rise tests of the High Temperature Engineering Test Reactor (HTTR) have been carried out aiming to achieve 100% power. So far, 50% of power operation and many tests have been carried out. In the HTTR, temperature change in core is so large to achieve the outlet coolant temperature of 950degC. To improve the calculation accuracy of the HTTR reactor physics characteristics, control rod positions at criticality and temperature coefficients were measured at each step to achieve 50% power level. The calculations were carried out using Monte Carlo code and diffusion theory with temperature distributions in the core obtained by reciprocal calculation of thermo-hydraulic code and diffusion theory. Control rod positions and temperature coefficients were calculated by diffusion theory and Monte Carlo method. The test results were compared to calculation results. The control rod positions at criticality showed good agreement with calculation results by Monte Carlo method with error of 50 mm. The control position at criticality at 100% was predicted around 2900mm. Temperature coefficients showed good agreement with calculation results by diffusion theory. The improvement of calculation will be carried out comparing the measured results up to 100% power level. (author)

  13. 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.

  14. A rule-based expert system for automatic control rod pattern generation for boiling water reactors

    International Nuclear Information System (INIS)

    This paper reports on an expert system for generating control rod patterns that has been developed. The knowledge is transformed into IF-THEN rules. The inference engine uses the Rete pattern matching algorithm to match facts, and rule premises and conflict resolution strategies to make the system function intelligently. A forward-chaining mechanism is adopted in the inference engine. The system is implemented in the Common Lisp programming language. The three-dimensional core simulation model performs the core status and burnup calculations. The system is successfully demonstrated by generating control rod programming for the 2894-MW (thermal) Kuosheng nuclear power plant in Taiwan. The computing time is tremendously reduced compared to programs using mathematical methods

  15. VVER-440 control rod follower induced local power peaking computational benchmark: MCNP and Karate solutions - 082

    International Nuclear Information System (INIS)

    With the original VVER-440 follower design the relatively large amount of water in the coupler between the absorber and fuel part of the control assembly can cause sharp power peaking in the fuel rods next to the coupler. The power peaking can be especially high after control rod withdrawal when the coupler reaches a low burnup level region of the adjacent assembly. Though the modernized coupler has a Hf plate in the critical region to suppress the power peak, the complicated structure needs a reference Monte Carlo calculation as a basis of engineering code validation. The coupler mathematical benchmark was solved by the KARATE code system using the same methods and approximations as in case of NPP applications and the results were compared to that of the reference MCNP. The need for treating the Hf burnout in the reflector region was also investigated. (authors)

  16. 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...

  17. 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)

  18. 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

  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. 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.

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

    International Nuclear Information System (INIS)

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

  4. Optimization of RIA-calculations : Simulating Falling Control Rods at Forsmark Nuclear Power Plant

    OpenAIRE

    Alex, Christian

    2013-01-01

    This report accounts for investigations of ways to reduce the calculation times forsimulations of falling control rods in boiling water reactors done prior to everyreactor startup, known as RIA-calculations. Two methodologies to lower thecalculation times have been proposed, developed and implemented in a set ofmatlab-scripts, which are fully compatible with the previously used methodology.The new methodologies have been applied on 17 authentic power cycles at the threeForsmark reactors, wher...

  5. 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

  6. SPND detectors response at the control rod drop in WWER-1000. Measurement and modelling results

    International Nuclear Information System (INIS)

    The paper analyzes and discusses possibility of neutron flux inspection in the WWER core during fast dynamic processes applying existing in-core monitoring system. The structure and functions of the system, basic principal of detector functioning and its temporal parameters are described briefly. To assess the ability of such dynamic monitoring the event with control rod drop happened during operation of Kozloduy NPP Unit 5 is observed - at the level of power close to nominal one of the rod from control group shifted to the lowest position at-2 seconds. In-core detectors readings at the process were registered and processed with mathematical methods that allow to single out only the prompt part of the signal. Results of the processing are presented. Furthermore, the process observing have been modeled with 3D dynamic code NOSTRA. Results of modeling are presenting in a paper, and comparing with experimental ones. A good agreement achieved. The analysis of measurements and its imitation give a hope that with an aggregate signal of detectors the measurement of control rod worth could be provided, and it allows to avoid of influence of spatial effects that are significant at standard technique with ex-core ion chambers (Authors)

  7. Factors influencing helium measurements for detection of control rod failures in BWR

    International Nuclear Information System (INIS)

    Much effort has been made to minimize the number and consequences of fuel failures at nuclear power plants. The consequences of control rod failures have also gained an increased attention. In this paper we introduce a system for on-line surveillance of control rod integrity which has several advantages comparing to the surveillance methods available today in boiling water reactors (BWRs). This system measures the helium released from failed control rods containing boron carbide (B4C). However, there are a number of factors that might influence measurements, which have to be taken into consideration when evaluating the measured data. These factors can be separated into two groups: 1) local adjustments, made on the sampling line connecting the detector to the off-gas system, and 2) plant operational parameters. The adjustments of the sample line conditions include variation of gas flow rate and gas pressure in the line. Plant operational factors that may influence helium measurements can vary from plant to plant. The factors studied at Leibstadt nuclear power plant (KKL) were helium impurities in injected hydrogen gas, variation of the total off-gas flow and regular water refill. In this paper we discuss these factors and their significance and present experimental results of measurements at KKL. (authors)

  8. On-line fuel and control rod integrity surveillance in BWRs

    International Nuclear Information System (INIS)

    Surveillance of fuel and control rod integrity in a BWR core is essential to maintain a safe and reliable operation of the nuclear power plant. Any actions to be taken in the event of a fuel failure during reactor operation should be based on the best available information regarding the failure and expected consequences. The detection of fuel and control rod failures in BWRs is usually performed by analyzing samples of off-gases and coolant taken with a certain time intervals, e.g. once a week or once a month. This procedure can, however, leave the failure undetected in the core for quite some time. Therefore, a sufficient improvement of the surveillance of fuel and control rods can be achieved by simultaneous measurements of He and gamma emitting noble gases on-line in the off gas system. In this paper, experiences of such measurements performed at Kernkraftwerk Leibstadt (KKL) in Switzerland and Forsmark nuclear power plant (NPP) in Sweden will be presented. (author)

  9. 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

  10. 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.

  11. 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)

  12. 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

  13. Experimental study on a heat pipe towards in-core decay heat removal control rod

    International Nuclear Information System (INIS)

    A novel in-core heat removal concept can be adopted in control rods for passive safety of nuclear power plants. The new concept is featured by a passive heat transfer device called heat pipe and combined with control rod. As the first step for this concept, stainless steel 316L heat pipes were tested in terms of heat removal capability under the same diameter condition with an actual control rod in a typical PWR. It has outer diameter of 3/4 inch (17.4 mm inner diameter), and the length of 1000 mm. Also, the capillary-driven heat pipe was compared with a bare tube with same diameter without wick structures called thermosyphon. As the results, heat transfer coefficients of the heat pipe were ∼34% higher than those of thermosyphon. The results were compared with existing correlations and a CFD analysis. The overall heat transfer characteristics of heat pipes such as thermal resistances were checked for potential uses in terms of in-core heat removal. (author)

  14. A power control system for the rod drive coil of control element drive mechanism in pressurized water reactor

    International Nuclear Information System (INIS)

    In this paper, we propose a new type of power control system for the rod drive coil of the CEDM of the PWR NPP in order to supply more reliable DC power. The electrical modelling of the controlled rod drive coil was done by referring related documentations. The design of the proposed system is based on this electrical model satisfying the existing specification. A high power DC-DC converter scheme is adopted utilizing the SMPS technique in the design of the proposed system. In order to show the effectiveness of the proposed system, an experimental system with the capability of 3.2 K Watt was set up for a rod with four cores and some computer simulations and experimentations were carried out. The result shows a very similar tracking performance with that of the existing system to the driving command. As a result of this, the proposed method can be applied to the power control system for the rod drive coil of the CEDM of the PWR NPP. (Author). 8 refs., 1 tab., 10 figs

  15. 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

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

    International Nuclear Information System (INIS)

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

  17. 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

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

    International Nuclear Information System (INIS)

    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

  19. 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.

  20. Control of the Peak Linear Power by Using Two Kinds of Fuel Rods in the AHR

    International Nuclear Information System (INIS)

    Korea Atomic Energy Research Institute (KAERI) is developing an Advanced HANARO research Reactor (AHR) based on the HANARO experiences through its design to operation stages. AHR will be a 20 MW multi purpose research reactor and loaded with the HANARO fuel assemblies of a rod type. AHR has a compact core with a high power density for achieving a high neutron flux that is most important in a research reactor. As the average power is high, the control of the peak linear power is very important. The reference core of the AHR shows an acceptable peak linear power in the fresh core. In the equilibrium core, the peak linear power had been assumed to be low. A recent evaluation shows that the peak linear power in the equilibrium core exceeds the target limit. The evaluation was performed with the HELIOS / VENTURE code system and is suspected to be overestimated when compared with the result by the MCNP code in the fresh core. The modeling and fuel management scheme will be improved. Fundamentally, measures to lower the high peak linear power should be prepared. HANARO uses two kinds of fuel rods for reducing the peak linear power. It is expected that the same method will work well in the AHR. This paper introduces measures to control the peak linear power including the adaptation of two kinds of fuel rods and evaluates the peak linear power for the equilibrium core using a Monte Carlo burn-up system

  1. Study of the WWR-K reactor automatic control rod materials microstructure

    International Nuclear Information System (INIS)

    The results of structure examination for the spent automatic control rod (ACR) material of the WWR-K reactor are presented. The rod was manufactured from both aluminium alloy SAV-1 and Cr18Ni10Ti austenitic stainless steel. The materials were irradiated by the thermal neutrons up to the dose 1.32·1021 neutron/cm2 at 70 deg.C temperature. Examination at the scanning electron microscope for the ACR components from aluminium alloy is showing the presence in their surface the cracks, forming a some grain structure with partial or full near-surface layer peeling. Transmission electron microscopy of these samples microstructure shows, that reactor irradiation leads to the material swelling (about 0.3%) in the result of the small vacancy pores formation with a high density

  2. Examination of control rod ejection in WWER-440 type reactors at different circumstances using the code DYN3D

    International Nuclear Information System (INIS)

    For nuclear reactors it is very important to examine the reactivity initiated transients caused by the ejection of a control rod. The event is found to be dependent on different thermal and neutronic parameters. In this paper the emphasis is laid on the effect of the power level at which the transient began and on the effect of the heat transfer coefficient measured in the gap between the fuel and the cladding. The most significant transients can be established by the ejection of the most effective control rod. So the first step is to determine the position of this rod. It was done by steady state calculations A calculation was carried out with all the rods inserted to the half level of the core, criticality was reached by adjusting the power level. Seven other calculations were made for each control rod at withdrawn position while the other six rods were inserted to the half plane of the core. From the results the most effective control rod could be determined.(authors)

  3. 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).

  4. 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)

  5. Parametric study of a reactivity accident in a pressurized water reactor: control rod cluster ejection

    International Nuclear Information System (INIS)

    This research thesis concerns a class 4 accident in a PWR: the ejection of a control rod cluster from the reactor core. It aims at defining, for such an accident, the envelope values which relate the reactivity to the hot spot factor within the frame of a mode A control. The report describes the physical phenomena and their modelling during the considered transient. It presents a simple mathematical solution of the accident which shows that the main neutron parameters are the released reactivity, the delayed neutron fraction, the Doppler coefficient, and the hot spot factor. It reports a temperature sensitivity study, and discusses three-dimensional calculations of irradiation distributions

  6. On the Rod Drop technique in integral reactivity measures in control banks and reactor safety

    International Nuclear Information System (INIS)

    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)

  7. Measurements of negative reactivity in Masurca and Phenix control rods: Prospects for Superphenix

    International Nuclear Information System (INIS)

    Experimental assessment of the negative reactivity of the control rods in an industrial reactor has recently been the subject of numerous studies conducted in the light of forthcoming startup tests on the core of Superphenix. Representative tests have been carried out both on Phenix and on the Masurca critical mockup, and a test programme for Superphenix has been drawn up. Subcritical measurements (source multiplication technique) have been carried out on Phenix without absolute measurement of a standard. However, a precise relative interpretation using two counters demonstrates good agreement following the correction of spatial effects. The chief value of the rod drop measurements conducted on Masurca was that it provided a means of cross-checking the kinetic method to be validated against a standard source multiplication method. The results demonstrate complete agreement between the two methods. The acceptability of the rod drop method is therefore considered to be established. The programme foreseen for startup of Superphenix and the objectives which have been set are briefly indicated. The calculation methods to be used in respect of the startup tests have been established on the basis of experience gained through interpreting the experiments conducted in the course of the Racine (Masurca) programme. An analysis of these experiments included, among other things, a parametric study that has made it possible to devise a standard calculation method for predicting Superphenix rod worth values. The main feature is a scattering calculation with three energy groups and three dimensions. Two-dimensional scattering and transport calculations are therefore necessary in order to define the corrective factors to be applied to this initial result. The final result of this analysis is thus made equivalent to a 25-energy-group transport calculation with an extremely small spatial mesh

  8. 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.

  9. A benchmark for investigating the radial dependence of C/E for control rod worths in large decoupled cores

    International Nuclear Information System (INIS)

    The first physics measurements of a heterogeneous core on the critical assembly ZPPR-7 at ANL showed that the C/E ratios with ENDF/B data for the worths of the control rods in the outer bank were several percent higher than those at the inner bank positions. This radial variation in the C/E for the rod worths was further confirmed in the analysis of the large conventional core ZPPR-10, and again in the analysis of the large heterogeneous core series ZPPR-13. In the design of a power reactor, the number of control rods, and their disposition, are determined by calculations. Misprediction of the worth of the control rods can lead to serious economic penalties by restricting the operation of the core. Retrofitting a core to accommodate more worth will be costly and is likely to lead to a non-optimized core. This document provides a discussion of these calculations. 7 refs., 7 figs., 12 tabs

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

    International Nuclear Information System (INIS)

    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

  11. 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.

  12. Influence of control rod enhanced expansion devices on the course of unprotected transients in the EFR

    International Nuclear Information System (INIS)

    In the safety analysis of fast reactors, unprotected accidents, such as ULOF and UTOP have to be considered, even when their frequency of occurrence lies far beyond the design basis accident. In the European Fast Reactor (EFR), the safety approach foresees further measures of risk minimization in the frame of the so-called Third Shutdown Level. One of the measures is a control rod enhanced expansion device, called ATHENa, which has been developed by KfK in collaboration with SIEMENS as a passive device to separate the absorbers from the drive lines in cases of accidental coolant temperature rises and to force the absorbers further into the core in case of failure to drop. The efficiency of the ATHENa devices to prevent soduim boiling and fuel melting in unprotected accidents in EFR has been investigated by calculations with the dynamics code DYANA2. In the case of ULOF accidents, sodium boiling can be prevented, if at least one out of 24 absorber rods equipped with ATHENa devices drops into the core after delatching from the drive lines. In the extremely remote case that all rods remain jammed after delatching, they are pushed by the ATHENa decices into the core with an enhanced expansion coefficient (∼ 10 times). Even then, sodium boiling could be prevented by extending of the pump coast down halving time from 10 to 12 s or by adjusting the delatching temperature to a value not higher than about 40 C above nominal coolant outlet. In UTOP accidents caused by the uncontrolled withdrawal of a control rod, the main concern is incipient fuel melting. The results of the calculations have shown that the power rise can be terminated by delatching the absorbers, before fuel melting occurs, if the ramp rate is mechanically limited to values of 1 /s or less. Again, even in the worst case that all rods remain jammed, fuel melting could be prevented by adjusting the delatching temperature to a similar value as in the ULOF case. (orig.)

  13. Development of carbon/carbon composite control rod for HTTR. 2. Concept, specifications and mechanical test of materials

    International Nuclear Information System (INIS)

    A concept and specifications of carbon/carbon composite (C/C) control rod were proposed, aiming at the application of the material to the HTTR. The outer diameter and length of the control rod were kept as the same as those of the present control rod, i.e., 113 mm and 3094 mm, respectively. According to the concept, the rod consists of ten units which are connected in series using bolts. Then, the stresses generated by dead loads in the control rod elements were estimated and compared with the design strengths which were derived from the results of measurements of tensile, compressive, bending and shear strengths of two candidate materials, AC250 (Across Co.) and CX-270 (Toyo Tanso Co.). Design strength was preliminarily determined as one-third or one-fifth of the mean strength. Ratio of the design strength to generated stress for the AC250 (2D) was : Tensile stress in the outer sleeve tube, 66, tensile and shear stresses in the M16 bolt, 8.8 and 8.5, shear stress in the plug support bolt M8, 2.43. These results are believed to indicate the mechanical integrity of the control rod structure. Data available on the candidate materials were also compiled in the Appendix. (author)

  14. 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)

  15. 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.

  16. 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

  17. Utilization of control rod drive (CRD) system for long term core cooling

    International Nuclear Information System (INIS)

    In this paper we consider an application of Probabilistic Risk Assessment (PRA) to risk management. Foreseeable risk management strategies to prevent core damage are constrained by the availability of first line systems as well as support systems. The actual trend in the evaluation of risk management options can be performed in a number of ways. An example is the identification of back-up systems which could be used to perform the same safety functions. In this work we deal with the evaluation of the feasibility, for BWR's, to use the Control Rod Drive system to maintain an adequate reactor core long term cooling in some accident sequences. This preliminary evaluation is carried out as a part of the Internal Events Analysis for Laguna Verde Nuclear Power Plant (LVNPP) that is currently under way by the Mexican Nuclear Regulatory Body. This analysis addresses the evaluation and incorporation of all the systems, including the safety related and the back-up non safety related systems, that are available for the operator in order to prevent core damage. As a part of this analysis the containment venting capability is also evaluated as a back-up of the containment heat removal function. This will prevent the primary containment overpressurization and loss of certain core cooling systems. A selection of accident sequences in which the Control Rod Drive system could be used to mitigate the accident and prevent core damage are discussed. A personal computer transient analysis code is used to carry out thermohydraulic simulations in order to evaluate the Control Rod Drive system performance, the corresponding results are presented. Finally, some preliminary conclusions are drawn. (author). 9 refs, 5 figs

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

    International Nuclear Information System (INIS)

    The aim of this work was to establish the 10B depletion model for CR 99 control rods by using the latest version of POLCA7. In order to obtain an understanding of the differences between the currently used 10B depletion models implemented in POLCA4 at O3 and in SIMULATE-3 at OL1, and the latest improved model implemented in the latest POLCA7, this work has been performed in three different parts. The first part of the work was to find out how large differences there exist in 10B depletion between the calculated data by using the latest core monitoring system (POLCA7 version 4.10.0) and the measured data obtained in the hot-cell laboratory in Studsvik. It was found that the 10B depletion computed by the latest POLCA7 version is in good agreement with the measured data from Studsvik. A poor agreement with a conservative overestimation in 10B depletion was also found between the old model and the measured data. The aim of the second part of the work was to compare the calculated 10B depletion values for two CR 99 rods from Olkiluoto 1 with the calculated 10B depletion value for a CR 99 rod from Oskarshamn 3, by using the new 10B depletion model implemented in the latest POLCA7 version. Swelling measurements of the boron carbide pins, used as absorber material, have indicated that the 10B depletion should be of similar magnitude for the rods in Olkiluoto 1 and the rod in Oskarshamn 3, whereas the calculated values by using the earlier 10B depletion models on the process computers showed a difference of about 20 %. By using the new 10B depletion model m POLCA7, it was found that the 10B depletion in the two studied cases was similar to each other and, thus, the hypothesis of a linear relationship between B4C swelling and thermal neutron fluence was supported. This third part of the work was carried out at KKL, Switzerland, and focused on comparing the B depletion models used in Westinghouse/POLCA7 and KKL/PRESTO-2. It was found that there is a slight difference in the

  19. 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

  20. Experimental and computer analyses of control rod drive systems seismic capacity

    International Nuclear Information System (INIS)

    The experimental and computer analyses of the 1/4 scale Control Rod Drive System (CRDS) model of WWER-440 reactor has been carried out. The experimental study has been undertaken on CVS 20 ton's capacity shaking table with modeling operability of CRDS during earthquake and operational vibration. A special PC computer program has been developed for evaluation of CRDS seismic and vibration margins. The program enables estimation of different nonlinear effects in bearings and gaps of CRDS including shocks and friction that highly influence on dynamic capacity of CRDS. The results of these investigations are presented in this paper. (author)

  1. 'DPS-1 SKODA' diagnostic system for the reactor control rod drives functional and lifetime tests

    International Nuclear Information System (INIS)

    The 'DSP-J SKODA' diagnostic system of the reactor control rod drives (VVER-440, 213 type) is described in this paper. The hardware structure, methods and utility software of the diagnostic system is explained. The main goal of this system is defined: to ensure the functional availability and longer lifetime of modernized drives (15 to 20 years). Experiences from the measurements, evaluation and analysis with the 'DSP-1 SKODA' system in die testing room in SKODA - Bolevec are introduced. The results of functional and lifetime tests of prototype drive reductors are presented. (author)

  2. The application for examination technology of rod cluster control assembly in nuclear power plant

    International Nuclear Information System (INIS)

    During nuclear power plant operation, three typical defects may be generated in the Rod Cluster Control Assembly (RCCA). Through operation situation of RCCA, this paper describes and analyzes three kinds of the reasons leading to the typical defects generated in RCCA. The theory of the special examination technique including ultrasonic and eddy current examination methods are introduced in this paper. According to examination practice of RCCA in nuclear power plant, the analysis methods for main defects and the examination results are concluded. This RCCA examination technology will be provided for the reference experience of the future RCCA examination. (authors)

  3. 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

  4. The Racine-1e critical experiments for control-rod method and data validation experimental and calculation results

    International Nuclear Information System (INIS)

    This paper summarizes the results of control rod experiments, performed in the heterogeneous double ring core RACINE-1E, investigated as part of the RACINE programme at the French zero-power-facility MASURCA. Measurements were made using both natural and highly enriched boron carbide adsorbers and comprised subcritical and critical experiments. The analysis was carried out using methods similar to those used by CEA for the prediction of rod worths in LMFBR power reactors

  5. 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)

  6. 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)

  7. 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

  8. 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

  9. 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.

  10. 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

  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. Adjustment method of deterministic control rods worth computation based on measurements and auxiliary Monte Carlo runs

    International Nuclear Information System (INIS)

    Highlights: • 3-group cross sections is collapsed by WIMS and SN2. Core is calculated by CITATION. • Engineering adjustments are made to generate better few group cross-sections. • Validation is made by JRR-3M measurements and Monte Carlo simulation. - Abstract: The control rods (CRs) worth is key parameter for the research reactors (RRs) operation and utilization. Control rods worth computation is a challenge for the full deterministic calculation methodology, including the few group cross section generation, and the core analysis. The purpose of this work is to interpret our codes system, and their applicability of obtaining reliable CRs worth by some engineering adjustments. Cross sections collapsing in three energy groups is made by WIMS and SN2 codes, while the core analysis is performed by CITATION. We use these codes for the design, construction, and operation of our research reactor CMRR (China Mianyang Research Reactor). However, due to the intrinsic deficiency of the diffusion theory and homogenizing approximation, the directly obtained results, such as CRs worth and neutron flux distributions are not satisfactory. So two points of simple adjustments are made to generate the few group cross-sections with the assistance of measurements and auxiliary Monte Carlo runs. The first step is to adjust the fuel cross sections by changing properly the mass of a non-fissile material, such as the mass of the two 0.4 mm Cd wires existing at both sides of each uranium plate, so that the core model of CITATION can get good eigenvalue when all CRs are completely extracted. The second step is to revise the shim absorber cross section of CRs by adjusting the hafnium mass, so that the CITATION model can get correct critical rods position. In this manuscript, the JRR-3M (Japan Research Reactor No. 3 Modified) reactor is employed as a demonstration. Final revised results are validated with the stochastic simulation and experimental measurement values, including the

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

    Directory of Open Access Journals (Sweden)

    M. Vaisnoras

    2008-05-01

    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.

  14. 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)

  15. Design of the control rod system for the 2400 MWTH generation IV gas-cooled fast reactor

    International Nuclear Information System (INIS)

    The present paper is related to the design of the principal control rod system for the reference large (2400 MWth) Generation IV Gas-cooled Fast Reactor (GFR), which makes use of CERCER plate-type fuel-assemblies with carbide fuel contained within a SiC inert matrix. For the calculations, the deterministic code system ERANOS-2.0 has been used in association with a RZ core model, with particular attention given to the heat generation within the control rods. Based upon the results obtained, a heterogeneous control rod pattern has been developed taking into account both neutronic and thermal-hydraulic constraints. Thus, the rod temperature was computed using the thermal-hydraulic code COPERNIC, for different helium flow rates and heat transfer correlations. It is found that it is necessary to dedicate a significant coolant mass flow rate (∼2.5% of the total for the core) to maintain acceptable cladding temperatures. Then, the control rod worths were computed by applying a methodology based on reactivity equivalence, in order to account fully for the heterogeneity of the control assembly design. Furthermore, the heterogeneity effects between the absorber pins leads to some rather large rod worth reductions in large sodium cooled fast reactors (∼20% in the case of Superphenix, for example). Therefore the design was set up in a manner to minimize this shadowing effect and as a consequence, the explicit consideration of the heterogeneity of the new design leads to a significantly lower rod worth reduction of ∼13%, as compared to the result obtained with the homogeneous model. (author)

  16. Calculational methods, codes and results of calculational and experimental investigations of control rod worth in power fast reactors

    International Nuclear Information System (INIS)

    The paper aims to present the main physical principles for selection of design characteristics of the fast reactor control rods (CR) system. The brief analysis of problems of CR physical calculations is given. Four components are described for the correction to the control rod worth calculated by the routine method based on the few - group three - dimensional diffusion code (TRIGEX) in hexagonal geometry. Principle considerations are given for the choice of the original task discretization methods implemented in this code to minimize the total error. Brief information is given about methods and codes used for the evaluation of error components of control rod worths calculated in a standard way. The results of experimental and calculational investigations of control rod physical characteristics are presented. These results were obtained at BFS critical assemblies simulating LMFBR cores. The investigations have been carried out for different types of core configurations. The experimental and calculated values are given on the distortion of power distribution due to the control rod insertion in the core. (author). 51 refs, 9 figs, 5 tabs

  17. Fabrication of UO2 fuels with non-destructive assay and rod scanner technology for production and final quality control

    International Nuclear Information System (INIS)

    The production of nuclear fuels requires a continuous in-process inspection which includes the control of equipment and process parameters as well as the quality control of the material. All components and intermediate products belonging to fuel fabrication, or being produced throughout the process, are examined by physical and chemical tests. The supervision of important process and quality parameters was improved by new techniques and instrumentations of nondestructive in-line and laboratory assays. For example the 235U enrichment is supervised by the UF6-incoming control up to a continuously working in-line instrumentation, which measures the 215U concentration in the feed hoppers before pressing. For final control of the fuel rods, an active rod scanner, which examines the rod's structure and composition, the 235U enrichment, and pellet mix-up, is used. (orig.)

  18. Controllable end shape modification of ZnO nano-arrays/rods by a simple wet chemical etching technique

    International Nuclear Information System (INIS)

    The well-aligned ZnO nano-arrays/rods synthesized by a chemical bath deposition method on a highly conductive Si substrate were chemically etched in an ammonia chloride aqueous solution. An obvious end shape modification of ZnO nano-arrays/rods was realized in this report. The hexagonal frustum end of ZnO nano-arrays/rods changed into a pyramid and the diameter of ZnO nano-arrays/rods decreased gradually with the increasing etching time. The evolution mechanism of the wet etching process was discussed based on a proposed evolution model. Photoluminescence measurements indicated that the near band edge emissions of ZnO nano-arrays/rods increased greatly after wet etching. The controllable end shape modification of ZnO nano-arrays/rods on a highly conductive Si substrate by this simple wet etching technique will further explore the application of ZnO in field emission devices and 1D based nano-devices with various end shapes. (paper)

  19. 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)

  20. Qualification test of the EPR control rod drive mechanism in the full scale component test facility KOPRA

    International Nuclear Information System (INIS)

    The control rod drive mechanism (CRDM) and the mobile set consisting of rod cluster control assembly (RCC-A) of the evolutionary power reactor (EPR) had to pass a full scale qualification test in representative site conditions. The KOPRA core test section in Erlangen is precisely designed for full scale tests on nuclear core components in respect to coolant temperature and volume flow of PWR site conditions. In the test channel the complete geometry of the central core position of the reactor pressure vessel is simulated with 1:1 scale. The performance test program has led to an optimized test sequence through small adjustments in operating parameters of CRDM. The endurance test program has demonstrated that all tested components, i.e. the CRDN, the control rod driveline and the components of the drop channel are able to function properly and to meet the specification goals.

  1. MIC program for calculational simulation of experiments on measuring the control rod efficiency by the method of interference corrections

    International Nuclear Information System (INIS)

    The MIC program aimed for methematical simulation of experiments on measuring reactor control system absorbing rod efficiency and obtained data processing is described. The method of interference corrections used for calculations permits on the base of known values of single control rod efficiency and their mutual influence to determine the efficiency of the system in a whole. The values of the interference corrections for the MIC program are calculated by means of the PNK heterogeneous program. The program is written in FORTRAN for the BESM-6 computers

  2. 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)

  3. 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.

  4. 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

  5. High-temperature oxidation of graphite rods with temperature control by combustion gas recycle

    International Nuclear Information System (INIS)

    The combustion of graphite (fuel blocks) is of fundamental importance in the fuel reprocessing scheme for the High-Temperature Gas-Cooled Reactor (HTGR). A study was made to evaluate a chunk-type burner for possible application in this reprocessing step. The combustion gases were recycled to allow operation at higher burn rates without an increase in graphite temperature. Graphite rods of two diameters were oxidized with makeup oxygen and recycled stack gases at various gas flow rates in an insulated reactor. Results of this study indicate a strong dependence of oxygen transfer on gas flow rate with little effect resulting from changes in graphite temperature. High carbon monoxide concentrations in the exit gas were not a problem except at oxygen concentrations below approx. 5%. Stable operation of a recycle controlled burner was achieved, avoiding the temperature excursions common in previous graphite burners

  6. Implementation of reactor control rod position sensing/display in a CPLD

    International Nuclear Information System (INIS)

    The design, simulation, implementation and test of a prototype of electronic unit intended as a replacement of the outdated russian type used presently for reactor control rod position sensing/display in old power reactors is presented. The implementation involves both analog and digital design. The designed digital circuit has 12 TTL outputs working in a 1-out-of-12 mode, excluding both double (2-out-of-12) and no-output state. The circuit mainly consists of combinatorial logic. MACH 111SP CPLD from VANTIS (AMD) is selected as implementation technology for logic part of the circuit. Analog comparators are used in the analog part. To avoid a flickering display during transition between two neighboring positions, sort of hysteresys is implemented. The cir circuit is prepared to handle slow excitation voltage changes in the +/- 10 percent range, without the need for readjustment of comparator levels. The circuit is implemented and tested as a prototype

  7. 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)

  8. Fuel pin behaviour under conditions of control rod withdrawal accident in CABRI-2 experiments

    International Nuclear Information System (INIS)

    Simulation of the control rod withdrawal accident has been performed in the international CABRI-2 experimental programme. The tests realized with industrial pins led to clarification of the influence of the pellet design and have shown the important role of fission products on the solid fuel swelling which promotes early pin failure with solid fuel pellet. With annular pellet design, large fuel swelling combined to low smear density leads to degradation of fuel thermal conductivity and thus reduces power to melt. However, the high margin to deterministic failure is confirmed with hollow pellets. Improvements of the modelling were necessary to describe such behaviours in computer codes as SAS-4A, PAPAS-2S and PHYSURAC. (author)

  9. Uncertainty analysis for control rod ejection accidents simulated by KIKO3D/TRABCO code system

    International Nuclear Information System (INIS)

    Recently, considerable conservatism must be applied in the traditional safety analyses for taking into account the uncertainties originating from the input parameters, approximations in the models, due to the safety reserves, etc. The extreme values for all of the input parameters are supposed in the traditional safety analysis at the same time. Additionally it must be mentioned that the selection of the input parameter values leading to conservative results often is not easy. The main goal of this paper is to present a more realistic methodology for the case of control rod ejection accidents. The applied consistent statistical approach leads to conservative results also, but avoids the unnecessary cumulative conservatism. A method based on a mathematical model ('Two-Sided Statistical Tolerance Intervals', [1-2]) was chosen for the realization of uncertainty analyses of Reactivity Initiated Accidents (RIA). (author)

  10. Mathematical formulation of temperature fluctuation and control rod vibration in PARR

    International Nuclear Information System (INIS)

    This report describes the mathematical interpretation of experimental neutron noise spectra obtained for PARR core. A one dimensional thermal-hydraulic model of PARR core was developed to calculate the magnitude of neutron noise as a result of fluctuation in the core inlet coolant temperature. The sink structure of the neutron power spectral density as well as the dependence of observed neutron spectra on coolant velocity is also explained by the thermal hydraulic model. An attempt is made to explain the phenomena of control rod vibration by a simple eigen frequency vibration model. The calculated neutron power spectral density due to vibration and temperature noise were added and compared with the experimental power spectra obtained for PARR. (orig./A.B.)

  11. Seismic calculation of Superphenix control rod and drive mechanism, comparison with test results

    International Nuclear Information System (INIS)

    In case of a seismic event the insertability of Superphenix primary shutdown system: SCP (systeme de commande principal) has to be demonstrated. As there was no existing facility in France to test this kind of slender structure (21 meters high) a new facility named VESUBIE was designed and installed in an existing pit located at the Saclay nuclear research center. The objectives of the tests were the following: demonstrate insertability of control rod, demonstrate absence of seismic induced damage to the SCP, measure increase of scram time, measure seismic induced stresses, obtain data for code correlation. After completion of the tests, measurements have been correlated with results obtained from a non linear finite element model. Time history correlations were achieved for SCP 1. Afterwards a calculation was performed in hot condition to find if there was some effect of temperature on SCP seismic response. (orig./GL)

  12. 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.

  13. Variable reluctance electric motor for moving control rods in a nuclear reactor

    International Nuclear Information System (INIS)

    Under the invention, the motor includes a mobile part formed by a rod in a magnetic material fitted with ring projections regularly spaced along the rod; at least three multipole stators consecutively around the rod, each polar core having on the rod side annular hollows, so as to form sector shaped projections of the same width and spacing as those of the rod the spacing between stators being such that the rod projections and those of the polar cores face each other for one of the stators only. A non-magnetic leak-tight sleeve is placed at the intersection of the stator magnetic circuit and magnetic parts extend the polar cores inside the sleeve. Application to nuclear reactors

  14. Testing and qualification of Control and Safety Rod and its drive mechanism of Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Prototype Fast Breeder Reactor (PFBR) has two independent fast acting diverse shutdown systems. The absorber rod of the first system is called Control and Safety Rod (CSR). CSR and its Drive Mechanism (CSRDM) are used for reactor control and for safe shutdown of the reactor by scram action. In view of the safety role, the qualification of CSRDM is one of the important requirements. CSR and CSRDM were qualified in two stages by extensive testing. In the first stage, the critical subassemblies of the mechanism, such as scram release electromagnet, hydraulic dashpot and dynamic seals and CSR subassembly, were tested and qualified individually simulating the operating conditions of the reactor. Experiments were also carried out on sodium vapour deposition in the annular gaps between the stationary and mobile parts of the mechanism. In the second stage, full-scale CSRDM and CSR were subjected to all the integrated functional tests in air, hot argon and subsequently in sodium simulating the operating conditions of the reactor and finally subjected to endurance tests. Since the damage occurring in CSRDM and CSR is mainly due to fatigue cycles during scram actions, the number of test cycles was decided based on the guidelines given in ASME, Section III, Div. 1. The results show that the performance of CSRDM and CSR is satisfactory. Subsequent to the testing in sodium, the assemblies having contact with liquid sodium/sodium vapour were cleaned using CO2 process and the total cleaning process has been established, so that the mechanism can be reused in sodium. The various stages of qualification programmes have raised the confidence level on the performance of the system as a whole for the intended and reliable operation in the reactor.

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

    International Nuclear Information System (INIS)

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

  16. Neutron monitoring system and rod control system upgrades for plant life extension

    International Nuclear Information System (INIS)

    Most nuclear power plants in operation were built in the 1970's and 1980's. Plants of this vintage would require extensive upgrade for the instrumentation and control (I and C) to be able to support plant life extension. The most challenging aspect of I and C upgrades for these plants is the upgrade of the neutron monitoring system (NMS) and the Reactor Manual Control System (RMCS). These are specialized instrumentation and would require high quality design and engineering products to ensure safe and efficient plant operation. GEHitachi (GEH) Energy's nuclear business provides a Wide Range Neutron Monitoring System (WRNM) to replace the existing Source Range Monitors (SRM) and Intermediate Range Monitors (IRM), a Power Range Neutron Monitoring System (PRNM) to replace the Average Power Range Monitor (APRM), and a Rod Control Management System (RCMS) to replace the original Reactor Manual Control System (RMCS) in the GE designed Boiling Water Reactors (BWR). The WRNM, PRNM, and RCMS are based on the Nuclear Monitoring Analysis and Control (NUMAC) platform, which is a microprocessor based system that provides improved system performance with standard features such as improved HMI, selftest and automatic calibration. In addition to enhancing the system functions, these upgrades also help to support the challenges of plant life extension. This paper presents the designs and experience of these GE-Hitachi Nuclear Energy systems in support for nuclear plants life extension. (author)

  17. Controlled fabrication of individual silicon quantum rods yielding high intensity, polarized light emission

    International Nuclear Information System (INIS)

    Elongated silicon quantum dots (also referred to as rods) were fabricated using a lithographic process which reliably yields sufficient numbers of emitters. These quantum rods are perfectly aligned and the vast majority are spatially separated well enough to enable single-dot spectroscopy. Not only do they exhibit extraordinarily high linear polarization with respect to both absorption and emission, but the silicon rods also appear to luminesce much more brightly than their spherical counterparts. Significantly increased quantum efficiency and almost unity degree of linear polarization render these quantum rods perfect candidates for numerous applications.

  18. 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)

  19. Detection circuit of solenoid valve operation and control rod drive mechanism utilizing the circuit

    International Nuclear Information System (INIS)

    Object: To detect the operation of a plunger and detect opening and closing operations of a solenoid valve driving device due to change in impedance of a coil for driving the solenoid valve to judge normality and abnormality of the solenoid valve, thereby increasing reliance and safety of drive and control apparatus of control rods. Structure: An arrangement comprises a drive and operation detector section wherein the operation of a solenoid driving device for controlling power supply to a coil for driving the solenoid valve to control opening and closing of the solenoid valve, and a plunger operation detector section for detecting change in impedance of the drive coil to detect that the plunger of the solenoid valve is either in the opening direction or closing direction, whereby a predetermined low voltage such as not to activate the solenoid valve even when the solenoid valve is open or closed is applied to detect a current flowing into the coil at that time, thus detecting an operating state of the plunger. (Yoshino, Y.)

  20. 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.

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

    International Nuclear Information System (INIS)

    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

  2. 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.)

  3. 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)

  4. Growth control of ZnO nano-rod with various seeds and photovoltaic application

    International Nuclear Information System (INIS)

    ZnO has attracted much interesting as one of unique materials. Especially, it is suitable for the easy fabrication of nano-structures such as rod, wire and tube as well as particles. ZnO nano-rod is one of good sensitized electrodes because it has good electron transfer and is easily fabricated. In the chemical bath deposition process, seed layer plays an important role in the growth of nano-rod. This work investigated and analyzed the effect of seed layer on the growth of ZnO nano-rod. Fabricated nano-rods were applied to dye-sensitized solar cell. For better performance, ZnO was surface-modified by TiO2. Surface-modified ZnO had improved electron transfer and wider surface area. Consequently, the current and fill factor were much improved and overall performance was also enhanced with them.

  5. Development of a personal-computer-based system for control rod worth determination in a nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ansari, S.A.; Majid, B.; Hussain, A. (Nuclear Engineering Div., Pakistan Inst. of Nuclear Science and Technology, P.O. Nilore, Islamabad (PK))

    1990-12-01

    A new and compact system based on a personal computer has been developed for on-line determination of control rod worth in the upgraded 10-MW core of the Pakistan Research Reactor (PARR-1). The system utilizes a locally designed pulse acquisition card interfaced with the PC that features multiple inputs, high signal resolution, and fast data acquisition in a single slot card. The system acquires the pulse signals from the start-up channels and computes the system negative reactivity along with the associated errors using the inverse kinetic rod drop (IKRD) algorithm. The affect on reactivity measurements of relative detector position in the core is discussed. The results of rod worth measurements under different core conditions on the PARR-1 and PARR-2 cores are described. The application of the system for reactivity determination in the case of a highly subcritical core and large gamma background is discussed.

  6. Neutron monitoring system and rod control system upgrades for plant life extension

    International Nuclear Information System (INIS)

    Most nuclear power plants in operation were built in the 1970's and 1980's. The most challenging aspect of I and C upgrades for these plants is the upgrade of the neutron monitoring system (NMS) and the Reactor Manual Control System (RMCS). These are specialized instrumentation and would require high quality design and engineering products to ensure safe and efficient plant operation. Specifically, GE Energy's nuclear business provides a Wide Range Neutron Monitoring System (WRNM) to replace the existing Source Range Monitors (SRM) and Intermediate Range Monitors (IRM), a Power Range Neutron Monitoring System (PRNM) to replace the Average Range Power Monitor (APRM), and a Rod Control Management System (RCMS) to replace the original RCMS in the GE designed Boiling Water Reactors (BWR). The WRNM, PRNM, and RCMS are based on the Nuclear Monitoring Analysis and Control (NUMAC) platform, which is a microprocessor based system that provides improved system performance with standard features such as improved HMI, self-test and automatic calibration

  7. 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

  8. Thermo-Mechanical Analysis of Coated Particle Fuel Experiencing a Fast Control Rod Ejection Transient

    International Nuclear Information System (INIS)

    A rapid increase of the temperature and the mechanical stress is expected in TRISO coated particle fuel that experiences a fast Total Control Rod Ejection (CRE) transient event. During this event the reactor power in the pebble bed core increases significantly for a short time interval. The power is deposited instantly and locally in the fuel kernel. This could result in a rapid increase of the pressure in the buffer layer of the coated fuel particle and, consequently, in an increase of the coating stresses. These stresses determine the mechanical failure probability of the coatings, which serve as the containment of radioactive fission products in the Pebble Bed Reactor (PBR). A new calculation procedure has been implemented at the Idaho National Laboratory (INL), which analyzes the transient fuel performance behavior of TRISO fuel particles in PBRs. This early capability can easily be extended to prismatic designs, given the availability of neutronic and thermal-fluid solvers. The full-core coupled neutronic and thermal-fluid analysis has been modeled with CYNOD-THERMIX. The temperature fields for the fuel kernel and the particle coatings, as well as the gas pressures in the buffer layer, are calculated with the THETRIS module explicitly during the transient calculation. Results from this module are part of the feedback loop within the neutronic-thermal fluid iterations performed for each time step. The temperature and internal pressure values for each pebble type in each region of the core are then input to the PArticle STress Analysis (PASTA) code, which determines the particle coating stresses and the fraction of failed particles. This paper presents an investigation of a Total Control Rod Ejection (TCRE) incident in the 400 MWth Pebble Bed Modular reactor design using the above described calculation procedure. The transient corresponds to a reactivity insertion of $3 (∼2000 pcm) reaching 35 times the nominal power in 0.5 seconds. For each position in the

  9. 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

  10. 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.

  11. 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.

  12. State of the art of the conceptual designs for ASTRID control and shutdown rods

    International Nuclear Information System (INIS)

    A critical look at the conceptual designs of control and shutdown rods and absorber elements, along with the lessons learnt from the operation of French fast reactors (Phénix and Super-Phénix especially) and the associated irradiation tests, has yielded improved and even innovative absorber assembly design concepts which are presented in this paper. To comply with the GEN IV objectives set for the 600 MWe Advanced Sodium Technological Reactor for Industrial Demonstration (ASTRID), these design concepts have been researched with a view to improved economy/sustainability and enhanced safety. The two main measures undertaken to achieve economy, among many others, have been to reduce the absorber subassembly dimensions and boron carbide enrichment, as well as to extend the residence time. To achieve enhanced safety, measures could include improved components and/or structural materials and guidance surface coatings/hard-facings in active shutdown systems. As part of these measures, a new kind of absorber assembly has also been designed – called SEPIA – pertaining to safety devices for the passive insertion of negative reactivity in the core. Preliminary thermal-hydraulic and structural mechanical analyses have been carried out with the CADET and LICOS project codes to show their feasibility. Further detailed analyses need to be carried out to achieve optimum dimensions that comply with the RAMSES II design rules. The paper discusses the basis of the conceptual designs, giving due consideration to emerging design concepts, analysis backups and further R&D required for design qualification. (author)

  13. 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

  14. Simulation of Phenix control rod withdrawal experiments with SIMMER-IV

    International Nuclear Information System (INIS)

    The “end-of-life” tests performed in the PHENIX reactor in 2009, in particular the Control Rod (CR) withdrawal experiments provide an excellent opportunity for validation and verification of the reactor physics computer codes and modelling approaches. SIMMER-IV, a modern three-dimensional reactor safety code, has been recently employed at KIT for simulating these 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 at KIT. The reactor reactivity, power and neutron flux distributions calculated with SIMMER-IV are in good agreement with advanced neutronics codes, such as ERANOS, while the CR reactivity worth is overestimated due to neglecting heterogeneity effects. We show that SIMMER neutronics model can be improved 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 parameters, provided that a special treatment is employed for CR modelling. (author)

  15. Optimization and performance characteristics of servo-piston hydraulic control rod drive mechanism

    International Nuclear Information System (INIS)

    This paper introduces the structure and working principles of the servo-piston hydraulic control rod drive mechanism (SHCM), which can be moved continuously and has self-lock capacity. The steady state characteristics of SHCM are simulated using FLUENT codes. Based on comparison with the experimental results, the simulation is proven to be credible as a tool to describe the steady state characteristics. Finally, the influence of structural parameters is analyzed to obtain an optimal design. The experimental results indicate that the traction of the servo-tube is larger in the starting and braking stages. The resistance coefficient of SHCM increases gradually in the starting and lifting stage, and then tends to be stable. This coefficient has a maximum value while the inlet pressure is low. Performance norms of SHCM, such as the anti-disturbance ability and positioning accuracy, are tested, the anti-disturbance ability of the actuator is strong while the inlet pressure is fluctuating. The positioning accuracy is high regardless of the action process (lifting or not). (author)

  16. Assessment of the MDNBR enhancement methodologies for the SMART control rods banks withdrawal event

    International Nuclear Information System (INIS)

    For an electricity generation and seawater desalination, a 330 MW System-integrated Modular Advanced ReacTor (SMART) was developed by KAERI. The safety level of the SMART is enhanced when compared to that of the typical commercial reactors, with the aid of an elimination of a large break loss of coolant accident by placing the major components of the primary system in a reactor vessel and the adoption of a new technology and a passive design concept into the safety system. However, the events related to reactivity and power distribution anomalies have been evaluated as vulnerable points when compared to the other initiating events in the SMART, since the reactivity worth of the control rods (CR) banks is quite large due to the boron free core concept. Especially, safety margins, i.e., minimum departure from nucleate boiling ratio (MDNBR), are significantly threatened during the CR banks withdrawal event. Therefore, MDNBR enhancement methodology for the CR banks withdrawal event should be considered to further enhance the safety level of the SMART design. Two methodologies have been suggested to enhance the MDNBR during the CR banks withdrawal event: the application of a DNBR trip function into a core protection system and a turbine trip delay methodology. Sensitivity studies are performed to evaluate the two MDNBR enhancement methodologies and show that the suggested methodologies could enhance the MDNBR during the CR banks withdrawal event of the SMART

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

    CERN Document Server

    Lunnemann, Per; van Dijk-Moes, Relinde J A; Pietra, Francesca; Vanmaekelbergh, Daniël; Koenderink, A Femius

    2013-01-01

    We demonstrate that a simple silver coated ball lens can be used to accurately measure the entire distribution of radiative transition rates of quantum dot nanocrystals. This simple and cost-effective implementation of Drexhage's method that uses nanometer-controlled optical mode density variations near a mirror, not only allows to extract 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 ultra-stable 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 and we show that the rate distribution-width, that amounts to about 30% of the mean decay rate, is strongly dependent on the l...

  18. PWR Control Rod Ejection Analysis with the Method Of Characteristic Code DeCART

    Energy Technology Data Exchange (ETDEWEB)

    Hursin, Mathieu; Downar, Thomas J. [University of California at Berkeley, Berkeley (United States); Thomas, Justin [Argonne National Laboratory, Argonne (United States)

    2008-07-01

    During the past several years, a comprehensive high fidelity reactor core modeling capability has been developed called the Numerical Nuclear Reactor (NNR) (Weber,2003) for detailed analysis of Light Water Reactors. The NNR achieves high fidelity with a whole-core neutron transport solution and ultra-fine-mesh computational fluid dynamics/heat transfer solution. Previous applications of the NNR have been to the steady-state analysis of both pressurized and boiling water reactors. Recently there has been interest in taking advantage of the NNR to improve the fidelity for PWR transient analysis. The work described in this paper is a preliminary demonstration of the ability of the whole core neutron transport code, DeCART, to provide a detailed intra-pin-power distribution during a control rod ejection accident. The current state of the art in analysis of this event relies upon the assembly averaged power from a whole core nodal neutronics simulator and some type of pin power reconstruction within the fuel assembly. Both methodologies are briefly presented and applied to model a super-prompt reactivity insertion accident. The difference in the results of both approaches are discussed and the benefit of the DeCART methodology is described. (authors)

  19. PWR Control Rod Ejection Analysis with the Method Of Characteristic Code DeCART

    International Nuclear Information System (INIS)

    During the past several years, a comprehensive high fidelity reactor core modeling capability has been developed called the Numerical Nuclear Reactor (NNR) (Weber,2003) for detailed analysis of Light Water Reactors. The NNR achieves high fidelity with a whole-core neutron transport solution and ultra-fine-mesh computational fluid dynamics/heat transfer solution. Previous applications of the NNR have been to the steady-state analysis of both pressurized and boiling water reactors. Recently there has been interest in taking advantage of the NNR to improve the fidelity for PWR transient analysis. The work described in this paper is a preliminary demonstration of the ability of the whole core neutron transport code, DeCART, to provide a detailed intra-pin-power distribution during a control rod ejection accident. The current state of the art in analysis of this event relies upon the assembly averaged power from a whole core nodal neutronics simulator and some type of pin power reconstruction within the fuel assembly. Both methodologies are briefly presented and applied to model a super-prompt reactivity insertion accident. The difference in the results of both approaches are discussed and the benefit of the DeCART methodology is described. (authors)

  20. 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)

  1. 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)

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

    International Nuclear Information System (INIS)

    A rather critical problem to be faced in developing a safe strategy for the management of tritiated solid wastes is dealing with the outgassing property of tritium. Releases of tritium under elemental or oxide form may occur from waste items at different temperatures and rates depending upon the nature of tritium bonds into the waste matrix as well as on its 'contamination history'. Apart from the commercial value of tritium, its release from waste packages anyhow represents a risk of tritium exposure that cannot be accepted by skippers, by store and disposal site operators as well as by the general public. Consequently it is mandatory to carry out the detritiation of such wastes before their packaging and storage or disposal. In the boron carbide control rods from the Lingen BWR after about three years of operation, tritium generated by neutron reaction was essentially retained in the B4C matrix. The objectives of the study are to demonstrate the feasibility of two processes aimed at reducing to the maximum practicable extent the level of tritium contamination in such waste management are facilitated

  3. A Comparative analysis for control rod drop accident in RETRAN DNB and CETOP DNB Model

    International Nuclear Information System (INIS)

    In Korea, the nuclear industries such as fuel manufacturer, the architect engineer and the utility, have been using the methodologies and codes of vendors, such as Westinghouse(WH), Combustion Engineering, for the safety analyses of nuclear power plants. Consequently the industries have kept up the many organizations to operate the methodologies and to maintain the codes for each vendor. It may occur difficulty to improve the safety analyses efficiency and technology related. So, the necessity another of methodologies and code systems applicable to Non- LOCA, beyond design basis accident and performance analyses for all types of pressurized water reactor(PWR) has been raised. Due to the above reason, the Korea Electric Power Research Institute(KEPRI) had decided to develop the new safety analysis code system for Korea Standard Nuclear Power Plants in Korea. As the first requirement, the best-estimate codes were required for applicable wider application area and realistic behavior prediction of power plants with various and sophisticated functions. After the investigation for few candidates, RETRAN-3D has been chosen as a system analysis code. As a part of the feasibility estimation for the methodology and code system, CRD(Control Rod Drop) accident which an event of Non-LOCA accidents for Uljin units 3 and 4 and Yonggwang 1 and 2 was selected to verify the feasibility of the methodology using the RETRAN-3D. In this paper, RETRAN DNB Model and CETOP DNB Model were analyzed by using comparative method

  4. 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)

  5. Gas cooled fast reactor control rod drive mechanism deceleration unit. Test program

    Energy Technology Data Exchange (ETDEWEB)

    Wagner, T.H.

    1981-10-01

    This report presents the results of the airtesting portion of the proof-of-principle testing of a Control Rod Scram Deceleration Device developed for use in the Gas Cooled Fast Reactor (GCFR). The device utilizes a grooved flywheel to decelerate the translating assembly (T/A). Two cam followers on the translating assembly travel in the flywheel grooves and transfer the energy of the T/A to the flywheel. The grooves in the flywheel are straight for most of the flywheel length. Near the bottom of the T/A stroke the grooves are spiraled in a decreasing slope helix so that the cam followers accelerate the flywheel as they transfer the energy of the falling T/A. To expedite proof-of-principle testing, some of the materials used in the fabrication of certain test article components were not prototypic. With these exceptions the concept appears to be acceptable. The initial test of 300 scrams was completed with only one failure and the failure was that of a non-prototypic cam follower outer sleeve material.

  6. Gas cooled fast reactor control rod drive mechanism deceleration unit. Test program

    International Nuclear Information System (INIS)

    This report presents the results of the airtesting portion of the proof-of-principle testing of a Control Rod Scram Deceleration Device developed for use in the Gas Cooled Fast Reactor (GCFR). The device utilizes a grooved flywheel to decelerate the translating assembly (T/A). Two cam followers on the translating assembly travel in the flywheel grooves and transfer the energy of the T/A to the flywheel. The grooves in the flywheel are straight for most of the flywheel length. Near the bottom of the T/A stroke the grooves are spiraled in a decreasing slope helix so that the cam followers accelerate the flywheel as they transfer the energy of the falling T/A. To expedite proof-of-principle testing, some of the materials used in the fabrication of certain test article components were not prototypic. With these exceptions the concept appears to be acceptable. The initial test of 300 scrams was completed with only one failure and the failure was that of a non-prototypic cam follower outer sleeve material

  7. 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

  8. Annex II. Technique on predictive maintenance for magnetic Jack type control rod driving system

    International Nuclear Information System (INIS)

    Currently, maintaining a high level of availability and reliability in Nuclear Power Plant (NPP) operation is one of the most important issues. In this point of view, the nuclear power generation industry is experiencing an increased awareness and emphasis on the benefits of predictive maintenance because the condition based predictive maintenance can enhance the operability and reliability of the plant and reduce possibility of unwanted reactor trips. KOPEC has developed the online Coil Current Monitoring System (CCMS) for the effective and predictive maintenance of the magnetic jack type Control Rod Driving System (CRDS) of the Korean Standard Nuclear Power Plants (KSNPs). The CRDS of KSNP is composed of 73 Control Element Drive Mechanisms (CEDMs) and CEDM Control System(CEDMCS). The CCMS is capable of monitoring all coil current traces of up to 8 CEDMs at the same time. All coil currents of selected CEDMs are monitored and acquired automatically at high speed sampling rate depending on the CEDM movement. The acquired coil current data is automatically analyzed to check the timing sequence of CEDM motion, coil current amplitude and pattern mismatch of waveform. The various analytical results and diagnosis information of abnormal conditions are provided on MMI displays. For a more precise analysis, the CCMS provides various data manipulation tools of noise filtering, data comparison and database generation for automatic fault detection. The CCMS has been supplied to six KSNP units and mainly used for the predictive maintenance of the CEDM and CEDMCS during plant overhaul period. The CCMS has shown that it is very useful to diagnosis quickly and exactly the status of many control components of the CEDM and CEDMCS such as CEDM coils, hall effect current sensor, timing control cards, voltage adjustment cards and Silicon Controlled Rectifier (SCR). If enough coil current data are accumulated historically, the degradation trends of CEDM coil, hall effect current sensor

  9. Dynamic Rod Worth Measurement

    International Nuclear Information System (INIS)

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

  10. Development of carbon/carbon composite control rod for HTTR. 1. Preparation of elements and their fracture tests

    International Nuclear Information System (INIS)

    For the High Temperature Engineering Test Reactor(HTTR) the control rod sleeve is made of Alloy 800H for which a particular process is imposed when the reactor needs to be scrammed. The less restricted operation of the reactor would be attained if there would be the control rod more resistant to high temperature and neutron irradiation. This report summarizes the results which have been obtained as of March 1996 in the course of the development of the C/C composite control rod. Materials used were pitch- or PAN-based fiber-reinforced 2-dimensional carbon composites, from which preforms of the elements of a control rod were fabricated. The preforms were carbonized at 1000degC after being impregnated with pitch. Then they were graphitized at 3000degC, followed by a purification treatment with halogen. The elements included the pellet holder, lace truck and pin. The pin was fabricated by the fiber laminating technique. A control rod is to consist of pellet holders which are connected by the lace trucks with pins. Various strength tests were carried out on these elements. An irradiation of the elements made of PAN-based material was performed in JRR-3 at 900±50degC to a neutron fluence of 1x1025 n/m2 (E>29fJ). As for the strength tests on the elements, there were some differences between PAN- and pitch-based composites: In general, elements made of PAN-based composite showed the more plastic behavior before they fractured, whereas those of pitch-based material behaved in the more brittle manner. Fracture tests of the irradiated elements showed that fracture load and fracture displacement enough for assuring the integrity of the control rod structure were maintained even after the irradiation. It was also found that if the applied load was parallel to the fiber felt plane both fracture load and strain increased, whereas the load increase and strain decrease were observed for the applied load against the plane. (J.P.N.)

  11. 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 ...

  12. 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.

  13. Measuring Tools Design of Control Rods Drop Time at the RSG-GAS Based on Labview V8.5 and DAQ6009

    International Nuclear Information System (INIS)

    The RSG-GAS reactor has 8 control rods that serve to control the rate of fission. Control rods are the most important technical safety systems and the last protective equipment to shut down the reactor in the event of abnormal incident. Testing of the control rods drop time is one way to ensure that the control rods can function in accordance with the requirements reactor operations. Existing test tools have limitations that can only measure one control rod at each measurement. Another problem is the difficulty of getting a replacement device with the same functionality in the market to replace existing tools if damaged Therefore, then we do design of control rods drop time based on Labview v8.5 and DAQ6009. The design has resulted design, components specification and programming that are expected to be applied to the manufacture of new control rods drop time measuring devices that have the same functionality as the previous tool with better facilities. (author)

  14. Rod drop measurement analysis

    International Nuclear Information System (INIS)

    In some cases control rod worth efficiencies evaluated by inverse point kinetics from out-core detector currents remarkable differ from direct calculations. Explanation of this effects is given and is supported by the analysis of some WWER-440 rod drop experiments. (Authors)

  15. Modeling of continuous withdrawal and falling out of CPS control rods accident, using QUABOX/CUBBOX-HYCA code

    International Nuclear Information System (INIS)

    At present, at the Ignalina NPP the process of a wider use of the new uranium-erbium fuel of higher saturation and the manual control rods of new design is going on. These actions are directed to reducing the reactor control and protection system (CPS) cooling circuit voiding effect and to improving the technical and economical reactor operation parameters. Continuous withdrawal and falling out of CPS control rods lead to the reactivity and power changes in the reactor core. Therefore, important for safety is the evaluation of the CPS ability to compensate for the resulting excess reactivity in the reactor core, having the changed core loading conditions during such accidents. This article presents the calculation results of the continuous withdrawal and falling out of CPS control rods for the specific reactor core conditions of the Ignalina NPP Unit 2, i.e. during its operation on the maximum allowed power level of 4200 MW. The German code QUABOX/CUBBOX-HYCA with the improved CPS logic was used for the simulation of the above-mentioned transients. (author)

  16. Experimental investigation of power peak in vicinity of WWER-440 control rod at end of fuel cycle

    International Nuclear Information System (INIS)

    This paper presents some results of the axial power (fission density) distribution measurements carried out on the light-water, zero-power reactor LR-0 in a WWER-440 type core in vicinity of the WWER-440 control rod model at zero boron concentration in moderator, modelling the conditions at the end of the WWER-440 fuel cycle. Further information concerns the control rod model description, specification of the LR-0 core, fuel assemblies and measurement conditions. The aim of performed experiment is enlargement of the available power peaking database to enable the validation of the calculation codes by means of the measured data that correspond to the end of WWER-440 fuel cycle. (author)

  17. Correction method for critical extrapolation of control-rods-rising during physical start-up of reactor with spatial effect

    International Nuclear Information System (INIS)

    Reasons why the extrapolated critical curve obtained by lifting control rods is cambered during the physical start-up of a reactor are analyzed. Spatial flux deformation factor is introduced, and a new method, by which influences of spatial effect in the reactor are avoided additionally, is proposed based on what is achieved by removing source neutrons. The new method is employed to a real example. Comparing the new results with those of real physical start-up and achieved only by removing source neutrons, it is shown that the new method avoids cambering phenomenon of the extrapolated curve much better, and obtains more precise critical position of control rods, so the reactor will reach the criticality more safely. (authors)

  18. Step dynamic process of the hydraulically-driven control rod, (II). Theoretical model on step-down process

    International Nuclear Information System (INIS)

    The HCRDS (hydraulic control rod driving system) is a new type of control rod driving system, which is designed by INET (Institute of Nuclear Energy Technology) and has been put into use in 5 MW nuclear heating reactor in Tsinghua University. The purpose of this paper is to theoretically analyze the step-down process of this new technology and establish fundamental basis for further analysis and research. The experimental loop of the HCRDS and the working principle on the step-down process are introduced in this paper. The theoretical model is established on the basis of analysis, simplification and hypothesis. Also given is the accurate mathematical description of this model. The comparison between the results of this model and that of the experiment proves the rationality and feasibility of the model. The selection of the working point is also introduced. (author)

  19. 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.

  20. 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...

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

    International Nuclear Information System (INIS)

    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

  2. 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)

  3. Towards a reference numerical scheme using MCNPX for PWR control rod tip fluence estimations

    International Nuclear Information System (INIS)

    Recent occurrences of cracks and fissures on the cladding tubes of PWR control rod (CR) fingers employed in the Swiss reactors prompted the need to develop more reliable analytical methods for CR tip fluence estimations. To partly address this need, a deterministic methodology based on SIMULATE-3/CASMO-4 was in recent years developed at PSI. Although this methodology has already been applied for independent support to licensing issues related to CR lifetime, two main questions are currently being the center of attention for further enhancements. First, the methodology relies on several assumptions that have so far not been verified. Secondly, an assessment of the achieved accuracy has not been addressed. In an attempt to answer both these open questions, it was considered appropriate to develop an alternative computational scheme based on the stochastic MCNPX code with the objective to provide reference numerical solutions. This paper presents the first steps undertaken in that direction. To start, a methodology for a volumetric neutron source transfer to full core MCNPX models with detailed CR as well as axial reflector representations is established. On this basis, the assumptions of the deterministic methodology are studied for selected CR configurations for two Beginning-of-Life cores by comparing the spatial neutron flux distributions obtained with the two approaches for the entire spectrum. Finally, for the high-energy range (E> 1 MeV) and for a few CRs, the new MCNPX scheme is applied to estimate the accumulated fluence over one real operated cycle and the results are compared with the deterministic approach. (authors)

  4. 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)

  5. 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.

  6. A buoyantly-driven shutdown rod concept for passive reactivity control of a Fluoride salt-cooled High-temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Blandford, Edward D., E-mail: edb@unm.edu [Department of Chemical and Nuclear Engineering, University of New Mexico, Albuquerque, NM 87131-0001 (United States); Peterson, Per F. [Department of Nuclear Engineering, University of California, Berkeley, CA 94720-1730 (United States)

    2013-09-15

    Highlights: • We develop a novel buoyantly-driven shutdown rod concept for a FHR. • Shutdown rod system can be actively or passively activated during transients. • Response of the rod was computationally simulated and experimentally validated. • Initial results indicate rod could provide effective transient reactivity control. -- Abstract: This paper presents a novel buoyantly-driven shutdown rod concept for use in Fluoride salt-cooled High-temperature Reactors (FHRs). The baseline design considered here is a 900 MWth modular version of the FHR class called the Pebble Bed Advanced High-Temperature Reactor (PB-AHTR) that uses pebble fuel. Due to the high volumetric heat capacity of the primary coolant, the FHRs operate with a high power density core with a similar average coolant temperature as in modular helium reactors. The reactivity control system for the baseline PB-AHTR uses a novel buoyantly-driven shutdown rod system that can be actively or passively activated during reactor transients. In addition to a traditional active insertion mechanism, the new shutdown rod system is designed to also operate passively, fulfilling the role of a reserve shutdown system. The physical response of the shutdown rod was simulated both computationally and experimentally, using scaling arguments where applicable, with an emphasis on key phenomena identified by a preliminary Phenomena Identification and Ranking Table (PIRT) study. This paper presents results from both the pre-predicted simulation and experimental validation efforts.

  7. A buoyantly-driven shutdown rod concept for passive reactivity control of a Fluoride salt-cooled High-temperature Reactor

    International Nuclear Information System (INIS)

    Highlights: • We develop a novel buoyantly-driven shutdown rod concept for a FHR. • Shutdown rod system can be actively or passively activated during transients. • Response of the rod was computationally simulated and experimentally validated. • Initial results indicate rod could provide effective transient reactivity control. -- Abstract: This paper presents a novel buoyantly-driven shutdown rod concept for use in Fluoride salt-cooled High-temperature Reactors (FHRs). The baseline design considered here is a 900 MWth modular version of the FHR class called the Pebble Bed Advanced High-Temperature Reactor (PB-AHTR) that uses pebble fuel. Due to the high volumetric heat capacity of the primary coolant, the FHRs operate with a high power density core with a similar average coolant temperature as in modular helium reactors. The reactivity control system for the baseline PB-AHTR uses a novel buoyantly-driven shutdown rod system that can be actively or passively activated during reactor transients. In addition to a traditional active insertion mechanism, the new shutdown rod system is designed to also operate passively, fulfilling the role of a reserve shutdown system. The physical response of the shutdown rod was simulated both computationally and experimentally, using scaling arguments where applicable, with an emphasis on key phenomena identified by a preliminary Phenomena Identification and Ranking Table (PIRT) study. This paper presents results from both the pre-predicted simulation and experimental validation efforts

  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. 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

  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. Report of a consultants meeting on control rod insertion reliability for WWER-1000 nuclear power plants. Extrabudgetary programme on the safety of WWER and RBMK nuclear power plants

    International Nuclear Information System (INIS)

    Starting from 1992, an increased drop time of control rods exceeding the design limit of four seconds has been observed in most of the operating WWER-1000 reactors in Russia and in the Ukraine. In some cases a dropped control rod became stuck in an intermediate position near the bottom of the core. In October 1994, a similar control rod problem was also observed at Unit 6 of the Kozloduy NPP. The issue of control rod insertion reliability was considered at a consultants' meeting on ''Core Control and Protection Strategy of WWER-1000 Reactors'' in April 1994. A consultants' meeting specifically focused on ''Control Rod Insertion Reliability'' was convened in Vienna in February 1995 attended by 15 international experts. The objectives of this meeting were: The exchange of international experience on problems and solutions related to anomalous control rod insertion; judgement of the safety concern of this issue for WWER-1000 reactors based on safety analyses; consideration of regulatory requirements and interim measures to continue operation in short term including modifications implemented or planned; and, status of root cause analyses and pending problems. The technical discussions were held in plenary sessions and in three working groups devoted to specific aspects of the issue. Refs, figs, tabs

  12. Implementation of control rod movement and boron injection options by using control variables in RELAP5/PARCS v2.7 coupled code

    International Nuclear Information System (INIS)

    To efficiently characterize realistic transients, as the Reactivity Insertion Accidents (RIA), using coupled neutronic-thermal-hydraulic 3D best estimate system codes, like RELAP5/PARCS v2.7 coupled code, it is necessary to introduce some improvements in simulations by adding the capability of control rod movement and boron injection by means of RELAP5 control variables, with the aim of being able to analyze dynamically asymmetric transient accidents in a nuclear power reactor, like RIA, reproducing all control systems present in commercial reactors. In actual neutron kinetics codes, control rods banks do not have the possibility of dynamic movement during the simulation of a transient; besides it is necessary to send the boron concentration from the thermal-hydraulic code to the neutronic code to account for changes in cross-sections due to boron dilution. For instance, control rod movements are pre-programmed with simple instructions introduced before the beginning of the calculation. Hence, control rod positions are not related to the core characteristics and the control systems at any time of the simulation. This work presents the changes introduced in RELAP5/PARCS v2.7 codes to achieve that control rods and the boron injection become more dynamic and realistic components in such kind of simulators. With these modifications, control rods can be moved automatically, activated by the RELAP code control system, and also they can depend on signals related to the reactor activity, like pressure, fuel temperature or moderator temperature, etc., improving the realism of the calculation and widening the simulation possibilities. RELAP5 calculates the boron concentration in each node of the channels representing the reactor core, sending this information to the PARCS neutronic code. The environment of work chosen have been the graphical environment of programming Compaq Visual Fortran 6.6A (CVF 6.6A). The fundamental reasons have been, on the one hand the facility of

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

    International Nuclear Information System (INIS)

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

  14. Design report on the guide box-reactivity and safety control plates for MPR reactor under normal operation conditions

    International Nuclear Information System (INIS)

    The reactivity control system for the MPR reactor (Multi Purpose Reactor) is a critical component regarding safety, it must ensure a fast shut down, maintaining the reactor in subcritical condition under normal or accidental operation condition. For this purpose, this core component must be designed to maintain its operating capacity during all the residence time and under any foreseen operation condition. The mechanical design of control plates and guide boxes must comply with structural integrity, maintaining its geometric and dimensional stability within the pre-established limits to prevent interferences with other core components. For this, the heat generation effect, mechanical loads and environment and irradiation effects were evaluated during the mechanical design. The reactivity control system is composed of guide boxes, manufactured from Aluminium alloy, located between the fuel elements, and control absorber plates of Ag-In-Cd alloy hermetically enclosed by a cladding of stainless steel sliding inside de guide boxes. The upward-downward movement is transmitted by a rod from the motion device located at the reactor lower part. The design requirements, criteria and limits were established to fulfill with the normal and abnormal operation conditions. The design verifications were performed by analytical method, estimating the guide box and control plates residence time. The result of the analysis performed, shows that the design of the reactivity control system and the material selected, are appropriate to fulfill the functional requirements, with no failures attributed to the mechanical design. (author)

  15. Neutronic design and comparative characteristics of new pin type control rod for 14-MW TRIGA-SSR reactor

    Energy Technology Data Exchange (ETDEWEB)

    Iorgulis, C.; Truta, C. [Institute for Nuclear Research, 0300 Pitesti (Romania)

    2001-07-01

    This paper presents the structural changes of the control rods (CR-s) which will be done in order to manufacture new ones and therefore to replace the original rods and to increase the present safety features of TRIGA 14 MW reactor in Pitesti. Few years ago two CR-s became inoperable due to the combined effect of welding corrosion, water penetration into the absorbent section of the CR and finally the swelling caused from the high internal pressure of the radiolysis-generated gases (gases which accelerated welding corrosion, water penetration, etc). Although safe reactor operation is not yet affected a decision to redesign a new CR was taken, entirely suitable to the general reactor design. The new CR will use boron carbide (same absorbent as in the original one); this baron carbide will be packed in a different way. A set of 16 Incoloy pins filled with a column of boron carbide pellets, leak-tight welded, will be manufactured. These 16 pins will be mounted in a square array with 5 pins on each side of the square (corner pins being the same far two sides). Neutronic analysis on the preliminary design was done using specific computer codes (WIMS, DFA, and MCNP). Reactivity worth of this new rod would be slight less (8%) than that of the original one, but still with full compensation capacity. Helium release analysis showed that gas pressure in the Incoloy tube after 15 years of normal operation would not exceed 57 ATM. (author)

  16. 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

  17. 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...

  18. Determination of transient temperature and heat flux on the surface of a reactor control rod based on temperature measurements at the interior points

    International Nuclear Information System (INIS)

    The paper presents heat transfer calculation results concerning a control rod of nuclear power plant. Apart from numerical calculation results, experimental heat transfer measurements of the control rod model are also presented. The control rod that is the object of interest is surrounded by a mixing region of hot and cold streams and, as a consequence, is subjected to thermal fluctuations. The paper describes a method based on the solution of the inverse heat conduction problem (IHCP) for determining heat flux on the outer surface of the rod. Numerical tests were conducted to validate the method by comparison of the results with the time changes of surface temperature and heat flux which were obtained from the computational fluid dynamics (CFD) simulation of the mixing process. A measuring instrument was designed to measure the heat flux at the outer surface of the control rod model. In addition, the principle of operation and construction of heat flux meter is presented in detail. -- Highlights: • Temperature and heat flux estimation during cooling of control rod are presented. • The inverse technique is based on the space marching method. • The instrument for surface heat flux measurement was manufactured and tested. • CFD simulations were used to validate the developed inverse technique. • Actual data were used to demonstrate practical applicability of the method

  19. Investigation of the Coupled Reactivity Effects of the Movable Reflector and Safety Control Rods in the GFR

    International Nuclear Information System (INIS)

    Since the transient behaviour of the reactor core depends also on the fraction of neutrons that leak out of the core, the core control and reactivity management may benefit from a system of partially moveable reflector incorporated in the design. In fast reactors a larger migration area leading to a significant leak of neutrons can be observed because especially the transport cross-sections are in general smaller as compared to light water reactors. The utilization of a moveable reflector system in conjunction with dedicated safety control rods can increase the ability of accident managing due to enhanced escaping neutrons which otherwise would be reflected back into the fuel zone. The paper demonstrates the possibility of better controlling the transient reactor by additionally moving selected reflector subassemblies with higher neutronics importance. The main purpose of the analysis of the Gas-cooled Fast Reactor (GFR) presented in the full paper are investigations of the kinetic parameters and of the control and reflector rod worths, as well as optimization of the parts used for partial reflector withdrawal. The results found in this study may serve for future design improvements. (author)

  20. 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)

  1. Piecewise linear approximation: application to control rod step counting in a nuclear reactor core and image contours characterization

    International Nuclear Information System (INIS)

    After a survey of main algorithms for piecewise linear approximation, a new method is suggested. It consists of two stages: a sequential detection stage and an optimization stage, which derives from general dynamic clustering principle. It is applied to control rod step counting in a nuclear reactor core and images contours characterization. Another version of our method is presented. Its originality cames from the variability of the line segments number during iterations. A comparative study is made by comparing the results of the proposed method with of another methods already existing thereby it attests the efficiency and reliability of our method

  2. 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

  3. Design simplification of a small nuclear reactor for large-diameter neutron transmutation doping silicon using control rods

    International Nuclear Information System (INIS)

    A design concept for a small nuclear reactor for neutron transmutation doping silicon (NTD-Si) using a Pressurized Water Reactor (PWR) full-length fuel assembly was proposed in our previous work. The excess reactivity was suppressed by a combination of Gd2O3 and soluble boron, which results in a flatter flux profile over the core than with control rod insertion; however, the soluble boron system for reactivity control is quite complex and expensive. The removal of this system would make the design much simpler. In the current work, the removal of soluble boron is considered. Criticality, neutron transportation and core burn-up calculations were performed using the MVP/GMVP II code and MVP-BURN code. The calculation results show that the insertion of control rods in five of the nine assemblies is enough to suppress reactivity. The thermal hydraulic analysis showed that heat removal from the core was possible under 1 atm operating pressure. Silicon ingots up to 30 cm in diameter could be irradiated with sufficient uniformity in the irradiation channels. (author)

  4. 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)

  5. 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)

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

    International Nuclear Information System (INIS)

    This paper deals with the neutronic design of ELSY (the European Lead-cooled SYstem), a 600 MWe Fast Reactor developed within the 6th EURATOM Framework Programme. ELSY aims at being an 'adiabatic' system (as far as possible) in order to fulfill both the requirements of sustainability and proliferation resistance. It represents the European solution for the Lead Fast Reactor (LFR), one of the six candidate typologies proposed by the Generation-IV International Forum (GIF). The analysis of the ELSY reference configuration, with typical pure MOX loading, is here presented. An introductory investigation of the adiabatic and, possibly, the burner options viability is also achieved by providing a rough estimate of the Minor Actinides (MAs) equilibrium concentrations and time constants. One of the main challenge-points in the design of the core, made up of wrapper-less square Fuel Assemblies (FAs) according to the common scheme of PWRs, is the small delta-T between the coolant average outlet temperature (480degC) and the allowable cladding one (550degC): it requires a rather flat radial power distribution, obtained by segmenting the core in three zones with different enrichments. Three different control sets have been introduced in order to achieve the required reliability for reactor shutdown and safety systems: eight traditional concept Control Rod (CR) assemblies together with two independent systems of sparse control 'Finger Absorber' Rods (FARs), small B4C rods that can be inserted, in principle, in the center of each FA. One of the two finger absorber systems includes a subset of rods devoted to the regulation of the criticality swing during the cycle: their number can be limited indeed since the small reactivity swing (some hundreds pcm) due to the about unitary breeding ratio. Such an innovative solution can also be positioned in order to maintain an optimal power flattening during the fuel cycle. To verify the feasibility of this solution, a very detailed

  7. 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)

  8. Concept and neutronic design on the new pin-type control rods for TRIGA SSR 14 MW-INR Pitesti

    International Nuclear Information System (INIS)

    This paper presents the structural changes of the control rods (CR-s) intended to improve the safety features of TRIGA 14 MW reactor Pitesti. Few years ago two CR-s became inoperable due to the combined effect of welding corrosion, water penetration into the absorbent section of the CR and finally the swelling caused by the high internal pressure of the radiolysis-generated gases (gases which accelerated welding corrosion, water penetration, etc.). Although safe reactor operation is not yet affected, a decision to re-design a new CR was taken, entirely suitable to the general reactor design. The new CR will use boron carbide (same absorbent as in the original one), which will be packed in a different way. A set of 16 Incoloy pins filled with a column of boron carbide pellets, leak-tight welded will be manufactured. These 16 pins will be mounted in a square array with 5 pins on each side. Neutronic analysis on the preliminary design was done using specific computer codes (WIMS, DFA and and MCNP). Reactivity worth of this new rod would be slightly less (8%) than that of the original one, but still with full compensation capacity. Helium leak analysis showed that gas pressure in the Incoloy tube after 15 years of normal operation would not exceed 57 ATM. The new design maintains unchanged all the components of the CR but the shape and the enclosure of the boron carbide absorbent. In conclusion: - same boron carbide is the neutron absorbent as cylindrical pellets of 70% theoretical density; - absorbent pellets are definitely cladded in stainless steel tubes for all the estimated time of use; - the absorbent pins will be disposed in square array, 5 on each side of a standard rod location. So the concept of peripheral absorbent with a central trap is preserved; - all hydraulic and thermal conditions are preserved; - most important, the absorbent structure has to fulfil the major demand of safe operation - shutdown with the most reactive control rod fully withdrawn

  9. Monte Carlo MSM correction factors for control rod worth estimates in subcritical and near-critical fast neutron reactors

    Directory of Open Access Journals (Sweden)

    Lecouey Jean-Luc

    2015-01-01

    Full Text Available The GUINEVERE project was launched in 2006, within the 6th Euratom Framework Program IP-EUROTRANS, in order to study the feasibility of transmutation in Accelerator Driven subcritical Systems (ADS. This zero-power facility hosted at the SCK·CEN site in Mol (Belgium couples the fast subcritical lead reactor VENUS-F with an external neutron source provided by interaction of deuterons delivered by the GENEPI-3C accelerator and a tritiated target located at the reactor core center. In order to test on-line subcriticality monitoring techniques, the reactivity of all the VENUS-F configurations used must be known beforehand to serve as benchmark values. That is why the Modified Source Multiplication Method (MSM is under consideration to estimate the reactivity worth of the control rods when the reactor is largely subcritical as well as near-critical. The MSM method appears to be a technique well adapted to measure control rod worth over a large range of subcriticality levels. The MSM factors which are required to account for spatial effects in the reactor can be successfully calculated using a Monte Carlo neutron transport code.

  10. 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

  11. Control rod cluster drop time anomaly Guandong nuclear power station (Daya bay) and Electricite de France nuclear power stations (1450 MWe N4 Series)

    International Nuclear Information System (INIS)

    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)

  12. 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)

  13. Controlled synthesis of CuO nanostructures on Cu foil, rod and grid

    International Nuclear Information System (INIS)

    CuO nanowires are synthesized by heating Cu foil, rod and grid in ambient without employing a catalyst or gas flow at temperatures ranging from 400 to 800 deg. C for a duration of 1-12 h. Scanning electron microscopy (SEM) investigation reveals the formation of nanowires. The structure, morphology and phase of the as-synthesized nanowires are analyzed by various techniques such as X-ray diffraction (XRD), transmission electron microscopy (TEM), X-ray photoelectron spectroscopy (XPS), thermogravimetric analysis (TGA) and Fourier transform infrared spectroscopy (FTIR). It is found that these nanowires are composed of CuO phase and the underlying film is of Cu2O. A systematic study is carried out to find the possibilities for the transformation of one phase to another completely. A possible growth mechanism for the nanowires is also discussed.

  14. Experimental critical loadings and control rod worths in LWR-PROTEUS configurations compared with MCNPX results

    International Nuclear Information System (INIS)

    The PROTEUS research reactor at the Paul Scherrer Inst. (PSI) has been operating since the sixties and has already permitted, due to its high flexibility, investigation of a large range of very different nuclear systems. Currently, the ongoing experimental programme is called LWR-PROTEUS. This programme was started in 1997 and concerns large-scale investigations of advanced light water reactors (LWR) fuels. Until now, the different LWR-PROTEUS phases have permitted to study more than fifteen different configurations, each of them having to be demonstrated to be operationally safe, in particular, for the Swiss safety authorities. In this context, recent developments of the PSI computer capabilities have made possible the use of full-scale SD-heterogeneous MCNPX models to calculate accurately different safety related parameters (e.g. the critical driver loading and the shutdown rod worth). The current paper presents the MCNPX predictions of these operational characteristics for seven different LWR-PROTEUS configurations using a large number of nuclear data libraries. More specifically, this significant benchmarking exercise is based on the ENDF/B6v2, ENDF/B6v8, JEF2.2, JEFF3.0, JENDL3.2, and JENDL3.3 libraries. The results highlight certain library specific trends in the prediction of the multiplication factor keff (e.g. the systematically larger reactivity calculated with JEF2.2 and the smaller reactivity associated with JEFF3.0). They also confirm the satisfactory determination of reactivity variations by all calculational schemes, for instance, due to the introduction of a safety rod pair, these calculations having been compared with experiments. (authors)

  15. 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.

  16. 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

  17. 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

  18. An evaluation of the control rod modelling approach used in VSOP by comparing its results to the experiments performed in the ASTRA critical facility

    International Nuclear Information System (INIS)

    The modelling of strong absorber regions in diffusion theory is a well-known problem and many methods have been developed to accommodate the transport effects in diffusion theory. In this work the method of equivalent cross sections is evaluated for the ASTRA critical facility at the Russian Research Centre - Kurchatov Institute in Moscow. The measured reactivity worths of the control rods situated in the side reflector, are compared with the calculated values making use of equivalent diffusion parameters in VSOP. Favourable results were obtained for the control rods positioned within the first ring of reflector blocks with larger errors obtained for control rods positioned further from the core. Furthermore, the use of an equivalent boron concentration to represent the absorber regions was also investigated and shown to be useful if applied correctly and with care. However, the practical difficulties and restrictions imposed by the two approaches make the investigation of an alternative method, which should remove these shortcomings, attractive. (author)

  19. Measurements and analysis of neutron reaction rates and gamma-ray energy deposition in a critical assembly containing a central simulated boron control rod

    International Nuclear Information System (INIS)

    The main contributions to the power in a Boron control rod are provided by the energy deposition rates from alpha rays generated in the Boron and from gamma rays issued in the absorber and the surrounding fuel material. To check the validity of calculational methods and data for such a system, a simulated enriched Boron control rod has been built in the centre of the critical assembly BALZAC DE-2 in the MASURCA facility. Neutron capture rates in B10, fission rates in U-235, U-238 and Pu-239 have been measured through the core and the control rod. Gamma heat deposition rates and doses have been measured with ionization chamber and thermoluminescent dosemeters respectively, in a joint effort involving three European laboratories. This paper presents the experimental results and compares them with theoretical calculations. (author). 27 refs, 6 figs, 2 tabs

  20. 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)

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

    International Nuclear Information System (INIS)

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

  2. Parametric studies of the effect of MOx environment and control rods for PWR-UOx burnup credit implementation

    International Nuclear Information System (INIS)

    The increase of PWR-UOX fuel initial enrichment and the extensive needs for spent fuel storage or cask capacities reinforce the interest in taking burnup credit into account in criticality calculations. However, this utilization of credit for fuel burnup requires the definition of a methodology that ensures the conservatism of calculations. In order to guarantee the conservatism of the spent fuel inventory calculation, a depletion calculation scheme for burnup credit is under development. This paper presents the studies on the main parameters which have an effect on nuclides concentration: the presence of control rods during depletion and the fuel assembly environment, particularly the presence of MOx fuels around the UO2 assembly. Reactivity effects which are relevant to these parameters are then presented, and physics phenomena are identified. (author)

  3. 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)

  4. 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)

  5. Control of the neutronic and thermohydraulic conditions of power ramps in an irradiation loop for PWR fuel rod

    International Nuclear Information System (INIS)

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

  6. 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)

  7. 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

  8. 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)

  9. 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.

  10. 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

  11. 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

  12. 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)

  13. 2-D NON-CAUSAL SYSTEMS: CONTROL OF TEMERATURE VARIATION IN A ROD

    Czech Academy of Sciences Publication Activity Database

    Augusta, Petr; Frízl, R.; Hurák, Z.

    Bratislava: Slovak University of Technology in Bratislava, 2007, 022-1-022-7. ISBN 978-80-227-2677-1. [Process Control 2007. Štrbské Pleso (SK), 11.06.2007-14.06.2007] R&D Projects: GA MŠk(CZ) 1M0567 Institutional research plan: CEZ:AV0Z10750506 Keywords : Distributed-parameter systems * spatially distributed systems * multidimensional systems * multivariable polynomials * polynomial methods Subject RIV: BC - Control Systems Theory

  14. 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

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

    International Nuclear Information System (INIS)

    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

  16. 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

  17. 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)

  18. 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.

  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. 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.