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Sample records for control rod withdrawal

  1. Control rod withdrawal monitoring device

    International Nuclear Information System (INIS)

    Ebisuya, Mitsuo.

    1984-01-01

    Purpose: To prevent the power ramp even if a plurality of control rods are subjected to withdrawal operation at a time, by reducing the reactivity applied to the reactor. Constitution: The control rod withdrawal monitoring device is adapted to monitor and control the withdrawal of the control rods depending on the reactor power and the monitoring region thereof is divided into a control rod group monitoring region a transition region and a control group monitoring not interfere region. In a case if the distance between a plurality of control rods for which the withdrawal positions are selected is less than a limiting value, the coordinate for the control rods, distance between the control rods and that the control rod distance is shorter are displayed on a display panel, and the withdrawal for the control rods are blocked. Accordingly, even if a plurality of control rods are subjected successively to the withdrawal operation contrary to the control rod withdrawal sequence upon high power operation of the reactor, the power ramp can be prevented. (Kawakami, Y.)

  2. Control rod excess withdrawal prevention device

    International Nuclear Information System (INIS)

    Takayama, Yoshihito.

    1992-01-01

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

  3. Control device for the withdrawal of control rod

    International Nuclear Information System (INIS)

    Ando, Masaki.

    1985-01-01

    Purpose: To significantly suppress the maximum value of the control-rod worth upon control rod withdrawal. Constitution: At first, a signal for designating the first class is sent from a class-control section to the group-control section. In the group-control section, the peripheral group among the first class is designated by which the withdrawal of the control rods other than the peripheral group is inhibited and the control-rods in the peripheral group are withdrawn one by one. When all of them have been withdrawn, the group-control section designates the central group of the first class. All the control rods of the central group have been withdrawn, then the group-control section designates the peripheral group of the second class. Thereafter, the central group in the second class is designated. The control rods are thus withdrawn in the same manner hereinafter. The maximum value for the control-rod worth can be decreased by such a withdrawing sequence for the control rods. (Horiuchi, T.)

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

    Science.gov (United States)

    Smith, J.E.

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

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

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  6. Valuation of power oscillations in a BWR after control rod banks withdrawal events

    International Nuclear Information System (INIS)

    Costa, A. L.; Pereira, C.; Da Silva, C. A. M.; Veloso, M. A. F.

    2009-01-01

    The out-of-phase mode of oscillation is a very challenging type of instability occurring in BWR (Boiling Water Reactor) and its study is relevant because of the safety implications related to the capability to promptly detect any such inadvertent occurrence by in-core neutron detectors, thus triggering the necessary countermeasures in terms of selected rod insertion or even reactor shutdown. In this work, control rod banks (CRB) withdrawal transient was considered to study the power instability occurring in a BWR. To simulate this transient, the control rod banks were continuously removed from the BWR core in different cases. The simulation resulted in a very large increase of power. To perform the instability simulations, the RELAP5/MOD3.3 thermal hydraulic system code was coupled with the PARCS/2.4 3D neutron kinetic code. Data from a real BWR, the Peach Bottom, have been used as reference conditions and reactor parameters. The trend of the mass flow rate, pressure, coolant temperature and the void fraction to four thermal hydraulic channels symmetrically located in the core with respect to the core centre, were taken. It appears that the velocity of the rod bank withdrawal is a very important aspect for reactor stability. The slowest CRB withdrawal (180 s) did not cause power perturbation while the fast removal (20 s) triggered a slow power oscillation that little by little amplified to reach levels of more 100% of the initial power after about 210 s. The investigation of the related thermo hydraulic parameters showed that the mass flow rate, the void fraction and also the coolant temperature began to oscillate at approximately the same time interval

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

    International Nuclear Information System (INIS)

    Papin, Joelle; Lemoine, Francette; Sato, Ikken; Struwe, Dankward; Pfrang, Werner

    1994-01-01

    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)

  8. Control Rod Withdrawal Events Analyses for the Prototype Gen-IV SFR

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Chiwoong; Ha, Kwiseo; Jeong, Taekyeong; Jeong, Jaeho; Chang, Wonpyo; Lee, Seungwon; An, Sangjun; Lee, Kwilim [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    To confirm the limiting condition, based on the maximum allowable reactivity insertion of 0.3 $, three cases from the end of cycle (EOC) are selected. In addition, assuming the failure of CRSS by earthquake, additional cases is defined at beginning of cycle (BOC). When the CRW occurs, the reactor can be protected by plant protection system (PPS). In this study, PPS mechanism is sequentially studied for all initiating events. For design basis accidents (DBA), the reactor can be scrammed by reactor protection system (RPS). The first and seconds RPS signals are checked during transients. When RPS is failed, so called as anticipated transient without scram (ATWS), the reactor will be protected by diverse protection system (DPS). In this study, in order to analyze various initiating events related control rod withdrawal, four kinds of operating condition is defined. TOP events are analyzed using MARS-LMR. The influence of various plant protection system such as RPS and DPS are investigated.

  9. Control rod drives

    International Nuclear Information System (INIS)

    Nakamura, Akira.

    1984-01-01

    Purpose: To enable to monitor the coupling state between a control rod and a control rod drive. Constitution: After the completion of a control rod withdrawal, a coolant pressure is applied to a control rod drive being adjusted so as to raise only the control rod drive and, in a case where the coupling between the control rod drive and the control rod is detached, the former is elevated till it contacts the control rod and then stopped. The actual stopping position is detected by an actual position detection circuit and compared with a predetermined position stored in a predetermined position detection circuit. If both of the positions are not aligned with each other, it is judged by a judging circuit that the control rod and the control rod drives are not combined. (Sekiya, K.)

  10. WWER safety analysis in case simultaneous introduction of positive reactivity by withdrawal of control rods and injection of pure condensate

    International Nuclear Information System (INIS)

    Kuchin, A.V.; Ovdienko, Yu.N.; Khalimonchuk, V.A.

    2011-01-01

    The positive reactivity introduction rate is calculated for conditions of simultaneous effect on reactivity of two reactivity control systems incorporated in the WWER design: withdrawal of control rods and decrease in boric acid concentration. It is concluded that the actuation time for preliminary protection PZ-2 should be increased in the interval of pure condensate transport to completely exclude potential introduction of positive reactivity simultaneously by two different reactivity control systems

  11. Control rods

    International Nuclear Information System (INIS)

    Maruyama, Hiromi.

    1984-01-01

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

  12. Control rod

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  13. Control rod drives

    International Nuclear Information System (INIS)

    Shimano, Kunio; Nakamura, Akira; Mizuguchi, Koji; Sakai, Kazuhito; Mitsui, Hisayasu.

    1994-01-01

    The present invention concerns upper-built-in type control rod drives of a BWR type reactor. Namely, high temperature linear motor driving type control rod drives are disposed in an upper space of the reactor pressure vessel, which generates electromagnetic power. In usual driving of control rods, driving shafts connected with control rods by a high temperature linear motor driving system comprising a driving shaft having an iron core inserted therein and electromagnetic coils is vertically moved to insert/withdraw the control rods to and from the reactor core. Upon occurrence of reactor scram, electric power source is interrupted, and the control rods are rapidly inserted to the reactor core. According to the present invention, since the control rod drives are disposed in the space above the reactor pressure vessel, pipelines or equipments passing through the bottom of the reactor pressure vessel can be saved. As a result, operation for maintenance and inspection is facilitated. (I.S.)

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

    International Nuclear Information System (INIS)

    Bubelis, E.; Pabarcius, R.; Tonkunas, A.

    2003-01-01

    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)

  15. Control rod

    International Nuclear Information System (INIS)

    Takahashi, Akio.

    1982-01-01

    Purpose: To prevent distortion in control rod elements such as cladding tubes by decreasing the temperature difference between them. Constitution: In the case of housing a plurality of control rod elements in a protection pipe, flow rate control members are disposed in the protection pipe to equalize the flow resistance in each of coolant flow passages formed between the control rod elements and between the control rod elements and the inner surface of the protection pipe, to thereby unify the flow rate of the coolants flowing through these coolant flowing passages. Accordingly, each of the control rod elements can be cooled uniformly to thereby unify the temperature distribution and avoid the distortion in the cladding tubes, which may be resulted from bending due to the difference in thermal expansion and ununiform swelling due to the temperature difference. (Aizawa, K.)

  16. Control rod shutdown system

    International Nuclear Information System (INIS)

    Miyamoto, Yoshiyuki; Higashigawa, Yuichi.

    1996-01-01

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

  17. Analysis of NEA-NSC PWR Uncontrolled Control Rod Withdrawal at Zero Power Benchmark Cases with NODAL3 Code

    Directory of Open Access Journals (Sweden)

    Tagor Malem Sembiring

    2017-01-01

    Full Text Available The in-house coupled neutronic and thermal-hydraulic (N/T-H code of BATAN (National Nuclear Energy Agency of Indonesia, NODAL3, based on the few-group neutron diffusion equation in 3-dimensional geometry using the polynomial nodal method, has been verified with static and transient PWR benchmark cases. This paper reports the verification of NODAL3 code in the NEA-NSC PWR uncontrolled control rods withdrawal at zero power benchmark. The objective of this paper is to determine the accuracy of NODAL3 code in solving the continuously slow and fast reactivity insertions due to single and group of control rod bank withdrawn while the power and temperature increment are limited by the Doppler coefficient. The benchmark is chosen since many organizations participated using various methods and approximations, so the calculation results of NODAL3 can be compared to other codes’ results. The calculated parameters are performed for the steady-state, transient core averaged, and transient hot pellet results. The influence of radial and axial nodes number was investigated for all cases. The results of NODAL3 code are in very good agreement with the reference solutions if the radial and axial nodes number is 2 × 2 and 2 × 18 (total axial layers, respectively.

  18. Final PANTHER solution to the NEA-NSC3-DPWR core transient benchmark. Uncontrolled withdrawal of control rods at zero power

    International Nuclear Information System (INIS)

    Kuijper, J.C.

    1996-10-01

    This report contains the final results of PANTHER calculations for the 'NEA-NSC 3-D PWR Core Transient Benchmark: Uncontrolled Withdrawal of Control Rods at Zero Power'. PANTHER was able to model the benchmark problems without modifications to the code. All the calculations were performed in 3-D. (orig.)

  19. Control rod

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  20. Control rod assemblies

    International Nuclear Information System (INIS)

    Yamanaka, Toshikatsu.

    1986-01-01

    Purpose: To obtain simple and practical control rod assemblies by bringing the exit temperature of the guide tube of a control rod main body closer to that of an adjacent fuel assembly and thereby suppressing the wasteful flow of coolants. Constitution: A flow control member comprises an annular flow control plate disposed above the control rod main body and bellows having a plurality of small paertures capable of passing coolants therethrough formed at the circumferencial surface. The bellows are to cause the flow control plate to resiliently abut on the upper surface of the control rod main body. Coolants flowing from below to above in the guide tube remove heat from the neutron absorbers and are discharged externally at an elevated temperature, while coolants at a lower temperature are entered and mixed through the apertures formed in the bellows. By the way, upon insertion of the control rod main body, flow of the coolants to the inside of the bellows is substantially interrupted by the extension contraction of the bellows, by which the flow rate is adjusted depending on the withdrawing stroke to suppress the occurrence of thermal problems. (Kamimura, M.)

  1. Digital control rod blocking monitor

    International Nuclear Information System (INIS)

    Funayama, Yoshio.

    1996-01-01

    The present invention system is used for monitoring of a power region of a reactor, and used for monitoring of simultaneous withdrawal of a plurality of control rods without increasing the size or complicating the system. Namely, the system processes signals from a neutron flux detectors at the periphery of control rods controlled for withdrawal. As a result of the processing, the digital monitoring system generates an alarm when the reactor power at the periphery of the control rods fluctuates exceeding an allowable range. In the system, a control rod information forming means prepares frame data comprising front data, positions of the control rods to be withdrawn, frame numbers and completion data. A serial data transmitting means transmits the frame data successively as repeating frame data rows. A control rod information receiving means takes up the frame data of each of control rods to be withdrawn from the transmitted frame data rows. Since the system of the present invention can monitor the withdrawal of a plurality of control rods simultaneously without increasing the size or complicating the system, cost can be saved and the maintenance can be improved. (I.S.)

  2. Control rod drives

    International Nuclear Information System (INIS)

    Kimura, Koichi.

    1994-01-01

    In control rod drives, differential pressure sensors are disposed at the inlet and the exit of a driving water pressure control valve disposed in a driving water supply device and, when deviation of fluctuation of the differential pressure from a set value is detected, a pressure control valve for driving water is controlled so as to make the differential pressure constant. The differential pressure sensors detect the differential pressure between the pressure of the control rod drives at the inlet and the exit of the driving water pressure control valve and a pressure in a reactor dome. A judging circuit judges whether the differential pressure between both sides of the driving water pressure control valve is deviated from a set value or not and, if it deviates from the set value, outputs of judging signal to the control device. In the control device, the opening degree of the driving water pressure control valve is controlled, so that the differential pressure between both sides of the driving water pressure control value is constant and does not deviate from the set value. There are provided advantageous effects of preventing abnormal control rod withdrawing phenomenon to improve safety and reliability for the control of the reactor operation. (N.H.)

  3. Control rod displacement

    International Nuclear Information System (INIS)

    Nakazato, S.

    1987-01-01

    This patent describes a nuclear reactor including a core, cylindrical control rods, a single support means supporting the control rods from their upper ends in spaced apart positions and movable for displacing the control rods in their longitudinal direction between a first end position in which the control rods are fully inserted into the core and a second end position in which the control rods are retracted from the core, and guide means contacting discrete regions of the outer surface of each control rod at least when the control rods are in the vicinity of the second end position. The control rods are supported by the support means for longitudinal movement without rotation into and out of the core relative to the guide means to thereby cause the outer surface of the control rods to experience wear as a result of sliding contact with the guide means. The support means are so arranged with respect to the core and the guide means that it is incapable of rotation relative to the guide means. The improvement comprises displacement means being operatively coupled to a respective one of the control rods for periodically rotating the control rod in a single angular direction through an angle selected to change the locations on the outer surfaces of the control rods at which the control rods are contacted by the guide means during subsequent longitudinal movement of the control rods

  4. Control rod blocking monitor

    International Nuclear Information System (INIS)

    Suzuki, Shigeru.

    1993-01-01

    The number of times for setting up a control rod blocking monitor of a BWR type power plant is remarkably reduced to mitigate operator's burden. In the control rod blocking monitor, trip levels, as a judging standard upon outputting control rod blocking inhibition signals, are set up stepwise depending on the power level around control rods put to blocking control. The present invention comprises an allowance judging means capable of setting up trip levels for each of power levels corresponding to a plurality of control rods at once if the power levels are within the set up allowable range. With such a constitution, the set up allowable range is determined previously in the allowance judging means. Accordingly, when a gang blocking is conducted to control rods, if power levels around the control rods are increased at once into the set up allowable range, the trip levels for each of the control rods are set up at once. (I.S.)

  5. Control rod control device

    International Nuclear Information System (INIS)

    Seiji, Takehiko; Obara, Kohei; Yanagihashi, Kazumi

    1998-01-01

    The present invention provides a device suitable for switching of electric motors for driving each of control rods in a nuclear reactor. Namely, in a control rod controlling device, a plurality of previously allotted electric motors connected in parallel as groups, and electric motors of any selected group are driven. In this case, a voltage of not driving predetermined selected electric motors is at first applied. In this state an electric current supplied to the circuit of predetermined electric motors is detected. Whether integration or failure of a power source and the circuit of the predetermined electric motors are normal or not is judged by the detected electric current supplied. After they are judged normal, the electric motors are driven by a regular voltage. With such procedures, whether the selected circuit is normal or not can be accurately confirmed previously. Since the electric motors are not driven just at the selected time, the control rods are not operated erroneously. (I.S.)

  6. Control rod drives

    International Nuclear Information System (INIS)

    Hayakawa, Hiroyasu.

    1979-01-01

    Purpose: To enable rapid control in a simple circuit by providing a motor control device having an electric capacity capable of simultaneously driving all of the control rods rapidly only in the inserting direction as well as a motor controlling device capable of fine control for the insertion and extraction at usual operation. Constitution: The control rod drives comprise a first motor control device capable of finely controlling the control rods both in inserting and extracting directions, a second motor control device capable of rapidly driving the control rods only in the inserting direction, and a first motor switching circuit and a second motor switching circuit switched by switches. Upon issue of a rapid insertion instruction for the control rods, the second motor switching circuit is closed by the switch and the second motor control circuit and driving motors are connected. Thus, each of the control rod driving motors is driven at a high speed in the inserting direction to rapidly insert all of the control rods. (Yoshino, Y.)

  7. Control rod drives

    International Nuclear Information System (INIS)

    Oonuki, Koji.

    1981-01-01

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

  8. CONTROL ROD DRIVE

    Science.gov (United States)

    Chapellier, R.A.

    1960-05-24

    BS>A drive mechanism was invented for the control rod of a nuclear reactor. Power is provided by an electric motor and an outside source of fluid pressure is utilized in conjunction with the fluid pressure within the reactor to balance the loadings on the motor. The force exerted on the drive mechanism in the direction of scramming the rod is derived from the reactor fluid pressure so that failure of the outside pressure source will cause prompt scramming of the rod.

  9. Control rod testing apparatus

    Energy Technology Data Exchange (ETDEWEB)

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

    1987-06-02

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

  10. Control rod testing apparatus

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  11. Control rod velocity limiter

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  12. Control rod driveline and grapple

    International Nuclear Information System (INIS)

    Germer, J.H.

    1987-01-01

    A control rod driveline and grapple for engaging and releasing a control rod from a control rod drive is described comprising an enlarged control rod handle including an upwardly flaring frustum and a rod extending from the control rod handle; a relatively moving outer member; a tension rod connected to the relatively moving outer member at the upper end and provided with a lower annular flange at the lower end, the tension rod including a female cavity for receiving the upwardly extending rod from the enlarged control rod handle; a discrete and independent grapple segments for surrounding and grappling the control rod handle, each grapple segment including a first indentation for engaging and gripping the flange of the tension rod at an upper and interior annulus

  13. Control rod housing alignment

    International Nuclear Information System (INIS)

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

    1990-01-01

    This patent describes a process for measuring the vertical alignment between a hole in a core plate and the top of a corresponding control rod drive housing within a boiling water reactor. It comprises: providing an alignment apparatus. The alignment apparatus including a lower end for fitting to the top of the control rod drive housing; an upper end for fitting to the aperture in the core plate, and a leveling means attached to the alignment apparatus to read out the difference in angularity with respect to gravity, and alignment pin registering means for registering to the alignment pin on the core plate; lowering the alignment device on a depending support through a lattice position in the top guide through the hole in the core plate down into registered contact with the top of the control rod drive housing; registering the upper end to the sides of the hole in the core plate; registering the alignment pin registering means to an alignment pin on the core plate to impart to the alignment device the required angularity; and reading out the angle of the control rod drive housing with respect to the hole in the core plate through the leveling devices whereby the angularity of the top of the control rod drive housing with respect to the hole in the core plate can be determined

  14. Hydraulically centered control rod

    International Nuclear Information System (INIS)

    Horlacher, W.R.; Sampson, W.T.; Schukei, G.E.

    1981-01-01

    A control rod suspended to reciprocate in a guide tube of a nuclear fuel assembly has a hydraulic bearing formed at its lower tip. The bearing includes a plurality of discrete pockets on its outer surface into which a flow of liquid is continuously provided. In one embodiment the flow is induced by the pressure head in a downward facing chamber at the end of the bearing. In another embodiment the flow originates outside the guide tube. In both embodiments the flow into the pockets produces pressure differences across the bearing which counteract forces tending to drive the rod against the guide tube wall. Thus contact of the rod against the guide tube is avoided

  15. Control rod driving mechanisms

    International Nuclear Information System (INIS)

    Maejima, Yoshinori.

    1986-01-01

    Purpose: To conduct reactor scram by an external signal and, also by a signal for the abnormal temperature from a temperature detector in the nuclear reactor. Constitution: Control rod driving mechanisms magnetically coupling the extension pipe with the elevating mechanism above the reactor core and the holding magnet, and retains a control rod to the lower portion of the extension pipe by way of a latch mechanism. The temperature detector is immersed in reactor coolants. If the temperature of the coolants rises abnormally, bimetal contacts of the temperature detector are opened to interrupt the current supply to the holding electromagnet. Then, the extension pipe released from the magnetic coupling is lowered and the control rod free from latch is rapidly dropped and inserted into the reactor core. Since this procedure is carried out for all of the control rods, the reactor scram can be attained. The feature of this invention resides in that the reactor scram can be attained also by the signal of the reactor core itself even if the signal system for the external signals should be failed. (Horiuchi, T.)

  16. Device for coupling a control rod and control rod drive

    International Nuclear Information System (INIS)

    Nishioka, Kazuya.

    1975-01-01

    Object: To obtain simple and reliable coupling between a control rod and control rod drive by equipping the lower end of the control rod with an extension provided with lateral protuberances and forming the upper end of an index tube with a recess provided with lateral holes. Structure: The tapering central extension of the control rod is inserted into the recess by lowering the control rod, and then it is further inserted by causing frictional movement of the inclined surfaces of lateral protuberances in frictional contact with guide surfaces. When the lateral protuberances are brought into contact with a stepped portion, the control rod is rotated to fit the lateral protuberances into the lateral holes. In this way, the control rod is coupled to the index tube of the control rod drive. (Yoshino, Y.)

  17. Description and characterization of the ACRR's programmable transient rod withdrawal mode

    International Nuclear Information System (INIS)

    Boldt, K.R.; Sullivan, W.H.; Kefauver, H.L.

    1980-01-01

    To satisfy experiment needs for Sandia's Advanced Reactor Safety Program, a programmable Transient Rod Withdrawal (TRW) mode has been developed for the Annular Core Research Reactor (ACRR). The programmable mode is a modification of the existing continuous-withdrawal TRW mode and permits speed and direction changes during the pulse sequence. Basically, a TRW operation is similar to a routine pulse operation except that transient rods are mechanically withdrawn rather than pneumatically fired. Being a pulse-type operation, the TRW mode complies with pulse-mode safety system settings. Control system interlocks prevent the pneumatic firing of rods in the TRW mode. The hardware for the programmable TRW mode includes three ACRR transient rods, the ACRR timer, two rod programmers, a minicomputer and a summing circuit for position indication. Each ACRR transient rod is mechanically driven by a stepping motor (rated torque at 4.24 joules) and is capable of a maximum TRW speed of 26.7 centimeters/ second. The maximum reactivity insertion rate is $2.45/second with a transient rod bank worth of $3.00 and $3.47/second with a bank worth of $4.25, which is expected to be installed soon. The ACRR timer is a multifunctional timer used in all operating modes of the reactor. In the programmable TRW mode, the timer starts the rod programmers and drops regulating rods to terminate the operation. Programmed withdrawal capability is provided by one of two rod programmers (a hardwire-based unit and a microprocessor-based unit). The hardwire unit has eight intervals in which speed, direction and distance are selected by switches on the front panel. The microprocessor-based unit has the capability of 64 intervals in which speed, direction, and distance or time can be specified. Programming this unit is accomplished from the front panel or by inputting data from an HP-9845. minicomputer via a digital I/O interface. Self-test programs in the software provide a continual check of an operating

  18. REACTOR CONTROL ROD OPERATING SYSTEM

    Science.gov (United States)

    Miller, G.

    1961-12-12

    A nuclear reactor control rod mechanism is designed which mechanically moves the control rods into and out of the core under normal conditions but rapidly forces the control rods into the core by catapultic action in the event of an emergency. (AEC)

  19. Control rod drives

    International Nuclear Information System (INIS)

    Kono, Nobuaki.

    1987-01-01

    Purpose: To remove movable portion and improve the reliability by the direct control to coil. Constitution: Coils are disposed vertically at a predetermined interval to the outside of a control rod drive guide tube and each of the coils is adapted to be directly controlled. The coils are arranged at such an interval that a plunger laps over the vertically adjacent coils. In the case of moving the plunger upwardly, a coil just above the coil that attract the plunger is energized while the coil attracting the plunger so far is denergized. Then, the plunger is pulled up to an aimed position by repeating the procedures. In the case of moving the plunger downwardly, the procedures are conducted in the manner opposite to the above. (Kawakami, Y.)

  20. Integrated control rod monitoring device

    International Nuclear Information System (INIS)

    Saito, Katsuhiro

    1997-01-01

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

  1. Control rod drive hydraulic device

    International Nuclear Information System (INIS)

    Takekawa, Toru.

    1994-01-01

    The device of the present invention can reliably prevent a possible erroneous withdrawal of control rod driving mechanism when the pressure of a coolant line is increased by isolation operation of hydraulic control units upon periodical inspection for a BWR type reactor. That is, a coolant line is connected to the downstream of a hydraulic supply device. The coolant line is connected to a hydraulic control unit. A coolant hydraulic detection device and a pressure setting device are disposed to the coolant line. A closing signal line and a returning signal line are disposed, which connect the hydraulic supply device and a flow rate control valve for the hydraulic setting device. In the device of the present invention, even if pressure of supplied coolants is elevated due to isolation of hydraulic control units, the elevation of the hydraulic pressure can be prevented. Accordingly, reliability upon periodical reactor inspection can be improved. Further, the facility is simplified and the installation to an existent facility is easy. (I.S.)

  2. Nuclear reactor with control rods

    International Nuclear Information System (INIS)

    Obermeyer, F.D.; Berringer, R.T.

    1979-01-01

    A liquid-cooled nuclear reactor including fuel assemblies mounted within a reactor vessel having linearly movable control rods passing through control rod guide tubes into respective aligned fuel assemblies is described. Reactor coolant circulates through the assemblies. Guide tubes and other vessel internals structures located above the assemblies and is discharged through an outlet nozzle positioned above the elevation of primary flow openings in the guide tube walls. The guide tube includes internal horizontal supports and a length limited continuous control rod guide which, in conjunction with the flow openings, alleviate detrimental coolant cross flows and frictional restraints imposed upon the control rods

  3. Control rod experiments in Racine

    International Nuclear Information System (INIS)

    Stanculescu, A.; Humbert, G.

    1981-09-01

    A survey of the control-rod experiments planned within the joint CEA/CNEN-DeBeNe critical experiment RACINE is given. The applicability to both heterogeneous and homogeneous large power LMFBR-cores is discussed. Finally, the most significant results of the provisional design calculations performed on behalf of the RACINE control-rod programme are presented

  4. Control rod drives

    International Nuclear Information System (INIS)

    Sato, Takao; Arita, Setsuo; Mizuno, Katsuhiro.

    1986-01-01

    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.)

  5. Control rod drive

    International Nuclear Information System (INIS)

    Watando, Kosaku; Tanaka, Yuzo; Mizumura, Yasuhiro; Hosono, Kazuya.

    1975-01-01

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

  6. Nuclear reactor control rod

    International Nuclear Information System (INIS)

    Cearley, J.E.; Izzo, K.R.

    1987-01-01

    This patent describes a vertically oriented bottom entry control rod from a nuclear reactor: a frame including an elongated central spine of cruciform cross section connected between an upper support member and a lower support member both of cruciform shape having four laterally extending arms. The arms are in alignment with the arms of the lower support member and each aligned upper and lower support members has a sheath extending between; absorber plates of neutron absorber material, different from the material of the frame, one of the absorber plates is positioned within a sheath beneath each of the arms; attachment means suspends the absorber plates from the arms of the upper support member within a sheath; elongated absorber members positioned within a sheath between each of the suspended absorber plates and an arm of the lower support member; and joint means between the upper ends of the absorber members and the lower ends of the suspended absorber plates for minimizing gaps; the sheath means encloses the suspended absorber plates and the absorber members extending between aligned arms of the upper and lower support members and secured

  7. Control rod position detection device

    International Nuclear Information System (INIS)

    Akita, Haruo; Ogiwara, Sakae.

    1996-01-01

    The device of the present invention is used in a back-up shut down system of an LMFBR type reactor which is easy for maintenance, has high reliability and can recognize the position of control rods accurately. Namely, a permanent magnet is disposed to a control rod extension tube connected to the lower portion of the control rod. The detector guide tube is disposed in the vicinity of the control rod extension tube. A detector having a detection coil is inserted into a detector tube. With such constitution, the control rod can be detected at one position using the following method. (1) the movement of the magnetic field of the permanent magnet is detected by the detection coil. (2) a plurality of grooves are formed on the control rod extension tube, and the movement of the grooves is detected. In addition, the detection coil is inserted into the detector guide tube, and the signals from the detection coil are inputted to a signal processing circuit disposed at the outside of the reactor vessel using an MI cable to enable the maintenance of the detector. Further, if the detector comprises a detection coil and an excitation coil, the position of a dropped control rod can be recognized at a plurality of points. (I.S.)

  8. Process and apparatus for controlling control rods

    International Nuclear Information System (INIS)

    Gebelin, B.; Couture, R.

    1987-01-01

    This process and apparatus is characterized by 2 methods, for examination of cluster of nuclear control rods. Foucault current analyzer which examines fraction by fraction all the control rods. This examination is made by rotation of the cluster. Doubtful rods are then analysed by ultrasonic probe [fr

  9. Means for driving control rod

    International Nuclear Information System (INIS)

    Sato, Haruo; Sasaki, Masayoshi.

    1974-01-01

    Object: To enable wire rope to be readily removed from guide pulleys for the inspection or replacement of control rods. Structure: A pair of guide pulleys disposed to oppose each other are provided on their periphery with respective notches which are arranged in a staggered fashion. In this way, the rope is made to be removed from the notches for inspection of the control rod or for other purposes. (Kamimura, M.)

  10. Reactor control rod supporting structure

    International Nuclear Information System (INIS)

    Akimoto, Tokuzo; Miyata, Hiroshi.

    1984-01-01

    Purpose: To enable stable reactor core control even in extremely great vertical earthquakes, as well as under normal operation conditions in FBR type reactors. Constitution: Since a mechanism for converting the rotational movement of a control rod into vertical movement is placed at the upper portion of the reactor core at high temperature, the mechanism should cause fusion or like other danger after the elapse of a long period of time. In view of the above, the conversion mechanism is disposed to the lower portion of the reactor core at a lower temperature region. Further, the connection between the control rod and the control rod drive can be separated upon great vertical earthquakes. (Seki, T.)

  11. Advanced gray rod control assembly

    Energy Technology Data Exchange (ETDEWEB)

    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.

  12. Control rod housing alignment apparatus

    International Nuclear Information System (INIS)

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

    1991-01-01

    This paper discusses an alignment device for precisely locating the position of the top of a control rod drive housing from an overlying and corresponding hole and alignment pin in a core plate within a boiling water nuclear reactor. It includes a shaft, the shaft having a length sufficient to extend from the vicinity of the top of the control rod drive housing up to and through the hole in the core plate; means for registering the top of the shaft to the hole in the core plate, the registering means including means for registering with an alignment pin in the core plate adjacent the hole

  13. Control rod controlling device of nuclear reactor

    International Nuclear Information System (INIS)

    Arita, Setsuo; Okido, Fumiyasu.

    1997-01-01

    The present invention concerns a control rod drive mechanism for use in a BWR, which does not apply undesired effects on monitoring of neutron instrumentation systems. Control rods are operated using an induction electric motor equipped with an electromagnetic brake as a driving source. The induction electric motor and the electromagnetic brake are driven by ON/OFF control. Since a switching element for driving the induction electric motor and the electromagnetic brake can be kept ON or OFF during control rod operation, electromagnetic noises are not generated during the operation of the control rods. Accordingly, the neutron instrumentation systems do not undergoing effects of electromagnetic noises during operation of control rods, and the neutron instrumentation system can accurately be monitored. (N.H.)

  14. Control rod removal blocking device

    International Nuclear Information System (INIS)

    Yoshioka, Ritsuo.

    1981-01-01

    Purpose: To prevent excess power increase resulted from erroneous control rod removal during high power operation in BWR type reactor by decreasing the continuous removal enabling distance for the control rods along with increase in the reactor power where the reactor power is greater than a predetermined level. Constitution: When control rod selection signals are supplied from a control unit to a control rod removal blocking device, the blocking device judges whether the reactor core power is greater than a predetermined value A or not, using reactor core power signals outputted from an average power monitor. Where the reactor core power exceeds the predetermined value A and if the reactor core power is relatively low, a large continuous removal enabling distance N 1 is calculated in the blocking device to allow the continuous removal as far as the notch N 1 . The continuous removal enabling distance is shortened as the reactor core power increases and the removal is blocked, for example, at notch N 2 . While on the other hand, if the reactor core power is below the predetermined value A, both of the notchwise removal and the continuous removal are enabled. (Seki, T.)

  15. Control rod behaviour in earthquakes

    International Nuclear Information System (INIS)

    Kawakami, S.; Akiyama, H.; Shibata, H.; Watabe, M.; Ichikawa, T.; Fujita, K.

    1990-01-01

    For some years the Japanese have been working on a major research programme to determine the likely effects of an earthquake on nuclear plant internals. One aspect of this was a study of the behaviour of Pressurized Water Reactor control rods as they are being inserted in the core, which is reported here. (author)

  16. Anti-ejection device, which can be released, for control rods of nuclear reactor

    International Nuclear Information System (INIS)

    Belz, G.

    1983-01-01

    The present invention proposes an anti-ejection device which allows to withdraw the control rod out of a PWR reactor core if the locking systems of the rod translation are streck. This device prohibits the control rod ejection as long as an effort lower than a predetermined value is not applied on the control rod. This limit value is determined with regard of the efforts which may be applied on the control rod in case of an external accidental source. Nevertheless, if the anti-ejection mechanism remains stuck, it is however possible to withdraw the control rod out of the core applying on its control rod drives an effort higher than the limit value [fr

  17. Testing device for control rod drives

    International Nuclear Information System (INIS)

    Hayakawa, Toshifumi.

    1992-01-01

    A testing device for control rod drives comprises a logic measuring means for measuring an output signal from a control rod drive logic generation circuit, a control means for judging the operation state of a control rod and a man machine interface means for outputting the result of the judgement. A driving instruction outputted from the control rod operation device is always monitored by the control means, and if the operation instruction is stopped, a testing signal is outputted to the control rod control device to simulate a control rod operation. In this case, the output signal of the control rod drive logic generation circuit is held in a control rod drive memory means and intaken into a logic analysis means for measurement and an abnormality is judged by the control means. The stopping of the control rod drive instruction is monitored and the operation abnormality of the control rod is judged, to mitigate the burden of an operator. Further, the operation of the control rod drive logic generation circuit can be confirmed even during a nuclear plant operation by holding the control rod drive instruction thereby enabling to improve maintenance efficiency. (N.H.)

  18. Duke Power Company's control rod wear program

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  19. Control-rod scram device

    International Nuclear Information System (INIS)

    Matsui, Yoshiro; Saito, Koji.

    1986-01-01

    Purpose: To eliminate the requirement for the nitrogen gas system in a scram device and enable safety and reliable shutdown of a water-cooled reactor power plant. Constitution: A piston and a spring are contained within a hydraulic vessel, and the piston is driven by the energy stored in the spring so as to supply hydraulic water to control mechanisms. During usual reactor operation, a scram valve is closed and a high water pressure of about 130 kg/cm 2 is applied to the water filled in the vessel through a check valve. Upon occurrence of abnormal conditions and generation of scram signals, the scram valve is opened to supply the water filled in the vessel through the scram valve to the control rod drive mechanisms. When the water pressure in the vessel is decreased, since the piston is urged upwardly by the energy stored in the spring, the water filled in the vessel is intermitently supplied to the control rod drive mechanisms. Thus, control rods can be inserted into the nuclear reactor to shutdown the same. (Horiuchi, T.)

  20. Snubber assembly for a control rod drive

    International Nuclear Information System (INIS)

    Matthews, J.C.

    1978-01-01

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

  1. Key developments of a rod control system - 15101

    International Nuclear Information System (INIS)

    Pouillot, M.; Jegou, H.; Duthou, A.

    2015-01-01

    The aim of the Rod Control System is to carry out the insertion and withdrawal of control rod clusters to provide the power required by the grid (G-mode control), to control the temperature of the reactor, or to provide negative reactivity margin when the reactor is shut down. The rod control system is not classified important for safety, but its correct operation is essential for the availability of the reactor, as the spurious drop of a single cluster usually results in a reactor trip. Rolls-Royce has been designing, manufacturing and providing rod control systems since 1977, in France, China, Belgium, Korea, and South Africa, as an original manufacturer and for modernization projects. All the corresponding nuclear units share the following features, key points for the system design: -) The power source is a three-phased 260 Vac with neutral, provided by zigzag-coupled alternators; -) The Control Rod Drive Mechanisms (CRDM) are 'three-coil type': Stationary Gripper (SG), Movable Gripper (MG) and Lift Coil (LC); -) Rod clusters are arranged in banks and sub-banks, the bank being composed of one or two sub-banks and a sub-bank is a set of 4 clusters moved simultaneously, the central cluster being an exception; and -) Most of those reactors are operated in G-mode (load following). (authors)

  2. High temperature control rod assembly

    Science.gov (United States)

    Vollman, Russell E.

    1991-01-01

    A high temperature nuclear control rod assembly comprises a plurality of substantially cylindrical segments flexibly joined together in succession by ball joints. The segments are made of a high temperature graphite or carbon-carbon composite. The segment includes a hollow cylindrical sleeve which has an opening for receiving neutron-absorbing material in the form of pellets or compacted rings. The sleeve has a threaded sleeve bore and outer threaded surface. A cylindrical support post has a threaded shaft at one end which is threadably engaged with the sleeve bore to rigidly couple the support post to the sleeve. The other end of the post is formed with a ball portion. A hollow cylindrical collar has an inner threaded surface engageable with the outer threaded surface of the sleeve to rigidly couple the collar to the sleeve. the collar also has a socket portion which cooperates with the ball portion to flexibly connect segments together to form a ball and socket-type joint. In another embodiment, the segment comprises a support member which has a threaded shaft portion and a ball surface portion. The threaded shaft portion is engageable with an inner threaded surface of a ring for rigidly coupling the support member to the ring. The ring in turn has an outer surface at one end which is threadably engageably with a hollow cylindrical sleeve. The other end of the sleeve is formed with a socket portion for engagement with a ball portion of the support member. In yet another embodiment, a secondary rod is slidably inserted in a hollow channel through the center of the segment to provide additional strength. A method for controlling a nuclear reactor utilizing the control rod assembly is also included.

  3. Control rod for a reactor

    International Nuclear Information System (INIS)

    Natori, Hisahide.

    1975-01-01

    Object: To change arrangement and density of each layer of neutron absorber in the control rod and to render rotation by each layer possible, whereby the neutron absorber may be rotated to readily flatten power distribution. Structure: Neutron absorbers such as boron and carbide are filled into stainless steel pipes, which are peripherally arranged in a multi-layer fashion. Arrangement and density of the neutron absorber by each layer are changed and rotation by each layer is made possible, whereby surface area of the absorber or the like is changed to flatten power distribution. (Furukawa, Y.)

  4. Attracting electromagnet for control rod

    International Nuclear Information System (INIS)

    Kato, Kazuo; Sasaki, Kotaro.

    1989-01-01

    Non-magnetic material plates with inherent resistivity of greater than 20 μΩ-cm and thickness of less than 3 mm are used for the end plates of attracting electromagnets for closed type control rods. By using such control rod attracting electromagnets, the scram releasing time can be shortened than usual. Since the armature attracting side of the electromagnet has to be sealed by a non-magnetic plate, a bronze plate of about 5 mm thickness has been used so far. Accordingly, non-magnetic plate is inserted to the electromagnet attracting face to increase air source length for improving to shorten the scram releasing time. This method, however, worsens the attracting property on one hand to require a great magnetomotive force. For overcoming these drawbacks, in the present invention, the material for tightly closing end plates in an electromagnet is changed from bronze plate to non-magnetic stainless steel SUS 303 or non-magnetic Monel metal and, in addition, the plate thickness is reduced to less than 5 mm thereby maintaining the attracting property and shortening the scram releasing time. (K.M.)

  5. Control rod selecting and driving device

    International Nuclear Information System (INIS)

    Isobe, Hideo.

    1981-01-01

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

  6. Method of inspecting control rod drive mechanism

    International Nuclear Information System (INIS)

    Sato, Tomomi; Tatemichi, Shin-ichiro; Hasegawa, Hidenobu.

    1988-01-01

    Purpose: To conduct inspection for control rod drives and fuel handling operations in parallel without taking out the entire fuel, while maintaining the reactor in a subcritical state. Method: Control rod drives are inspected through the release of connection between control rods and control rod drives, detachment and dismantling of control rod drives, etc. In this case, structural materials having neutron absorbing power equal to or greater than the control rods are inserted into the gap after taking out fuels. Since the structural materials have neutron absorbing portion, subcriticality is maintained by the neutron absorbing effect. Accordingly, there is no requirement for taking out all of the fuels, thereby enabling to check the control rod drives and conduct handling for the fuels in parallel. As a result, the number of days required for the inspection can be shortened and it is possible to improve the working efficiency for the decomposition, inspection, etc. of the control rod drives and, thus, improve the operation efficiency of the nuclear power plant thereby attaining the predetermined purpose. (Kawakami, Y.)

  7. Control rod drive of nuclear reactor

    International Nuclear Information System (INIS)

    Zhuchkov, I.I.; Gorjunov, V.S.; Zaitsev, B.I.

    1980-01-01

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

  8. The analysis of RWAP(Rod Withdrawal at Power) using the KEPRI methodology

    International Nuclear Information System (INIS)

    Yang, C. K.; Kim, Y. H.

    2001-01-01

    KEPRI developed new methodology which was based on RASP(Reactor Analysis Support Package). In this paper, The analysis of RWAP(Rod Withdrawal at Power) accident which can result in reactivity and power distribution anomaly was performed using the KEPRI methodology. The calculation describes RWAP transient and documents the analysis, including the computer code modeling assumptions and input parameters used in the analysis. To validity for the new methodology, the result of calculation was compared with FSAR. As compared with FSAR, result of the calculation using the KEPRI Methodology is similar to FSAR's. And result of the sensitivity of postulated parameters were similar to the existing methodology

  9. 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.

  10. Control rod housing alignment and repair apparatus

    International Nuclear Information System (INIS)

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

    1991-01-01

    This patent describes a welding a repair device for precisely locating and welding the position of the top of a control rod drive housing attached from a stub tube from a corresponding aperture and alignment pin in a core plate within a boiling water nuclear reactor, the welding and repair device. It comprises: a shaft, the shaft extending from the vicinity of the top of the control rod drive housing up to and through the aperture in the core plate; means for registering to the aperture and the alignment pin on the core plate; a fixture attached to the bottom end of the shaft for mating to the top of the control rod drive housing in precise mating relationship; the fixture attached to the bottom end of the shaft whereby the fixture, when mated to the control rod drove housing and the registering means when registered to the alignment pin and aperture on the core plate imparts to the shaft, and angularity between the top of the control rod drive housing and the hole in the core plate; a hollow cylinder, the cylinder mounted for depending and sealed support with respect to the shaft above, about and below the control rod drive housing top; the cylinder depending down below the control rod drive housing to an elevation below the top of the sub tube; a rotating welding apparatus with a welding head for dispensing weldment mounted for rotation with respect to the shaft; the welding head disposed at the juncture between the side of the control rod drive housing and the stub tube; and means for flooding the cylinder with gas whereby the cylinder may be lowered. flooded in a gas environment and effect a weld between the top of the stub tube and the control rod drive housing

  11. Apparatus for handling control rod drives

    International Nuclear Information System (INIS)

    Akimoto, A.; Watanabe, M.; Yoshida, T.; Sugaya, Z.; Saito, T.; Ishii, Y.

    1979-01-01

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

  12. Control rod for a nuclear reactor

    International Nuclear Information System (INIS)

    Roman, W.G.; Sutton, H.G. Jr.

    1979-01-01

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

  13. Control rod for a nuclear reactor

    International Nuclear Information System (INIS)

    Roman, W.G.; Sutton, H.G. Jr.

    1976-01-01

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

  14. Estimation of irradiated control rod worth

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  15. Control rod driving hydraulic pressure device

    International Nuclear Information System (INIS)

    Ishida, Kazuo.

    1990-01-01

    Discharged water after actuating control rod drives in a BWR type reactor is once discharged to a discharging header, then returned to a master control unit and, subsequently, discharged to a reactor by way of a cooling water header. The radioactive level in the discharging header and the master control unit is increased by the reactor water to increase the operator's exposure. In view of the above, a riser is disposed for connecting a hydraulic pressure control unit incorporating a directional control valve and the cooling water head. When a certain control rod is inserted, the pressurized driving water is supplied through a hydraulic pressure control unit to the control rod drives. The discharged water from the control rod drives is entered by way of the hydraulic pressure control unit into the cooling water header and then returned to the reactor by way of other hydraulic pressure control unit and the control rod drives. Thus, the reactor water is no more recycled to the master control unit to reduce the radioactive exposure. (N.H.)

  16. Control rod for FBR type reactor

    International Nuclear Information System (INIS)

    Nakai, Koichi.

    1993-01-01

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

  17. Magnetic switch for reactor control rod. [LMFBR

    Science.gov (United States)

    Germer, J.H.

    1982-09-30

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

  18. Performance analysis of LMFBR control rods

    International Nuclear Information System (INIS)

    Pitner, A.L.; Birney, K.R.

    1975-01-01

    Control rods in the FFTF and LMFBR's will consist of pin bundles of stainless steel-clad boron carbide pellets. In the FFTF reference design, sixty-one pins of 0.474-inch diameter each containing a 36-inch stack of 0.362-inch diameter boron carbide pellets comprise a control rod. Reactivity control is provided by the 10 B (n,α) 7 Li reaction in the boron carbide. This reaction is accompanied by an energy release of 2.8 MeV, and heating from this reaction typically approaches 100 watts/cm 3 for natural boron carbide pellets in an LMFBR flux. Performance analysis of LMFBR control rods must include an assessment of the thermal performance of control pins. In addition, irradiation performance with regard to helium release, pellet swelling, and reactivity worth depletion as a function of service time must be evaluated

  19. Nuclear reactor remote disconnect control rod coupling indicator

    International Nuclear Information System (INIS)

    Vuckovich, M.

    1977-01-01

    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

  20. Method of diagnosing the control rod operation

    International Nuclear Information System (INIS)

    Nakaniwa, Tomoko; Kudo, Mitsuru.

    1983-01-01

    Purpose: To confirm the soundness of a control rod operation at high accuracy. Method: By utilizing the fact that a control rod position indicating system outputs digitalized decimal signals, position signals are outputted as below: a stop position signal when a magnet is opposed to an odd number limit switch; a drift position signal when it is opposed to an even number limit switch; and a blank position signal when it is not opposed to a limit switch. Since the standard signal pattern is present for the order of signal change due to the displacement of the magnet i.e., stop position→blank position→drift position, absolute position change and the time for passing the limit switch, the standard pattern is stored in an operation diagnosis unit and compared with the data upon actual driving to diagnose the control rod operation collectively. (Sekiya, K.)

  1. Storing device for control rod drive

    International Nuclear Information System (INIS)

    Tomatsu, Tsutomu; Miura, Teruo.

    1989-01-01

    A water supply device for supplying clean water, a recycling pump and a filter disposed between the water supply device and a water vessel by way of recycling pipelines are disposed to a water vessel containing storing water for immerging and storing control rod drives for BWR type reactors upon periodical inspection, etc. Clean water is supplied from the water supply device into the control rod drives immerged in the storing water to remove radioactive cruds, etc. deposited at the surface thereof and water is supplied through the recycling pipelines to the filter to remove solid impurities contained therein and the clean water is returned to the water supply device. Since the clean water is always recycled to the inside of the control rod drives, chemical corrosion and electrical corrosion of nitride parts are prevented and radioactive cruds are processed in separated waste processing systems, the atmospheric radiation doses in the operation chamber is reduced. (S.K.)

  2. HIGH STRENGTH CONTROL RODS FOR NEUTRONIC REACTORS

    Science.gov (United States)

    Lustman, B.; Losco, E.F.; Cohen, I.

    1961-07-11

    Nuclear reactor control rods comprised of highly compressed and sintered finely divided metal alloy panticles and fine metal oxide panticles substantially uniformly distributed theretbrough are described. The metal alloy consists essentially of silver, indium, cadmium, tin, and aluminum, the amount of each being present in centain percentages by weight. The oxide particles are metal oxides of the metal alloy composition, the amount of oxygen being present in certain percentages by weight and all the oxygen present being substantially in the form of metal oxide. This control rod is characterized by its high strength and resistance to creep at elevated temperatures.

  3. Automatic operation device for control rods

    International Nuclear Information System (INIS)

    Sekimizu, Koichi

    1984-01-01

    Purpose: To enable automatic operation of control rods based on the reactor operation planning, and particularly, to decrease the operator's load upon start up and shutdown of the reactor. Constitution: Operation plannings, demand for the automatic operation, break point setting value, power and reactor core flow rate change, demand for operation interrupt, demand for restart, demand for forecasting and the like are inputted to an input device, and an overall judging device performs a long-term forecast as far as the break point by a long-term forecasting device based on the operation plannings. The automatic reactor operation or the like is carried out based on the long-term forecasting and the short time forecasting is performed by the change in the reactor core status due to the control rod operation sequence based on the control rod pattern and the operation planning. Then, it is judged if the operation for the intended control rod is possible or not based on the result of the short time forecasting. (Aizawa, K.)

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

    International Nuclear Information System (INIS)

    Baron, Clemens

    2008-01-01

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

  5. Control rod for nuclear reactor

    International Nuclear Information System (INIS)

    Tada, Kaoru; Kawano, Shohei

    1998-01-01

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

  6. Analysis of buffering process of control rod hydraulic absorber

    International Nuclear Information System (INIS)

    Bao Jishi; Qin Benke; Bo Hanliang

    2011-01-01

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

  7. Removable control rod drive shaft guide

    International Nuclear Information System (INIS)

    Ales, M.W.; Brown, S.K.; Dixon, L.D.

    1988-01-01

    A removable control rod drive shaft guide is described for a control rod ''guide'' structure card, comprising: a. a substantially annular shaped main body portion having a central axial bore for receiving a control rod drive shaft and an upper exterior groove for receiving removal tooling; b. the main body portion having a reduced outer diameter at its lower section; c. a shoulder portion integral with the main body portion for supporting the main body portion on the guide structure card; d. the shoulder portion having a substantially radial bore and the reduced outer diameter lower section having a slot in alignment with the radial bore; e. a locking arm ''pivotaly'' mounted in the radial bore which protrudes into the slot and is movable between a first normal locking position for engaging the guide structure card and a second release position; f. a spring received within a second axial bore in the main body portion and biased against the locking arm for urging and locking arm into the first normal locking position; and g. a release tab at one end of the locking arm for moving the locking arm into the second release position

  8. Modeling and Experimental Tests on the Hydraulically Driven Control Rod option for IRIS Reactor

    International Nuclear Information System (INIS)

    Cammi, Antonio; Ricotti, Marco E.; Vitulo, Alessia

    2004-01-01

    The adoption of Internal Control Rod Drive Mechanisms (ICRDMs) represents a valuable alternative to classical, external CRDMs based on electro-magnetic devices, as adopted in current PWRs. The advantages on the safety features of the reactor are apparent: inherent elimination of the Rod Ejection accidents and of possible concerns about the vessel head penetrations. A further positive feedback on the design is the reduction of the primary system overall dimensions. Within the frame of the ICRDM concepts, the Hydraulically Driven Control Rod solution is investigated as a possible option for the IRIS integral reactor. After a brief comparison of the solutions currently proposed for integral reactors, the configuration of the Hydraulic Control Rod device for IRIS, made up by an external movable piston and an internal fixed cylinder, is described. A description of the whole control system is reported as well. Particular attention is devoted to the Control Rod profile characterization, performed by means of a Computational Fluid Dynamics (CFD) analysis. The investigation of the system behavior has been carried out, including the dynamic equilibrium and its stability properties, the withdrawal and insertion step movement and the sensitivity study on command time periods. A suitable dynamic model has been set up for the mentioned purposes: the models corresponding to the various Control Rod system devices have been written in an Object-Oriented language (Modelica), thus allowing an easy implementation of such a system into the simulator for the whole reactor. Finally, a preliminary low pressure, low temperature, reduced length experimental facility has been built. Tests on HDCR stability and operational transients have been performed. The results are compared with the dynamic system model and CFD simulation model, showing good agreement between simulations and experimental data. During these preliminary tests, the control system performed correctly, allowing stable dynamic

  9. Rolls-Royce digital Rod Control System

    International Nuclear Information System (INIS)

    Pouillot, M.

    2010-01-01

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

  10. Electronic circuit for control rod attracting electromagnet

    International Nuclear Information System (INIS)

    Ito, Koji.

    1991-01-01

    The present invention provides a discharging circuit for control rod attracting electromagnet used for a reactor which is highly reliable and has high performance. The resistor of the circuit comprises a non-linear resistor element and a blocking rectification element connected in series. The discharging circuit can be prevented from short-circuit by selecting a resistor having a resistance value about ten times as great as the coil resistance, even in a case where the blocking rectification element and the non-linear resistor element are failed. Accordingly, reduction of attracting force and the increase of scream releasing time can be minimized. (I.S.)

  11. Minimization of PWR reactor control rods wear

    International Nuclear Information System (INIS)

    Ponzoni Filho, Pedro; Moura Angelkorte, Gunther de

    1995-01-01

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

  12. Maintenance of BWR control rod drive mechanisms

    International Nuclear Information System (INIS)

    Greene, R.H.

    1991-01-01

    Control rod drive mechanism (CRDM) replacement and rebuilding is one of the highest dose, most physically demanding, and complicated maintenance activities routinely accomplished by BWR utilities. A recent industry workshop sponsored by the Oak Ridge National Laboratory, which dealt with the effects of CRDM aging, revealed enhancements in maintenance techniques and tooling which have reduced ALARA, improved worker comfort and productivity, and have provided revised guidelines for CRDM changeout selection. Highlights of this workshop and ongoing research on CRDM aging are presented in this paper

  13. Analysis of control rod drop accidents postulated in the Laguna Verde power plant

    International Nuclear Information System (INIS)

    Cuevas V, G.F.; Gonzalez C, J.

    2001-01-01

    In this paper, the results of the simulation of the Control Rod Drop Accident (CRDA) using the Studsvik- Scandpower Fuel Management System (FMS) are presented. The simulation conditions were selected to represent three operational conditions of the reactor core at Laguna Verde Nuclear Power Station (CLV). To avoid unnecessary symmetry considerations the full reactor core is modeled with a fuel load formed with a blend of two assembly types namely 10 x 10 and 8 x 8 arrays. Control Rod Worth was calculated for each of the inserted control rods using a representation of a realistic BWR core configuration, 444 fuel assemblies, l 09 control rods. Individual control rod withdrawals were simulated in order to identify the largest control rod worth. As part of the description of the core configuration, nuclear data banks for cold/hot conditions were generated. Likewise, the burnup and relative power distributions were calculated to resemble the operational conditions of the fifth fuel cycle of CLV unit 2. The main analysis code for this simulation, FMS RAMONA-3, provides dynamic behavior of the neutronic thermal hydraulic BWR feedback adequate for the characteristics of this accident. Furthermore, using a parameterization of core conditions the response of hot-spot fuel-clad enthalpies and temperatures were calculated. For the most-severe core condition, from the neutronics standpoint, potential fuel damage is assessed either for a hypothetical 293 isothermal core condition (fuel and moderator) or for a more realistic operational startup condition. (Author)

  14. Shielding device for control rod in nuclear reactor

    International Nuclear Information System (INIS)

    Sakamaki, Kazuo; Tomatsu, Tsutomu.

    1995-01-01

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

  15. Rebirth of a control rod at the Phenix power plant

    International Nuclear Information System (INIS)

    De Carvalho, Corinne; Vignau, Bernard; Masson, Marc

    2007-01-01

    This paper outlines the operations involved in cleaning the control rod for the complementary shutdown system in the Phenix Power Plant, the French sodium-cooled fast reactor. The Phenix reactor is controlled by six control rods and a complementary shutdown system. The latter comprises a control rod and a mechanism maintaining the rod in position by means of an electromagnet. The electromagnet is continuously supplied with power and holds the rod control assembly in position by magnetisation on a plane circular surface made from pure iron. The bearing capacity of the mechanism on the rod was initially 80 daN with a rod weight of 26.3 daN. This deteriorated progressively over time. The bearing surface of the rod and the electromagnet became contaminated with a deposit of sodium oxides and metallic particles, thus creating an air gap. This reached a figure of 36 daN in 2005 and was deemed not to be sufficient to prevent the rod from dropping at the wrong time during reactor operation. The Power Plant thus decided to replace the rod mechanism in the reactor in an initial phase, followed by the control rod itself. As the Phenix Power Plant had no spare control rods left, they initiated a 'salvage' plan, over two stages, for the rod removed from the reactor and placed in the fuel storage drum: - Inspection of the bearing surface of the rod by means of a borescope to check whether the rod could be salvaged, - A cleaning operation on the bearing face and checks on the bearing capacity of the rod. The operation is subject to very stringent requirements: the rod must not be taken out of the sodium to ensure that it can be reused in the reactor. The operation must thus take place in the fuel storage drum where there are no facilities for such an operation and where operating conditions are very hostile: high temperatures (the sodium in the fuel storage drum is at a temperature of 150 deg. C, high dose rate (3 mGy/h on the bearing surface) and the bearing surface is submerged

  16. Hydraulically driven control rod concept for integral reactors: fluid dynamic simulation and preliminary test

    International Nuclear Information System (INIS)

    Ricotti, M.E.; Cammi, A.; Lombardi, C.; Passoni, M.; Rizzo, C.; Carelli, M.; Colombo, E.

    2003-01-01

    The paper deals with the preliminary study of the Hydraulically Driven Control Rod concept, tailored for PWR control rods (spider type) with hydraulic drive mechanism completely immersed in the primary water. A specific solution suitable for advanced versions of the IRIS integral reactor is under investigation. The configuration of the Hydraulic Control Rod device, made up by an external movable piston and an internal fixed cylinder, is described. After a brief description of the whole control system, particular attention is devoted to the Control Rod characterization via Computational Fluid Dynamics (CFD) analysis. The investigation of the system behavior, including dynamic equilibrium and stability properties, has been carried out. Finally, preliminary tests were performed in a low pressure, low temperature, reduced length experimental facility. The results are compared with the dynamic control model and CFD simulation model, showing good agreement between simulations and experimental data. During these preliminary tests, the control system performs correctly, allowing stable dynamic equilibrium positions for the Control Rod and stable behavior during withdrawal and insertion steps. (author)

  17. Controlling a nuclear reactor with dropped control rods

    International Nuclear Information System (INIS)

    Mc Atee, K.R.; Alsop, B.H.

    1987-01-01

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

  18. RODDRP - A FORTRAN program for use in control rod calibration by the rod drop method

    International Nuclear Information System (INIS)

    Wilson, W.E.

    1972-01-01

    The different methods to measure reactivity which are applicable to control rod calibration are discussed. They include: 1) the positive period method, 2) the rod drop method, 3) the source-jerk method, 4) the rod oscillation method, and 5) the pulsed neutron method. The instrument setup used at WSU for rod drop measurements is presented. To speed up the analysis of power fall-off trace, a FORTRAN IV program called RODDRP was written to simultaneously solve the in-hour equation and relative neutron flux. The procedure for calculating the worth of the rod that produced the power trace is given. The reactivity for each time relative flux point is obtained. Conclusions about the status of the equipment are made

  19. Shock analysis on hydraulic drive control rod during scram

    International Nuclear Information System (INIS)

    Song Wei; Qin Benke; Bo Hanliang

    2013-01-01

    Control rod hydraulic drive mechanism (CRHDM) is a new invention of Institute of Nuclear and New Energy Technology of Tsinghua University. The hydraulic absorber buffers the control rod when it scrams. The control rod fast drop impact experiment was conducted and the key parameters of control rod hydraulic buffering performance were obtained. Based on the test results and according to D'Alembert principle, the maximum inertial impact force on the control rod during the fast drop period was applied as equivalent static load force on the control rod. The deformations and stress distributions on the control rod in this worst case were calculated by using finite element software ABAQUS. Calculation results were compared with the experiment results, and it was verified that nonlinear transient dynamics analysis in this problem can be simplified as static analysis. Damage criterion of the control rod fast drop impact process was also given. And it lays foundation for optimal design of the control rod and hydraulic absorber. (authors)

  20. Rapid and accurate control rod calibration measurement and analysis

    International Nuclear Information System (INIS)

    Nelson, George W.; Doane, Harry J.

    1990-01-01

    In order to reduce the time needed to perform control rod calibrations and improve the accuracy of the results, a technique for a measurement, analysis, and tabulation of integral rod worths has been developed. A single series of critical rod positions are determined at constant low power to reduce the waiting time between positive period measurements and still assure true stable reactor period data. Reactivity values from positive period measurements and control rod drop measurements are used as input data for a non-linear fit to the expected control rod integral worth shape. With this method, two control rods can be calibrated in about two hours, and integral and differential calibration tables for operator use are printed almost immediately. Listings of the BASIC computer programs for the non-linear fitting and calibration table preparation are provided. (author)

  1. Optimal control rod programs in power reactors

    International Nuclear Information System (INIS)

    Fadilah, S.M.; Lewins, J.

    1975-01-01

    Control rod programming is investigated with respect to optimising the power peaking factor and hence the utilisation of a nuclear reactor. A simplified diffusion model, initially with a finite number of regions, in cylindrical geometry, is used to enable optimal trajectories to be completely synthesised. The average discharge burnup problem is posed both as an external and as an internal optimisation. The connection between optimum power shape and the maximisation of the average discharge burnup is explored in a wider context. It is shown that optimum trajectories combine an initial singular solution of the Haling type with a terminal bang-bang solution. An extension to a higher number of regions and, on passing to the limit, to a diffusion model, provides an alternative proof of Haling's principle without the restriction to monotonic reactivity decrease with burnup. Numerical results in the two-region model are given to show the general scope of optimisation available. (author)

  2. Thermal hydraulic performance of naturally aspirated control rod housing assemblies

    International Nuclear Information System (INIS)

    Geiger, G.T.; Randolph, H.W.; Paik, I.K.; Foti, D.J.

    1992-01-01

    Savannah River Site reactors are comprised of heat generating fuel/target assemblies, control rods which regulate reactor power, and heavy water which acts as the coolant and as a moderator. The fuel/target assemblies are cooled by the downflow of heavy water while the control rods are cooled via upflow. Five control rods are grouped with two safety rods in seven-channel assemblies called septifoils. Under normal operating conditions, the reactor power level, radial shape flux and axial power flux are regulated by the positioning of the control rods. The control rods are solid rods of a lithium-aluminum alloy with an thin aluminum outer sheath. Lithium is a good absorber of neutrons and, thus control rod temperatures rise with reactor power. At conditions of sufficiently high reactor power and degraded coolant flow, the control rods could heat sufficiently to cause a metallurigical failure of the sheath leading to molten material coming in contact with water and the possibility of a steam explosion. An accident has been postulated as part of the analysis involving the safety upgrade of Savannah River Site reactors in which the housing is not seated on the pin. Coolant from the upflow pin would not be directed into the housing but, into the moderator space surrounding the housing. Only naturally aspirated cooling due to buoyancy effects would be available to cool the control rods and the coolant mass flow rate would drop significantly from its nominal value. In this study, the mechanisms and limits of cooling heated rods housed in an unseated septifoil are addressed. Experiments were conducted on a shortened, prototypic housing with electrically heated rods to gain an understanding of the phenomena governing the cooling in such a case and develop data which can be used to evaluate predictive models. These experiments are described, their results discussed, and the predictions of current models is presented

  3. Hydraulic pressure control unit for control rod drive

    International Nuclear Information System (INIS)

    Watabe, Yukio.

    1990-01-01

    The pressure invention concerns a hydraulic pressure control unit for control rod drives in BWR type reactors. The space above a floating piston possessed by an accumulator and the housing of control rod drives are connected by means of a pipeline. The pipeline has a scram valve which is opened upon occurrence of reactor scram. A pump is disposed between the accumulator and the scram valve for communicating a discharge port to apply a high pressure water to the accumulator. According to the present invention, a control unit is disposed between the scram valve and the housing of the control rod drives in the hydraulic pressure control unit for maintaining the cross sectional area of the flow channel of the pipeline to a usual size when the pressure in a pressure vessel is under a rated operation pressure, while limiting the cross sectional area of the flow channel when the pressure is lower than that in the rated operation. Thus, whole insertion of the control rod substantially at a constant speed is enabled irrespective of the level of the pressure in the pressure vessel. (I.S.)

  4. 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.

  5. Optimized Control Rods of the BR2 Reactor

    International Nuclear Information System (INIS)

    Kalcheva, Silva; Koonen, E.

    2007-01-01

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

  6. Study on evaluating the reactivity worth of the control rods of the PWR 900 MWe

    International Nuclear Information System (INIS)

    Phan Quoc Vuong; Tran Vinh Thanh; Tran Viet Phu

    2015-01-01

    Control rods of a nuclear reactor are divided into two groups: shut down and power control. Reactivity worth of the control rods depends nonlinearly on the rods' compositions and positions where the rods are inserted into the core. Therefore, calculation of control rod worth is of high important. In this study, we calculated the reactivity worth of the power control rod bank of the Mitsubishi PWR 900 MWe. The results are integral and differential worth calibration of the control rods. (author)

  7. Fast, accurate control rod calibration using a programmable desk calculator

    International Nuclear Information System (INIS)

    Naugle, N.W.; Randall, John D.

    1972-01-01

    In an attempt to develop a simple least squares program for the rapid calibration of control rods it was necessary to verify that all rods, with the exception of the transient rod, could be accurately defined by a single analytical expression. Since the vertical flux distribution in the core region follows a cosine function, a cosine squared variation was tested. The solution which involves the inversion of a 3 x 3 matrix is performed using a Hewlett Packard Model 9100B programmable desk calculator. The least squares program was applied to a number of control rod calibrations that had previously been analyzed by hand. The agreement was excellent in all cases

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

    International Nuclear Information System (INIS)

    John, T.M.; Mohanakrishnan, P.; Mahalakshmi, B.; Singh, R.S.

    1988-01-01

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

  9. Control rod homogenization in heterogeneous sodium-cooled fast reactors

    International Nuclear Information System (INIS)

    Andersson, Mikael

    2016-01-01

    The sodium-cooled fast reactor is one of the candidates for a sustainable nuclear reactor system. In particular, the French ASTRID project employs an axially heterogeneous design, proposed in the so-called CFV (low sodium effect) core, to enhance the inherent safety features of the reactor. This thesis focuses on the accurate modeling of the control rods, through the homogenization method. The control rods in a sodium-cooled fast reactor are used for reactivity compensation during the cycle, power shaping, and to shutdown the reactor. In previous control rod homogenization procedures, only a radial description of the geometry was implemented, hence the axially heterogeneous features of the CFV core could not be taken into account. This thesis investigates the different axial variations the control rod experiences in a CFV core, to determine the impact that these axial environments have on the control rod modeling. The methodology used in this work is based on previous homogenization procedures, the so-called equivalence procedure. The procedure was newly implemented in the PARIS code system in order to be able to use 3D geometries, and thereby be take axial effects into account. The thesis is divided into three parts. The first part investigates the impact of different neutron spectra on the homogeneous control-rod cross sections. The second part investigates the cases where the traditional radial control-rod homogenization procedure is no longer applicable in the CFV core, which was found to be 5-10 cm away from any material interface. In the third part, based on the results from the second part, a 3D model of the control rod is used to calculate homogenized control-rod cross sections. In a full core model, a study is made to investigate the impact these axial effects have on control rod-related core parameters, such as the control rod worth, the capture rates in the control rod, and the power in the adjacent fuel assemblies. All results were compared to a Monte

  10. 77 FR 43967 - Account Ownership and Control Report; Withdrawal

    Science.gov (United States)

    2012-07-26

    ... 15, 16, 17, et al. Account Ownership and Control Report; Withdrawal; Ownership and Control Reports... RIN 3038-AC63 Account Ownership and Control Report; Withdrawal AGENCY: Commodity Futures Trading... published for public comment a Notice of Proposed Rulemaking that proposed to collect certain account...

  11. ABWR-II Core Design with Spectral Shift Rods for Operation with All Control Rods Withdrawn

    International Nuclear Information System (INIS)

    Moriwaki, Masanao; Aoyama, Motoo; Anegawa, Takafumi; Okada, Hiroyuki; Sakurada, Koichi; Tanabe, Akira

    2004-01-01

    An innovative reactor core concept applying spectral shift rods (SSRs) is proposed to improve the plant economy and the operability of the 1700-MW(electric) Advanced Boiling Water Reactor II (ABWR-II). The SSR is a new type of water rod in which a water level is naturally developed during operation and changed according to the coolant flow rate through the channel. By taking advantage of the large size of the ABWR-II bundle, the enhanced spectral shift operation by eight SSRs allows operation of the ABWR-II with all control rods withdrawn. In addition, the uranium-saving factor of 6 to 7% relative to the reference ABWR-II core with conventional water rods can be expected due to the greater effect of spectral shift. The combination of these advantages means the ABWR-II with SSRs should be an attractive alternative for the next-generation nuclear reactor

  12. The development and validation of control rod calculation methods

    International Nuclear Information System (INIS)

    Rowlands, J.L.; Sweet, D.W.; Franklin, B.M.

    1979-01-01

    Fission rate distributions have been measured in the zero power critical facility, ZEBRA, for a series of eight different arrays of boron carbide control rods. Diffusion theory calculations have been compared with these measurements. The normalised fission rates differ by up to about 30% in some regions, between the different arrays, and these differences are well predicted by the calculations. A development has been made to a method used to produce homogenised cross sections for lattice regions containing control rods. Calculations show that the method also reproduces the reaction rate within the rod and the fission rate dip at the surface of the rod in satisfactory agreement with the more accurate calculations which represent the fine structure of the rod. A comparison between diffusion theory and transport theory calculations of control rod reactivity worths in the CDFR shows that for the standard design method the finite mesh approximation and the difference between diffusion theory and transport theory (the transport correction) tend to cancel and result in corrections to be applied to the standard mesh diffusion theory calculations of about +- 2% or less. This result applies for mesh centred finite difference diffusion theory codes and for the arrays of natural boron carbide control rods for which the calculations were made. Improvements have also been made to the effective diffusion coefficients used in diffusion theory calculations for control rod followers and these give satisfactory agreement with transport theory calculations. (U.K.)

  13. Nuclear reactor shutdown control rod assembly

    International Nuclear Information System (INIS)

    Bilibin, K.

    1988-01-01

    This patent describes a nuclear reactor having a reactor core and a reactor coolant flowing therethrough, a temperature responsive, self-actuated nuclear reactor shutdown control rod assembly, comprising: an upper drive line terminating at its lower end with a substantially cylindrical wall member having inner and outer surfaces; a lower drive line having a lower end adapted to be attached to a neutron absorber; a ring movable disposed about the outer surface of the wall member of the upper drive line; thermal actuation means adapted to be in heat exchange relationship with coolant in an associated reactor core and in contact with the ring, and balls located within the openings in the upper drive line. When reactor coolant approaches a predetermined design temperature the actuation means moves the ring sufficiently so that the balls move radially out from the recess and into the space formed by the second portion of the ring thereby removing the vertical support for the lower drive line such that the lower drive line moves downwardly and inserts an associated neutron absorber into an associated reactor core resulting in automatic reduction of reactor power

  14. Control rod driving hydraulic pressure device

    International Nuclear Information System (INIS)

    Ogawa, Masahide.

    1993-01-01

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

  15. Electromotor control rod drive for nuclear reactors

    International Nuclear Information System (INIS)

    Baker, S.M.

    1975-01-01

    The positioning of a control rod arranged in a pressure vessel takes place with a drive. This protrudes out of the pressure vessel through a support and is formed from a rotating field motor with energy source, e.g. alternating current connection. Its stator surrounds a section of a pressure casing which covers the length of the drive. The rotor is arranged in the pressure casing and interacts with a shaft lying in the rotation axis. Furthermore, segments are hinged on it, each of which forms two arms of a rocker. Each segment can be revolved against a storing force in a plane containing the rotation axis, through the stator field acting on one of the rocker arms. In order that the drive motor is automatically blocked should the electricity supply fail, the other rocker arm can be connected with a fixed cased component of the drive having the effect of a friction break or a form-locking mechanical catch. (DG/LH) [de

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

    International Nuclear Information System (INIS)

    Telford, A.R.R.

    1975-10-01

    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)

  17. Review of FFTF and CRBRP control rod systems designs

    International Nuclear Information System (INIS)

    Pitterle, T.A.; Lagally, H.O.

    1977-01-01

    The evolution of the primary control rod system design for FFTF and CRBR, beginning with the initial choice of the basic concepts, is described. The significant component and systems tests are reviewed together with the test results which referenced the development of the CRBR primary control rod system design. Modifications to the concepts and detail designs of the FFTF control rod system were required principally to satisfy the requirements of CRBR, and at the same time incorporating design refinements shown desirable by the tests

  18. Post irradiation examination of control rod assembly of FBTR

    International Nuclear Information System (INIS)

    Anandaraj, V.; Raghu, N.; Venkiteswaran, C.N.; Visweswaran, P.; Vijayakumar, Ran; Jayaraj, V.V.; Padmaprabu, P.; Saravanan, T.; Philip, John; Muralidharan, N.G.; Joseph, Jojo; Kasiviswanathan, K.V.

    2010-01-01

    Six control rods with boron carbide pellets are used in FBTR for shutdown and control of reactor power. One control rod after being subjected to a fluence level of 7.2 x 10 22 n/cm 2 was received for post irradiation examination (PIE) to assess its irradiation behavior and to investigate the incident of dropping of control rod. Examinations carried out include precise dimensional measurements to investigate the possibility of interference between the control rod and outer sheath, Neutron radiography and x-radiograph to assess the integrity of the boron carbide pellets and other internals, density measurements to assess the swelling behaviour of boron carbide pellets and metallographic examinations to study the cracking behaviour and microstructural changes in the pellet and the clad. Depletion of B 10 in the pellet was studied using time of flight mass spectrometry. The paper highlights the examinations and results of the PIE carried out. (author)

  19. Impact loading of a BWR control rod during braking

    International Nuclear Information System (INIS)

    Heeschen, U.

    1977-01-01

    In an emergency case the control rods of a boiling water reactor are shot into the RPV from below against the weight of the rods with drive motors. According to the position of the control rods between the fuel elements the rods can reach in that case velocities up to 4 m/s. The moved masses of the control rods and of the pistons (both of them are connected by a coupling) are braked through a cup spring which transfers its forces to the RPV-bottom sphere. The spring has to be designed that in this case tthe complete kinetic energy of he control rods of about 1000Nm can be taken up. The spring power and the inertia of the moved masses cause extremely high loadings during and shortly after the impact onto the spring. The shock-like loading propagates along the whole rod at the speed of sound, and this is also the reason why the weaker cross-sections have to endure considerable short-term stress peaks. (Auth.)

  20. Computation of reactor control rod drop time under accident conditions

    International Nuclear Information System (INIS)

    Dou Yikang; Yao Weida; Yang Renan; Jiang Nanyan

    1998-01-01

    The computational method of reactor control rod drop time under accident conditions lies mainly in establishing forced vibration equations for the components under action of outside forces on control rod driven line and motion equation for the control rod moving in vertical direction. The above two kinds of equations are connected by considering the impact effects between control rod and its outside components. Finite difference method is adopted to make discretization of the vibration equations and Wilson-θ method is applied to deal with the time history problem. The non-linearity caused by impact is iteratively treated with modified Newton method. Some experimental results are used to validate the validity and reliability of the computational method. Theoretical and experimental testing problems show that the computer program based on the computational method is applicable and reliable. The program can act as an effective tool of design by analysis and safety analysis for the relevant components

  1. Design requirement on KALIMER control rod assembly duct

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-03-01

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

  2. Design requirement on KALIMER control rod assembly duct

    International Nuclear Information System (INIS)

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

    1998-03-01

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

  3. Safety coupling for a control rod of a nuclear reactor

    International Nuclear Information System (INIS)

    Mindnich, F.R.; Friedrichs, H.; Schoettle, J.

    1978-01-01

    A coupling is presented between a control rod and the drive shafft arranged below. The construction of this coupling is designed in such a way that the usual sealing maesures against the escape of coolant are reduced. (TK) [de

  4. Components inspection of Monju, a sodium bonded type control rod

    International Nuclear Information System (INIS)

    Harada, Kiyoshi; Matsushita, Yuichi; Lee, Chunchan; Abe, Hideaki; Watahiki, Naohisa

    2002-03-01

    This Report addresses a result of a sodium test conducted on components of a Double Poral Filter Sodium Bonded Type Control Rod that is expected to be a next generation, long life Control Rod. Upper and lower Poral Filter Sodium Bonded Type Control Rod components were mocked up to conduct a sodium test. During the test, sodium chargeability, formation of Gas Plenum at the upper part of the components, sodium drain-ability and NaOH clean-ability were recognized under actual plant condition. The following are results obtained: (1) Sodium Chargeability at Control Rod Insertion to EVST. Sodium was charged into the components when the mocked-up was inserted in sodium of 190degC, with insertion speed of 6 m/min which is an actual insertion speed to EVST. (2) Formation of Upper Gas Plenum by Helium Gas generated in Control Rod Components Gas Plenum formation within deviation of 9% was confirmed by releasing helium gas into the mocked-up which is immersed in sodium of 620degC and 190degC. Length of Gas Plenum is confirmed to be retained in certain length even if helium gas is further released into formed Gas Plenum. (3) Sodium Drain-ability of Control Rod Components when Drawing from EVST. Drain-ability was confirmed to be sufficient and no sodium residue was found in the mocked-up when the mocked-up was drawn out from sodium of 190degC, with drawing speed of 6 m/min which is an actual drawing speed from EVST. (4) Clean-ability of NaOH Solution against Sodium Residue in Control Rod Components. Sodium and NaOH solution reacted calmly, however, clean-ability was not sufficient. When Sodium fully remained in Control Rod Components, it made circulation of NaOH solution not enough. (author)

  5. Normal temperature ventilating homogenity test of control rod drive mechanism

    International Nuclear Information System (INIS)

    Zhu Longxing.

    1987-01-01

    This paper describes the cooling for the control rod drive mechanism. It emphatically introduces some problems which must be considered in the test on ventilating homogenity under normal temperature for the control rod drive mechanism at the top of a reactor such as the selection of cooling shroud assemblies, installed position of ducts, volume of the static pressure container, etc. The test data of blowing-in and induced draft are compared and analysed

  6. Computer program for automatic generation of BWR control rod patterns

    International Nuclear Information System (INIS)

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

    1990-01-01

    A computer program named OCTOPUS has been developed to automatically determine a control rod pattern that approximates some desired target power distribution as closely as possible without violating any thermal safety or reactor criticality constraints. The program OCTOPUS performs a semi-optimization task based on the method of approximation programming (MAP) to develop control rod patterns. The SIMULATE-E code is used to determine the nucleonic characteristics of the reactor core state

  7. Reactor core and control rod assembly in FBR type reactor

    International Nuclear Information System (INIS)

    Fujimura, Koji; Kawashima, Katsuyuki; Itooka, Satoshi.

    1993-01-01

    Fuel assemblies and control rod assemblies are attached respectively to reactor core support plates each in a cantilever fashion. Intermediate spacer pads are disposed to the lateral side of a wrapper tube just above the fuel rod region. Intermediate space pads are disposed to the lateral side of a control rod guide tube just above a fuel rod region. The thickness of the intermediate spacer pad for the control rod assembly is made smaller than the thickness of the intermediate spacer pad for the fuel assembly. This can prevent contact between intermediate spacer pads of the control guide tube and the fuel assembly even if the temperature of coolants is elevated to thermally expand the intermediate spacer pad, by which the radial displacement amount of the reactor core region along the direction of the height of the control guide tube is reduced substantially to zero. Accordingly, contribution of the control rod assembly to the radial expansion reactivity can be reduced to zero or negative level, by which the effect of the negative radial expansion reactivity of the reactor is increased to improve the safety upon thermal transient stage, for example, loss of coolant flow rate accident. (I.N.)

  8. Calculation method for control rod dropping time in reactor

    International Nuclear Information System (INIS)

    Nogami, Takeki; Kato, Yoshifumi; Ishino, Jun-ichi; Doi, Isamu.

    1996-01-01

    If a control rod starts dropping, the dropping speed is rapidly increased, then settled substantially constant, rapidly decreased when it reaches a dash pot. A second detection signal generated by removing an AC component from a first detection signal is differentiated twice. The time when the maximum value among the twice differentiated values is generated is determined as a time when the control rods starts dropping. The time when minimum value among the twice differentiated values is generated is determined as a time when the control rod reaches the dash pot of the reactor. The measuring time within a range from the time when the control rod starts dropping to the time when the control rod reaches the dash pot of the reactor is determined. As a result, processing for the calculation of the dropping start time and dash pot reaching time of the control rod can be automatized. Further, it is suffice to conduct differentiation twice till the reaching time, which can facilitate the processing thereby enabling to determine a reliable time range. (N.H.)

  9. Device and method of cooling control rod drives

    International Nuclear Information System (INIS)

    Togashi, Hidetoshi; Mase, Noriaki; Matsumura, Yuichi.

    1985-01-01

    Purpose: To prevent the generation of local temperature rise depending on the reactor core position of the control rod drives and control the temperature to an averaged state in BWR type reactors. Method: Control rod drives having a large charging length of the housing in the pressure vessel involve such a factor that the temperature of the control rod drives is increased by the synergistic effect due to the radiation heat from the reactor core and to the unevenness of the cooling water flow rate, which renders an appropriate temperature control difficult for the reactor core position. A cooling water flow rate controlling device having a restriction mechanism is disposed on the cooling water feed path for each of the hydraulic control units of the control rod drives, so that flow rate to the control rod drives is increased at the center of the reactor core and decreased at the periphery thereof. As a result, average temperature state can be set, temperature increase due to cloggings can be prevented and the thermal effect can be eliminated to thereby improve the reliability. (Moriyama, K.)

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

    International Nuclear Information System (INIS)

    Malang, S.

    1975-11-01

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

  11. Control Rod Driveline Reactivity Feedback Model for Liquid Metal Reactors

    International Nuclear Information System (INIS)

    Kwon, Young-Min; Jeong, Hae-Yong; Chang, Won-Pyo; Cho, Chung-Ho; Lee, Yong-Bum

    2008-01-01

    The thermal expansion of the control rod drivelines (CRDL) is one important passive mitigator under all unprotected accident conditions in the metal and oxide cores. When the CRDL are washed by hot sodium in the coolant outlet plenum, the CRDL thermally expands and causes the control rods to be inserted further down into the active core region, providing a negative reactivity feedback. Since the control rods are attached to the top of the vessel head and the core attaches to the bottom of the reactor vessel (RV), the expansion of the vessel wall as it heats will either lower the core or raise the control rods supports. This contrary thermal expansion of the reactor vessel wall pulls the control rods out of the core somewhat, providing a positive reactivity feedback. However this is not a safety factor early in a transient because its time constant is relatively large. The total elongated length is calculated by subtracting the vessel expansion from the CRDL expansion to determine the net control rod expansion into the core. The system-wide safety analysis code SSC-K includes the CRDL/RV reactivity feedback model in which control rod and vessel expansions are calculated using single-nod temperatures for the vessel and CRDL masses. The KALIMER design has the upper internal structures (UIS) in which the CRDLs are positioned outside the structure where they are exposed to the mixed sodium temperature exiting the core. A new method to determine the CRDL expansion is suggested. Two dimensional hot pool thermal hydraulic model (HP2D) originally developed for the analysis of the stratification phenomena in the hot pool is utilized for a detailed heat transfer between the CRDL mass and the hot pool coolant. However, the reactor vessel wall temperature is still calculated by a simple lumped model

  12. Calculation of the effectiveness of manual control rods for the reactor of Ignalina NPP Unit 2

    International Nuclear Information System (INIS)

    Bubelis, E.; Pabarcius, R.

    2001-01-01

    On the basis of one of the recent databases of the reactor of Ignalina NPP Unit 2, calculations of the effectiveness of separate manual control rods, groups of manual control rods and axial characteristic of effectiveness of separate manual control rods were performed. The results of the calculations indicated, that all analyzed separate manual control rods have approximately the same effectiveness, which doesn't depend on the location of a control rod in the reactor core layout Manual control rod of the new design has about 10% greater effectiveness than manual control rod of the old design. (author)

  13. Disassembling and cleaning apparatus for control rod drives

    International Nuclear Information System (INIS)

    Yoshida, Tomiji; Sasaki, Masayoshi; Numata, Nobumasa; Oowada, Masataka; Arazoe, Masatoshi.

    1981-01-01

    Purpose: To shorten the working period, as well as decrease the number of operators and radiation exposure doses for disassembling and cleaning of the control rod drives. Constitution: The apparatus comprises a cleaning tank, clamp units, inner and outer surface cleaning nozzles, a high pressure pump, a drying tank unit, electronic components and the likes. This enables to render the various works automatic, such as wash out of outer and inner filters and strainers at high radiation dose levels, extraction and cleaning of a piston tube and an index tube, cleaning for the control rod drive main body, and cleaning and drying of small components, whereby the disassembling, cleaning and assembling works for the control rod drives, among all, the disassembling and cleaning works resulting relatively high radiation exposure doses to the operator can be made in a remote control and semi-automatic manner. (Furukawa, Y.)

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

    International Nuclear Information System (INIS)

    Sanda

    2010-01-01

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

  15. Design of controller for control rod of research reactors

    International Nuclear Information System (INIS)

    Abou-Zaid, R.M.F.M

    2008-01-01

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

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

    International Nuclear Information System (INIS)

    Sease, J.D.

    1998-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Sease, J.D.

    1998-03-01

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

  18. Control device for handling device of control rod drive

    International Nuclear Information System (INIS)

    Sasaki, Toshiya

    1998-01-01

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

  19. 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

  20. CEA data and methods for control rod calculations

    International Nuclear Information System (INIS)

    Salvatores, M.

    1988-01-01

    Methods and data used at CEA for LMFBR control rod calculations are presented. The performances of both the design methods and of the refined methods for the SUPER PHENIX start-up experiments are given. New developments for future core design and operation are also reported, together with the role of integral experiments to validate them. (author). 15 refs, 4 figs, 11 tabs

  1. Determination of the control rod worth for research reactors

    International Nuclear Information System (INIS)

    Aldama, D.L.; Gual, M.R.

    2000-01-01

    Nowadays there is a big interest in developing neutronic analysis methods for research reactor and particularly for the determination of the control rods worth under different operation conditions and core configurations. The reactivity associated with the control rods is of interest in the shutdown margin and in calculations of possible abnormal conditions related to reactivity accidents. For theses studies several computer codes have been developed. The present work is aimed at the validation of the calculation methods of the Nuclear Technology Center of Cuba. For this purpose, in order to evaluate the safety of this type of installations, the reactivity worth of the control rods of the cylindrical configuration of the Brazilian critical assembly IPEN/MB-01 is determined. These calculations, however, are a relatively complex task that requires the use of three-dimensional models. Because of this, the validation of the calculation methods used for this purpose is of great importance. In fact, it is one of the requirements called upon by the quality assurance programs for the development, maintenance and utilization of the calculation codes used in safety analysis. For the calculation of control rod worth the lattice code WIMS-D/4 [8] and the diffusion code SNAP-3D [9] were used. This work presents the obtained results and gives a comparison with the experimental values

  2. Possibilities and limits of the reactivity determination of control rods

    International Nuclear Information System (INIS)

    Buenemann, D.

    1975-01-01

    Basic physical facts of the reactivity determination of control rods are presented. A survey of currrently applied methods is given, and the drawbacks of the various methods are pointed out. Special problems are presented by the interpretation of highly subcritical assemblies which are not really important in practical reactor operation but desirable for a consistant comparison between theory and experiments. (orig./AK) [de

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

    International Nuclear Information System (INIS)

    West, G.

    1982-11-01

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

  4. Electromagnetic design calculation of the control rod drive mechanism

    International Nuclear Information System (INIS)

    Zhu Qirong; Zhu Jingchang

    1991-01-01

    Electromagnetic design calculation of the step-by-step magnetic jacking control rod drive mechanism includes magnetic field force calculation and design calculation of magnetomotive force for three electromagnetic iron and their coilds. The basic principle and method of electromagnetic design calculation had been expounded to take the lift magnet and lift coil for example

  5. Boiling water reactor radiation shielded Control Rod Drive Housing Supports

    Energy Technology Data Exchange (ETDEWEB)

    Baversten, B.; Linden, M.J. [ABB Combustion Engineering Nuclear Operations, Windsor, CT (United States)

    1995-03-01

    The Control Rod Drive (CRD) mechanisms are located in the area below the reactor vessel in a Boiling Water Reactor (BWR). Specifically, these CRDs are located between the bottom of the reactor vessel and above an interlocking structure of steel bars and rods, herein identified as CRD Housing Supports. The CRD Housing Supports are designed to limit the travel of a Control Rod and Control Rod Drive in the event that the CRD vessel attachement went to fail, allowing the CRD to be ejected from the vessel. By limiting the travel of the ejected CRD, the supports prevent a nuclear overpower excursion that could occur as a result of the ejected CRD. The Housing Support structure must be disassembled in order to remove CRDs for replacement or maintenance. The disassembly task can require a significant amount of outage time and personnel radiation exposure dependent on the number and location of the CRDs to be changed out. This paper presents a way to minimize personal radiation exposure through the re-design of the Housing Support structure. The following paragraphs also delineate a method of avoiding the awkward, manual, handling of the structure under the reactor vessel during a CRD change out.

  6. Control rod blow out protection system

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  7. Anti-ejection system for control rod drives

    International Nuclear Information System (INIS)

    Matthews, J.C.

    1977-01-01

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

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

    International Nuclear Information System (INIS)

    Jahnke, S.; Ple, P.

    1989-01-01

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

  9. Control rod drive mechanism with shock absorber for nuclear reactor

    International Nuclear Information System (INIS)

    Chevereau, G.

    1989-01-01

    The mechanism usable in a PWR has a shaft carrying the bar vertically displaceable in the reactor internals and a dash pot with a hydraulic cylinder and a piston. The cylinder has a large diameter perforated upper section to the cylinder, a small diameter lower section, a piston traversed by the control rod sized to fit into the upper section and forced downwards when the control descends. The shock absorbing chamber is defined between the piston and the upper section [fr

  10. Production control in HTGR fuel rod fabrication

    International Nuclear Information System (INIS)

    Downing, D.J.; Bailey, M.

    1979-06-01

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

  11. Drive mechanism nuclear reactor control rod

    International Nuclear Information System (INIS)

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

    1978-01-01

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

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

    International Nuclear Information System (INIS)

    Gil, Choong-Sup; Okumura, Keisuke; Ishiguro, Yukio

    1991-11-01

    Investigations were made for the heterogeneity effects caused by insertion or withdrawal of a B 4 C control rod on fine structure of reaction rates distributions in a tight pitched PuO 2 -UO 2 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)

  13. Improvement to the control rod drive of a nuclear reactor

    International Nuclear Information System (INIS)

    Desfontaines, Guy.

    1981-01-01

    Improvement to the devices that move the control rods of a nuclear reactor. The slow movements of the rods are generally carried out by screw and nut gear, the nut being blocked as to rotation and the screw as to translation movement. Additionally, a mechanism enables the control rods to be inserted rapidly by release of the screw and nut gear, the nut remaining constantly in gear with the screw. The presence of extra poles and coils under the stator of the actuating motor of the screw add length and weight to the mechanism and hence increase the strains and deformations which affect the latter in the event of an earthquake. The device of the invention makes it possible to overcome this drawback and leads to a more simple mechanism. It is characterized in that the rotor of the motor actuating the screw is also provided with clamps, in its high position, controlled by electromagnetic action as from the coils of the actuating motor stator so that they are in the closed position on the screw when the stator is powered and in the open position when it is no longer so, in order to allow the screw and nut assembly drop, and in that it includes a device to lock the clamps, enabling these to be kept in the open position when the control screw is not in the high holding position [fr

  14. Control-rod removal-blocking monitor (RBM)

    International Nuclear Information System (INIS)

    Matsumiya, Shoichi.

    1983-01-01

    Purpose: To provide an apparatus capable of checking with ease whether predetermined local power region neutron signals are selected correctly depending on selected control rods. Constitution: A computer is adapted to read the entire LPRM (local power range monitor) signals and a matrix calculation function is further provided for obtaining predetermined selected LPRM signals from the entire LPRM signals by the selected control rods. Then, a comparison function is provided for comparing the above mentioned selected LPRM signals read from RBM and signals obtained in the inside of the computer. The comparison functions judge whether both of them coincide to each other or not and, if it is judged not coincide, produces an abnormal-operation signal. Since the operator can thus recognize the failure in the RBM and carry out repairement directly, the reliability can be improved. (Seki, T.)

  15. Auxiliary device for the assembly of cruciform BWR control rods

    International Nuclear Information System (INIS)

    Lippert, H.J.

    1980-01-01

    The mounting auxiliary equipment has got a frame of the size of a core cell that can be put on the core grid. Diagonally arranged link bodies serve to center the cross-shaped control rod. The mounting auxiliary equipment consists of a single piece and therefore facilitates manipulation for which a diagonal handle is provided at the upper end. It may be swung aside. (RW) [de

  16. Auxiliary device for the assembly of cruciform BWR control rods

    International Nuclear Information System (INIS)

    Lippert, H.J.

    1982-01-01

    The mounting auxiliary equipment has got a frame of the size of a core cell that can be put on the core grid. Diagonally arranged link bodies serve to center the cross-shaped control rod. The mounting auxiliary equipment consists of a single piece and therefore facilitates manipulation for which a diagonal handle is provided at the upper end. It may be swung aside. (orig./HP)

  17. Control rod effects with plutonium recycle in a PWR

    International Nuclear Information System (INIS)

    Nash, G.; Muehl, G.J.; Gibson, I.H.

    1979-03-01

    A study has been made on a PWR loaded partly and wholly with plutonium to determine the changes in shutdown margin compared with an enriched uranium core. Lattice calculations are used to generate cell constants for core calculations. Three fuel loadings were considered, all uranium, 30% (approximately) of the assemblies plutonium in natural uranium, and all plutonium. The equilibrium fuel management schemes adopted in each case are based on the standard three cycle equal size batch scheme. Detailed calculations of power and irradiation distributions through the cycles have been carried out to provide a starting point for the control rod worth and requirement calculations. Control rod worths are reduced in a plutonium core because of the harder spectrum and higher fuel absorption cross sections. Furthermore, the control rod requirements for shutdown increase because of the increase in fuel and moderator temperature coefficients. This results in a reduction in shutdown margin. The magnitude of these changes is fully analysed in the report. The significance of these reductions depends on the detail of the safety argument but reductions of these sizes are unlikely to be acceptable. The data provided in this report could be used to give a first estimate of the plutonium loading acceptable given the safety assessment of the normal uranium core. (U.K.)

  18. Fabrication Of Bar graph Assembly For Control Rod Indicator Of Reactor Kartini

    International Nuclear Information System (INIS)

    Supriyanto

    1996-01-01

    The Bar graph instrument for monitor control rod position have been made. This instrument is located on the panel diagram in the control room of reactor Kartini. In this instrument, every control rod position is indicated by 10 LED on, that can move up or down as moving of control rod, total 20 LED is needed one assembly that is 10 LED for full down position and 10 LED for full up position. This bar graph can only monitor the control rod position every 10% of maximum control rod movement

  19. Development of control system of coating of rod hydraulic cylinders

    Science.gov (United States)

    Aizhambaeva, S. Zh; Maximova, A. V.

    2018-01-01

    In this article, requirements to materials of hydraulic cylinders and methods of eliminating the main factors affecting the quality of the applied coatings rod hydraulic cylinders. The chromium plating process - one of ways of increase of anti-friction properties of coatings rods, stability to the wear and corrosion. The article gives description of differences of the stand-speed chromium plating process from other types of chromium plating that determines a conclusion about cutting time of chromium plating process. Conducting the analysis of technological equipment suggested addressing the modernization of high-speed chromium plating processes by automation and mechanization. Control system developed by design of schematic block diagram of a modernized and stand-speed chromium plating process.

  20. The behaviour of control rod absorber under irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Bourgoin, J. [Electricite de France, Avoine (France). Groupe des Laboratoires; Couvreur, F.; Gosset, D. [CEA-Saclay, Service d' Etude des Materiaux Irradies, 91191, Gif-sur-Yvette (France); Defoort, F.; Monchanin, M. [Framatome Nuclear Fuel, 10 rue J. Recamier 69456, Lyon (France); Thibault, X. [Electricite de France, SEPTEN, 12-14 avenue Dutrievoz, 69628, Villeurbanne (France)

    1999-12-01

    Increase of rod diameters and cracking of PWR control rod claddings may occur in operation. In order to understand the contribution of the absorber properties to this damage, EDF and FRAMATOME launched a programme of examinations concerning the silver-indium-cadmium alloy constituting the absorber bars. Density measurements and microstructural investigations such as micrography, microanalysis were carried out in the EDF Hot Laboratory, X-ray diffraction was performed by CEA. The results show that transmutations induce chemical modifications inside the FCC alloy and, further, formation of an HCP phase similar to the {zeta} phase of the silver alloys. The chemical and crystallographic changes account for the major part of the absorber swelling. (orig.)

  1. Analysis of Explanted Magnetically Controlled Growing Rods From Seven UK Spinal Centers.

    Science.gov (United States)

    Joyce, Thomas J; Smith, Simon L; Rushton, Paul R P; Bowey, Andrew J; Gibson, Michael J

    2018-01-01

    Analysis of explanted MAGnetic Expansion Control (MAGEC) growing rods. To analyze explanted MAGEC rods used in management of early onset scoliosis and identify the mode of failure in such cases. Magnetically controlled growing rods are increasingly used as the option of choice for early onset scoliosis. However, being more complex than conventional growing rods they are perhaps more likely to succumb to multifarious failure modes. In addition, metallosis has been reported around failed MAGEC rods. Explanted MAGEC rods from seven UK spinal centers were obtained for independent analysis. Thirty-four MAGEC rods, from 18 children, explanted for reasons including failure of rod lengthening and maximum rod distraction reached, were cut open to allow internal components to be evaluated and assessed. Externally, all MAGEC rods showed localized marks, which were termed "growth marks" as they indicated growth of the rod in vivo, on the extending bar component. After cutting open, titanium wear debris was found inside all 34 (100%) MAGEC rods. Ninety-one percent (31/34) of MAGEC rods showed measurable wear of the extending bar, towards the magnet end. Substantial damage to the radial bearing was seen inside 74% (25/34) of MAGEC rods while O-ring seal failure was seen in 53% (18/34) of cases. In 44% (15/34) of MAGEC rods the drive pin was fractured but this was felt to be an effect of rod failure, not a cause. The combination of high volumes of titanium wear debris alongside O-ring seal damage likely accounts for the metallosis reported clinically around some MAGEC rods. Based on this explant data, a failure mechanism in MAGEC rods due to the natural off axis loading in the spine was proposed. This is the largest data set reporting a complete analysis of explanted MAGEC rods to date. 4.

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

    International Nuclear Information System (INIS)

    Palomo, M.; Urrea, M.; Arnaldos, A.

    2011-01-01

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

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

    International Nuclear Information System (INIS)

    Curiel, M.; Palomo, M. J.; Urrea, M.; Arnaldos, A.

    2010-10-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-10-15

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

  5. 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)

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

    International Nuclear Information System (INIS)

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

    1974-01-01

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

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

    International Nuclear Information System (INIS)

    Kussmaul, G.

    1994-10-01

    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.) [de

  8. 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.

  9. Clinch River Breeder Reactor secondary control rod system

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  10. Control rod drive mechanism stator loss of coolant test

    International Nuclear Information System (INIS)

    Besel, L.; Ibatuan, R.

    1977-04-01

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

  11. Integral core degradation test with B4C control rod

    International Nuclear Information System (INIS)

    Windberg, P.; Nagy, I.; Matus, L.; Balasko, M.; Hozer, Z.; Czitrovszky, A.; Nagy, A.

    2001-01-01

    The CODEX-B4C VVER bundle test has been successfully performed on 25 th May 2001 in the framework of the COLOSS project of the EU 5 th FWP. The experiment was carried out according to a scenario selected in favour of methane formation. Degradation of control rod and fuel bundle took place at temperatures approx. 2000 oC, cooling down of the bundle was performed in steam atmosphere. The gas composition measurement indicated no methane production during the experiment. The on-line measured data are collected into a database and are available for code validation and development.(author)

  12. Managing the aging of BWR control rod drive systems

    International Nuclear Information System (INIS)

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

    1992-01-01

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

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

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

    International Nuclear Information System (INIS)

    Kinoshita, Mitsuo; Yamada, Naoyuki; Kiguchi, Takashi

    1988-01-01

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

  15. Countercurrent flow-limiting characteristics of a Savannah River Plant control rod septifoil

    International Nuclear Information System (INIS)

    Anderson, J.L.

    1992-07-01

    Experiments were performed at the Idaho National Engineering Laboratory to investigate the counter-current flow limiting characteristics of a Savannah River Plant control rod septifoil assembly. These experiments were unheated, using air and water as the working fluids. Results are presented in terms of the Wallis flooding correlation for several different control rod configurations. Flooding was observed to occur in the vicinity of the inlet slots/holes of the septifoil, rather than within the rod bundle at the location of the minimum flow area. Nearly identical flooding characteristics of the septifoil were observed for configurations with zero, three, and four rods inserted, but significantly different results occurred with 5 rods inserted

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

    International Nuclear Information System (INIS)

    Jacquelin, Roland; Michot, Gilbert.

    1982-01-01

    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 [fr

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-02-15

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

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

    International Nuclear Information System (INIS)

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

    1998-01-01

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

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

    International Nuclear Information System (INIS)

    Gallozzi-Ulmann, A.; Couturier, P.; Amgarou, K.; Rothan, D.; Menaa, N.; Chard, P.

    2015-01-01

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

  20. Scram characteristics of control rods of pressurized water reactor under seismic conditions

    International Nuclear Information System (INIS)

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

    1987-01-01

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

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

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  2. Apparatus for coupling and rotatably securing control rod systems of nuclear reactors

    International Nuclear Information System (INIS)

    Jabsen, F.S.

    1977-01-01

    An apparatus is described for coupling and rotatably securing the control rod system of a nuclear reactor having a control rod assembly, a fuel assembly and a drive to translate longitudinally the control rod assembly within the fuel assembly. This apparatus is of importance during reactor maintenance when fuel assemblies are removed from the core for refuelling and is particularly applicable in the case of the consolidated nuclear steam generator type of reactor. (U.K.)

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

    Energy Technology Data Exchange (ETDEWEB)

    Norinder, O.; Nyman, K.

    1962-06-15

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

  4. High temperature study of the control rod behaviour under accident conditions

    International Nuclear Information System (INIS)

    Troyanov, V.; Pomeschikov, A.; Sougonjaev, V.; Ponomarenko, V.; Scheglov, A.

    2000-01-01

    Control rod behaviour under accident conditions is considered. A comparative assessment of various types of absorbers is carried out in this respect. It was found that the best absorbers are dysprosium based absorbers. Dysprosium based control rods show no failure at temperatures below 1300 deg. C. Boron based control rods may be liquefied at a temperature of 1150 deg. C. Besides, the cladding may burst at lower temperature, due to the released helium. Experiments indicate that the absorber materials may initiate severe core damage under severe accidents. The beginning of core structure degradation is caused by the boron control rod liquefaction. (author)

  5. Development of control rod position indicator using reed switch for SMART CEDM

    International Nuclear Information System (INIS)

    Yoo, Jae Yong; Kim, Jong In; Kim, Ji Ho; Huh, Hyoung; Jang, Moon Hee

    2001-01-01

    The reliability and accuracy of the information on control rod position are very important to the reactor safety and the design of the core protection system. In this study, a thorough investigation on the RSPT(Reed Switch Position Transmitter) type control rod position indication system and its actual implementation in the exiting nuclear power plants in Korea was performed first. A design of the control rod position indication system using reed switch for the CEDM on the integral reactor SMART was developed based on the position indicator technology identified through the investigation. The feasibility of the design was evaluated by test of manufactured control rod position indicator using reed switch for SMART

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

    International Nuclear Information System (INIS)

    Padilla M, C.E.

    1998-01-01

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

  7. Method of cleaning pipeline in control rod drive

    International Nuclear Information System (INIS)

    Baba, Mikiya.

    1993-01-01

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

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

    International Nuclear Information System (INIS)

    Shen Xiaoyao

    2012-01-01

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

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

    International Nuclear Information System (INIS)

    Gunther, W.; Sullivan, K.

    1990-01-01

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

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Persson, Rolf

    1966-10-15

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

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

    International Nuclear Information System (INIS)

    Persson, Rolf

    1966-10-01

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

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-05-01

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

  15. Monte Carlo simulation of neutron transport in a homogeneous reactor with a partially inserted control rod

    International Nuclear Information System (INIS)

    Karlsson, J.K.H.; Linden, P.

    1997-01-01

    The neutron transport in a bare, cylindrical and homogeneous reactor, with and without the presence of a central partially inserted control rod, has been simulated by using a Monte Carlo transport code. The behaviour of both the flux and current in this system have been investigated. We have found that the flux and especially the current are strongly affected by the presence of the control rod in its close vicinity. The results indicate the feasibility to identify the position and especially the tip of the rod from the flux and current. Further, the direction to the rod can be found from the current vector. The information content regarding the position of the rod, in both the neutron flux and the current, decays strongly as a function of distance and it is dependent on the size of the rod. In our model, the practical range over which the flux or current can be a useful indicator of the position of the tip of the rod is about 10-15 cm for a rod with a diameter of 2 cm. The practical range for identification of the position of the rod is greater for a rod of larger diameter

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

    International Nuclear Information System (INIS)

    Lecourt, G.

    1998-01-01

    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.)

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

    International Nuclear Information System (INIS)

    Yoshida, Kenji.

    1984-01-01

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

  18. Transient rod failure in a pulsing TRIGA Mark I reactor

    International Nuclear Information System (INIS)

    Draper, E.L. Jr.; Atkinson, G.D. Jr.

    1972-01-01

    Full text: On July 7, 1970 the University of Texas at Austin TRIGA Mark I Pulsing Reactor experienced a failure of the transient control rod. Although no danger to personnel or damage to the reactor other than the pulse rod occurred, the failure was promptly reported to the USAEC regional compliance office. The first indication of an abnormal situation was unusual multiplication behavior during the first start-up of the day. As usual for steady state operation, the operator removed the transient rod and began to withdraw the shim and regulating rods. After partial withdrawal, he noticed that the count rate was not increasing as rapidly as was customary. While remaining at the console,the operator had a technician make a visual inspection of the core. The technician observed the transient drive rod was swinging freely in the pool and the poison section was detached. It was concluded, based on the indications of the.reactor instrumentation and visual inspection, that the transient control rod had broken off and remained in position in the core. The regulating and shim rods were inserted and the transient rod was manually cranked to the down position. The manual manipulation of the transient rod, instead of dropping the rod by gravity, was used so that the connecting rod could be reinserted in the control rod guide tube. The reactor core was then partially unloaded so that a critical mass was not present. The transient rod drive and connecting rod were removed from the pool. The poison section was retrieved from its position in the core by welding a tap to a long rod and tapping into the top of the poison section. Visual inspection of the poison section showed that the weld joining the male threads on the poison section to the main body of the control rod had failed. The threads remained screwed in the control rod drive shaft upon separation and the poison section remained fully inserted in the core. A new control rod was fabricated by Gulf General Atomic and shipped

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

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

    International Nuclear Information System (INIS)

    Savary, F.

    1986-01-01

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

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

    International Nuclear Information System (INIS)

    Li Peng; Yan Yiren; Lu Li; Liu Tong; Zhou Yuemin; Guo Yan

    2013-01-01

    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)

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

    International Nuclear Information System (INIS)

    Chang Zhengke; Shao Dinghong

    2004-11-01

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

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

    International Nuclear Information System (INIS)

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

    1989-01-01

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

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

    International Nuclear Information System (INIS)

    Gomez, A.; Waldman, R.M.

    1993-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-07-01

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

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

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

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  9. Aging assessment of BWR control rod drive systems

    International Nuclear Information System (INIS)

    Greene, R.H.

    1991-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Takeda, Takeshi E-mail: takeda@lstf3.tokai.jaeri.go.jp; Tachibana, Yukio

    2003-07-01

    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.

  11. Optical and mechanical attachment to coaxility control rod channels when mounting WWER-1000 reactor

    International Nuclear Information System (INIS)

    Vajnberg, B.I.; Rodionov, N.F.

    1987-01-01

    Optical and mechanical attachment to control coaxiality of control rod channels when mounting WWER-1000 reactor is described. It is indicated, that this attachment was applied at the Kalinin NPP second power unit

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

    International Nuclear Information System (INIS)

    Christensen, L.J.

    1983-01-01

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

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

    Science.gov (United States)

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

    2016-01-01

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

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

    International Nuclear Information System (INIS)

    Pena V, J.D.

    2016-01-01

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

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

    International Nuclear Information System (INIS)

    Nonboel, E.

    1989-08-01

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

  16. Review of design technology of control rod position indicators

    International Nuclear Information System (INIS)

    Yu, Je Yong; Huh, Hyung; Kim, Ji Ho; Kim, Jong In; Chang, Moon Hee

    1999-10-01

    An integral reactor SMART is under development at KAERI. The design characteristics of SMART are radically different from those employer in currently operating loop type water reactors in Korea. The objective of this report is to review the design technology of position indicator, and to study the various sensors which can be used in rod position indicator. Design criteria that rod position indicator should satisfy are also examined. Following position indicators are reviewed in this report. 1. Digital positioning indicator (DRPI), 2. Reed switch type position indicator (RSPT), 3. Choke sensor type position indicator, 4. Ultrasonic sensor type position indicator, 5. Comparison of each position indicator. (author)

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

    International Nuclear Information System (INIS)

    Markiewicz, Mario; Estevez, Esteban

    2000-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-06-01

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

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

    International Nuclear Information System (INIS)

    Song Qing; Yang Hongyi; Jing Yueqing; Wen Jing; Liu Guijuan; Sun Lei

    2008-01-01

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

  20. Control system for nuclear reactor with a series of flexible control rods, diverging towards the core from a common channel

    International Nuclear Information System (INIS)

    Doll, D.W.

    1976-01-01

    The description is given of a nuclear reactor control system comprising a pressure vessel having at least one leaktight crossing and a core area enclosed in the pressure vessel and in which a series of control rod guiding channels are distributed over an area exceeding the leaktight crossing section. This control includes a series of flexible control rods, intended to enter the guiding channels at the rate of one per channel, a set of guide tubes which diverge from the leaktight crossing to respective guide channels to guide the control rods, and operating gear to move the control rods in the guide tubes to cause them to take up and leave insertion positions in the guide channels. Each control rod is made up of several rigid sections joined by articulated joints. This invention is intended for high temperature reactors [fr

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

    International Nuclear Information System (INIS)

    1975-03-01

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

  2. Control rod upper pin and roller removal and replacement

    International Nuclear Information System (INIS)

    Giordano, B.

    1995-01-01

    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

  3. Failure Mode and Effects Analysis (FMEA) of the solid state full length rod control system

    International Nuclear Information System (INIS)

    Shopsky, W.E.

    1977-01-01

    The Full Length Rod Control System (FLRCS) controls the power to the rod drive mechanisms for rod movement in response to signals received from the Reactor Control System or from signals generated through Reactor Operator action. Rod movement is used to control reactivity of the reactor during plant operation. The Full Length Rod Control System is designed to perform its reactivity control function in conjunction with the Reactor Control and Protection System, to maintain the reactor core within design safety limits. By the use of a Failure Mode and Effects Analysis, it is shown that the FLRCS will perform its reactivity control functions considering the loss of single active components. That is, sufficient fault limiting control circuits are provided which blocks control rod movement and/or indicates presence of a fault condition at the Control Board. Reactor operator action or automatic reactor trip will thus mitigate the consequences of potential failure of the FLRCS. The analysis also qualitatively demonstrates the reliability of the FLRCS to perform its intended function

  4. Experience on reliability of insertion of control rod with PWRs in Japan

    International Nuclear Information System (INIS)

    Nakata, H.

    1995-01-01

    This paper is prepared to introduce, in response to the IAEA Letter dated 1995-02-08, national experience concerning to the reliability of insertion of control rod with PWRs which is relatively similar to WWER 1000. 3 figs

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  6. Sample Federal Facility Land Use Control ROD Checklist and Suggested Language (LUC Checklist)

    Science.gov (United States)

    The LUC Checklist provides direction on describing and documenting land use controls (LUCs) in federal facility actrions under CERCLA in Records of Decision (RODs), remedial designs (RDs), and remedial action work plans (RAWPs).

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

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

    OpenAIRE

    E.K. Boafo; E. Alhassan; E.H.K. Akaho; C. Odoi

    2013-01-01

    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 an...

  9. Measurements and calculations of 10B(n,He) reaction rates in a control rod in ZPPR

    International Nuclear Information System (INIS)

    Brumbach, S.B.; Collins, P.J.; Grasseschi, G.L.; Oliver, B.M.

    1986-01-01

    The helium accumulation fluence monitor (HAFM) technique has been used to measure the 10 B(n,He) reaction rate within B 4 C pellets in a control rod in ZPPR. Knowledge of this reaction rate is important to control rod design studies because helium production leads to control rod swelling, buildup of gas pressure and a reduction in thermal conductivity which can limit the lifetime of a control rod. We believe these to be the first measurements of boron capture within boron pins in a fast reactor spectrum. Previously reported measurements used 235 U foils to measure fission rates in a control rod, and to infer boron capture rates

  10. Wear plates control rod guide tubes top internal reactor vessel C. N. VANDELLOS II

    International Nuclear Information System (INIS)

    2010-01-01

    The guide tubes for control rods forming part of the upper internals of the reactor vessel, its function is to guide the control rod to permit its insertion in the reactor core. These guide tubes are suspended from the upper support plate which are fixed by bolts and extending to the upper core plate which is fastened by clamping bolts (split pin) to prevent lateral displacement of the guide tubes, while allowing axial expansion.

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

    International Nuclear Information System (INIS)

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

    2015-01-01

    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. - 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

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

    International Nuclear Information System (INIS)

    Giese, H.

    1992-10-01

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

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

    DEFF Research Database (Denmark)

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

    2008-01-01

    Effective control of vortex breakdown in a cylinder with a rotating lid was achieved with small rotating rods positioned on the stationary lid. After validation with accurate measurements using a novel stereoscopic particle image velocimetry (SPIV) technique, analysis of numerical simulations using...... a high-order spectral element method has been undertaken. The effect of a finite length rod creates additional source terms of vorticity as the rod rotates. These additional source terms and their spatial locations influence the occurrence of the vortex breakdown....

  14. 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)

    1964-01-01

    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

  15. Studies on neutron noise diagnostics of control rod vibrations by neural networks

    International Nuclear Information System (INIS)

    Roston, G.; Kozma, R.; Kitamura, M.; Garis, N.S.; Pazsit, I.

    1996-01-01

    This work is focussed on the study of a neutron noise based technique for the diagnostics of reactor core internal, in particular, excessively vibrating control rods. The use of a combination of physical models and neural networks offers an alternative way of performing the inversion procedure. The application of a neural network technique to determine the rod position from the detector spectra is much faster, more effective and simpler to use than the conventional method. (author). 5 refs., 1 fig., 1 tab

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

    Energy Technology Data Exchange (ETDEWEB)

    Lind, Terttaliisa, E-mail: terttaliisa.lind@psi.c [Paul Scherrer Institut, Villigen (Switzerland); Csordas, Anna Pinter; Nagy, Imre [HAS KFKI Atomic Energy Research Institute, Budapest (Hungary); Stuckert, Juri [Forschungszentrum Karlsruhe, Karlsruhe (Germany)

    2010-02-15

    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 approx 1570 K at hottest bundle elevation. A very large burst of aerosols was detected 660 s later at the bundle temperature maximum of T approx 1650 K, followed by a

  17. Estimation of the activity generated by neutron activation in control rods of a BWR

    International Nuclear Information System (INIS)

    Rodenas, Jose; Gallardo, Sergio; Juan, Violeta

    2008-01-01

    Full text: Control rods in a BWR are cross shaped; every blade of the cross having 18 stainless steel tubes containing CB 4 as neutron absorber. These tubes are closed by welding taps at both sides, and they are contained in a stainless steel gain with holes that allow proper cooling. The estimated life of one control rod is about 15 years. During its stay into the reactor the components of the rod are activated by neutron reactions. The activation is produced mainly in impurities contained in stainless steel as well as in the elements composing the steel alloy. The activity so generated will produce a dose around the rod, not important while it is inside the reactor, but it has to be taken into account when the rod is withdrawn from the reactor. In this work, activation reactions produced in stainless steel have been modelled by the MCNP code based on the Monte Carlo method. The code gives the number of reactions that can be converted into activity by the expression: A(t) = Q(1 - exp(-λt i )) exp(-λt e ) where Q is the number of reactions produced per cm 3 and second; t i is the irradiation time; and t e the time since the rod is discharged. Two different materials have been considered in the model: stainless steel (with impurities) and an absorbent mix composed by CB 4 , stainless steel (tube gains) and H 2 O. The control rod handle has been modelled as pure stainless steel. Results show that the greatest activation is produced in stainless steel and mainly in the rod handle as expected being the piece more exposed to neutron flux. For a thermal flux the most important reaction is radioactive capture and radionuclides with greater activities are by this order 59 Ni, 55 Fe, 60 Co, and 63 Ni. Figures are reproduced showing a schematic design of model as well as some details of the cells

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

    International Nuclear Information System (INIS)

    Figueredo, A.M. de.

    1985-12-01

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

  19. Performance estimation of control rod position indicator due to aging of magnet

    International Nuclear Information System (INIS)

    Yu, Je Yong; Kim, Ji Ho; Huh, Hyung; Choi, Myoung Hwan; Sohn, Dong Seong

    2009-01-01

    The Control Element Drive Mechanism (CEDM) for the integral reactor is designed to raise and lower the control rod in steps of 2mm in order to satisfy the design features of the integral reactor which are the soluble boron free operation and the use of a nuclear heating for the reactor start-up. The actual position of the control rod could be achieved to sense the magnet connected to the control rod by the position indicator around the upper pressure housing of CEDM. It is sufficient that the actual position information of control rod at 20mm interval from the position indicator is used for the core safety analysis. As the magnet moves upward along the position indicator assembly from the bottom to the top in the upper pressure housing, the output voltage increases linearly step-wise at 0.2VDC increments. Between every step there are transient areas which occur by a contact closing of three reed switches which is the 2-3-2 contact closing sequence. In this paper the output voltage signal corresponding to the position of control rod was estimated on the 2-1-2 contact closing sequence due to the aging of the magnet.

  20. Method and means for remote removal of guide balls from nuclear reactor control rods

    International Nuclear Information System (INIS)

    Krieg, A.H.

    1988-01-01

    This patent describes a method of remotely removing guide balls from nuclear reactor control rods using a punch mechanism, comprising: (a) providing attachment means in the punch mechanism for attaching the punch mechanism to means for reversibly lowering the punch mechanism over the top of one of the control rods; (b) providing a die within the punch mechanism; (c) providing cylinder means within the punch mechanism operatively connected to the die for axially moving the die in a back-and-forth direction; (d) providing a die block within the punch mechanism cooperating with the die; (e) providing guide means within the punch mechanism for self-aligning the punch mechanism so that the die and the die block are automatically aligned with a first one of the guide balls therebetween when the punch mechanism is lowered over the top of the control rod; (f) lowering the punch mechanism over the control rod so that the die, the die block, and the first guide ball are in alignment; and (g) then operating the cylinder means so that the die advances into the die block, thereby removing the first guide ball from the control rod

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

    International Nuclear Information System (INIS)

    Silva, Jose Eduardo R. da; Terremoto, Luis A.A.; Castanheira, Myrthes; Zeituni, Carlos A.; Damy, Margaret de A.

    2002-01-01

    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)

  2. Material Tracking with Dynamic Torque Adaptation for Tension Control in Wire Rod Mill

    Directory of Open Access Journals (Sweden)

    Tomas Borovsky

    2017-01-01

    Full Text Available Material tracking is an important part of the automation control system which has a major impact on the product quality. This paper addresses a stand load identification in wire rod mill as a new algorithm added to existing control system. Tension control approaches are described and a modification of existing tracking system is proposed in order to eliminate tracking faults. Proposed method is based on dynamic torque calculation and its performance was experimentally verified on the industrial wire rod mill. Experimental results show significant reduction of the errors.

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

    International Nuclear Information System (INIS)

    Cancino P, G.

    2016-01-01

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

  4. 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)

    Tavolara, Ivan; Desgeorge, Romain; Verburgh, Pierre

    2014-01-01

    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

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

    International Nuclear Information System (INIS)

    Okamura, Hajime.

    1996-01-01

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

  6. Design of control and safety rod and its drive mechanism of PFBR

    International Nuclear Information System (INIS)

    Rajan Babu, V.; Govindarajan, S.; Chetal, S.C.

    1997-01-01

    Control and Safety Rod (CSR) is one of the two types of absorber rods in shutdown systems of PFBR. Control and Safety Rod Drive Mechanism (CSRDM) actuates CSR to have vertical translatory motion in reactor core. The dual responsibilities entrusted on CSR to control reactor power during normal operating condition and to shutdown the reactor by scram action during abnormal condition, necessitate highly reliable design, analysis, testing and surveillance of CSR and CSRDM. The paper discusses on the salient features of CSR and CSRDM and design and analysis of individual sub-assemblies, viz., gripper, scram-release electromagnet, hydraulic dash pot, seals. Also it discusses on the developmental activities proposed and surveillance test requirements. (author)

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

    International Nuclear Information System (INIS)

    Bowsher, B.R.; Jenkins, R.A.; Nichols, A.L.; Rowe, N.A.; Simpson, J.A.H.

    1986-04-01

    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)

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

    International Nuclear Information System (INIS)

    Krohn, John L.; Fisher, Thomas H.

    1988-01-01

    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

  9. Rod examination gauge

    Energy Technology Data Exchange (ETDEWEB)

    Bacvinskas, W.S.; Bayer, J.E.; Davis, W.W.; Fodor, G.; Kikta, T.J.; Matchett, R.L.; Nilsen, R.J.; Wilczynski, R.

    1991-12-31

    The present invention is directed to a semi-automatic rod examination gauge for performing a large number of exacting measurements on radioactive fuel rods. The rod examination gauge performs various measurements underwater with remote controlled machinery of high reliability. The rod examination gauge includes instruments and a closed circuit television camera for measuring fuel rod length, free hanging bow measurement, diameter measurement, oxide thickness measurement, cladding defect examination, rod ovality measurement, wear mark depth and volume measurement, as well as visual examination. A control system is provided including a programmable logic controller and a computer for providing a programmed sequence of operations for the rod examination and collection of data.

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

    International Nuclear Information System (INIS)

    Lee, E. K.; Shin, H. C.; Bae, S. M.; Lee, Y. G.

    2004-01-01

    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)

  11. Eddy-current inspection of high flux isotope reactor nuclear control rods

    International Nuclear Information System (INIS)

    Smith, J.H.; Chitwood, L.D.

    1981-07-01

    Inner control rods for the High Flux Isotope Reactor were nondestructively inspected for defects by eddy-current techniques. During these examinations aluminum cladding thickness and oxide thickness on the cladding were also measured. Special application techniques were required because of the high-radiation levels (approx. 10 5 R/h at 30 cm) present and the relatively large temperature gradients that occurred on the surface of the control rods. The techniques used to perform the eddy-current inspections and the methods used to reduce the associated data are described

  12. Experiment monitoring system of a new electromagnet drive for nuclear reactor control rod

    International Nuclear Information System (INIS)

    Zhang Jige; Wang Xiaoguang; Wu Yuanqiang; Zhang Zhengming

    2003-01-01

    In order to deal with some unsolved problems in the engineering prototype design of a new electromagnet drive device for nuclear reactor control rod, the property experiment in view of principle prototype is carried out. Actual displacement of nuclear reactor control rod is measured by means of raster ruler and the test data is obtained by means of computer. The computer communicates with PLC using RS232 serial port. The experimental results show that the monitoring system have the properties of high reliability and high precision, and ensures the experiment to accomplish successfully

  13. Determination of curve 1/M profile as a function of control rod bank position

    International Nuclear Information System (INIS)

    Pereira, Valmir; Martinez, Aquilino Senra; Silva, Fernando Carvalho da

    2002-01-01

    Determination of the subcritical multiplication curve profile (1/M) as a function of control rod bank position is of paramount importance to the development of a system which allows to foresee and also anticipate determination of criticality of a PWR reactor core. This work aims at determining this profile. For that, the 3D- two group-diffusion equations for a subcritical PWR reactor core with external neutron source is solved for different control rod bank positions. Results obtained are compared with the results from the corresponding eigenvalue problem, in order to verify how the external neutron source interferes with the reactor criticality search. (author)

  14. Heat flux and temperature determination on the control rod outer surface

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  15. Verification of heat flux and temperature calculation on the control rod outer surface

    Science.gov (United States)

    Taler, Jan; Cebula, Artur

    2011-12-01

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

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

    International Nuclear Information System (INIS)

    Nguyen Hoang Hai; Do Quang Binh

    2011-01-01

    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)

  17. Development of the predictive maintenance system prototype for the rod control system

    International Nuclear Information System (INIS)

    Lim, H. S.; Hong, H. P.; Koo, J. M.; Kim, Y. B.; Han, H. W.

    2003-01-01

    The demand for safety and reliability of Nuclear Power Plants (NPPs) has been constantly increasing and economical operation is also an important issue. Developing and adopting predictive maintenance technology for the major systems or equipment is considered as a way to achieve these goals. This paper describes the development of a predictive maintenance system prototype for the Rod Control System, which adopts an advanced methodology. Bayesian Belief Networks (BBN) has been adopted for the real time fault diagnosis and prediction of the system. Through a simulation test, it was confirmed that the prototype monitors and secures sound operability of rod drive mechanism and its control system, and also provides the predictive maintenance information

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

    International Nuclear Information System (INIS)

    Hosoya, Kazuya; Kanazawa, Takashi; Matsuzaki, Tokunori.

    1986-01-01

    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.)

  19. Control rod drop accident analysis for the mixed core project in Ling Ao NPS

    International Nuclear Information System (INIS)

    Zhang Shishun; Zhou Zhou; Xiao Min

    2004-01-01

    AFA-2G assemblies in Ling Ao NPS (LNPS) have been replaced gradually by AFA-3G assemblies from cycle 2 and subsequent cycles. the enrichment of the fuels will be increased from 3.2% to 3.7% from cycle 3 in Ling Ao. Therefore, the study of ling Ao mixed core and increased enrichment have been performed since 2001. Lots of accidents need to be re-analyzed in Ling Ao NPS in order to verify its safety requirements for the new fuel management. Control rod drop accident for LNPS was re-analyzed in 2001 in frame of FRAMATOME ANP analytical methodology. The analytical codes used in the accident analysis include SCIENCE, ESPADON, CINEMA, CANTAL and FLICA III. The control rod drop accident analysis is performed with respect to the 10 reference cycles of the generic fuel management design for Ling Ao mixed core and increased enrichment study. The pre-drop FδH for the first transition cycles and other cycles are 1.52 and 1.55, respectively. For detected dropped rod configurations, the negative flux rate protection system actuates a reactor trip. For the non-detected dropped rod configurations, the minimum DNBR values have been evaluated with conservative analysis methodology and assumptions and the DNBR fuel design limit is respected the analytical results shows that, for all the non-detected dropped rod configurations, the minimum DNB margin is about 2% which occurs in AFA-2G fuel assembly in the first transition cycle. (author)

  20. System for fuel rod removal from a reactor module

    International Nuclear Information System (INIS)

    Matchett, R.L.; Roof, O.R.; Kikta, T.J.; Wilczynski, R.; Nilsen, R.J.; Bacvinskas, W.S.; Fodor, G.

    1990-01-01

    This patent describes a robotic system for remote underwater withdrawal of the fuel rods from fuel modules of a light water breeder reactor includes a collet/grapple assembly for gripping and removing fuel rods in each module, which is positioned by use of a winch and a radial support means attached to a vertical support tube which is mounted over the fuel module. A programmable logic controller in conjunction with a microcomputer, provides control for the accurate positioning and pulling force of the rod grapple assembly. Closed circuit television cameras are provided which aid in operator interface with the robotic system

  1. Evaluation of the Control Rod Super Alloy Material of HTR-PM

    International Nuclear Information System (INIS)

    Li Pengjun; Yan He; Diao Xingzhong

    2014-01-01

    The control rod drive mechanism (CRDM) system is served as the first reactivity control and shutdown system for the high temperature reactor pebble-bed module (HTR-PM) in Shandong, China. And the control rod, which is pulled up and down by a chain sprocket mechanism of CRDM to realize reactivity control, compensation and shutdown, has to be durable under temperature as high as 550℃ for a long time. Thus the material persistent strength under high temperature is quite important for the reliability of the CRDM. In this paper, a review on material selection of control rod of high temperature gas cooled reactors, including AVR and THTR-300 in Germany, HTTR in Japan, PBMR in South Africa and Dragon in Britain, was summarized. The major parameters of two kinds of high temperature alloy, incoloy 800H and alloy 625, were compared and discussed. According to the ASME NH volume, a design criterion for the control rod was established and applied in the analysis of the chain by using finite element method. The numerical simulations showed that the chain made of alloy 625 could meet the condition and work for a long time under high temperature. (author)

  2. Evaluation of Tehran research reactor (TRR) control rod worth using MCNP4C computer code

    International Nuclear Information System (INIS)

    Hosseini, Mohammad; Vosoughi, Naser; Hosseini, Seyed Abolfazl

    2010-01-01

    The main objective of reactor control system is to provide a safe reactor starting up, operation and shutting down. Calculation or measurement of precise values of control rod worth is of great importance in Tehran Research Reactor (TRR), considering the fact that they are the only controlling tools in the reactor. In present paper, simulation of TRR in First Operation Cycle (FOC) and in cold and clean core for the calculation of total and integral worth of control nods is reported. MCNP4C computer code has been used for all simulation process. Two method have been used for control rods worth calculation in this paper, namely the direct approach and perturbation method. It is shown that while the direct approach is appropriate for worth calculation of both the shim and the regulating control rods, the perturbation method is just suitable for tiny reactivity changes, i.e. for small initial part of regulating rods. Results of simulation are compared with the reported data in Safety Analysis Report (SAR) of Tehran research reactor and showed satisfactory agreement. (author)

  3. 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.

  4. AgInCd control rod failure in the QUENCH-13 bundle test

    Energy Technology Data Exchange (ETDEWEB)

    Sepold, L. [Forschungszentrum Karlsruhe, Institut fuer Materialforschung, Nuclear Safety Program (NUKLEAR), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany)], E-mail: leo.sepold@imf.fzk.de; Lind, T. [Paul Scherrer Institut, Laboratory for Thermalhydralics (LTH), Department of Nuclear Energy and Safety (NES), 5232 Villigen PSI (Switzerland); Csordas, A. Pinter [Fuel Materials Department, HAS KFKI AEKI, 1121 Budapest (Hungary); Stegmaier, U.; Steinbrueck, M.; Stuckert, J. [Forschungszentrum Karlsruhe, Institut fuer Materialforschung, Nuclear Safety Program (NUKLEAR), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany)

    2009-09-15

    The QUENCH off-pile experiments performed at the Karlsruhe Research Center are to investigate the high-temperature behavior of Light Water Reactor (LWR) core materials under transient conditions and in particular the hydrogen source term resulting from the water injection into an uncovered LWR core. The typical LWR-type QUENCH test bundle, which is electrically heated, consists of 21 fuel rod simulators with a total length of approximately 2.5 m. The Zircaloy-4 rod claddings and the grid spacers are identical to those used in Pressurized Water Reactors (PWR) whereas the fuel is represented by ZrO{sub 2} pellets. In the QUENCH-13 experiment the single unheated fuel rod simulator in the center of the test bundle was replaced by a PWR-type control rod. The QUENCH-13 experiment consisting of pre-oxidation, transient, and quench water injection at the bottom of the test section investigated the effect of an AgInCd/stainless steel/Zircaloy-4 control rod assembly on early-phase bundle degradation and on reflood behavior. Furthermore, in the frame of the EU 6th Framework Network of Excellence SARNET, release and transport of aerosols of a failed absorber rod were to be studied in QUENCH-13, which was accomplished with help of aerosol measurements performed by PSI-Switzerland and AEKI-Hungary. Control rod failure was initiated by eutectic interaction of steel cladding and Zircaloy-4 guide tube and was indicated at about 1415 K by axial peak absorber and bundle temperature responses and additionally by the on-line aerosol monitoring system. Significant releases of aerosols and melt relocation from the control rod were observed at an axial peak bundle temperature of 1650 K. At a maximum bundle temperature of 1820 K reflood from the bottom was initiated with cold water at a flooding rate of 52 g/s. There was no noticeable temperature escalation during quenching. This corresponds to the small amount of about 1 g in hydrogen production during the quench phase (compared to 42 g

  5. High patient-parent satisfaction with magnetically controlled growing rod treatment in EOS

    DEFF Research Database (Denmark)

    Skov, Simon Toftgaard; Bünger, Cody

    Background: Magnetic controlled growing rod (MCGR) use is increasing worldwide and was by some considered a revolution in EOS management. Now, reports on MCGR-failures and complications similar to traditional growing rods are being reported but the treatment is still considered to be a major.......8 (range 9-10) in favor of MCGR, considering this method most gentle to their child (0 conventional most gentle, 10 MCGR most gentle). Conclusion: We found a uniform high overall satisfaction, and the parents with experience from other growth instrumentation would choose MCGR again if it was an option...

  6. 77 FR 8900 - Certain Vaginal Ring Birth Control Devices; Termination of the Investigation Based on Withdrawal...

    Science.gov (United States)

    2012-02-15

    ... From the Federal Register Online via the Government Publishing Office INTERNATIONAL TRADE COMMISSION Certain Vaginal Ring Birth Control Devices; Termination of the Investigation Based on Withdrawal... within the United States after importation of certain vaginal birth control devices by reason of...

  7. Calculation of the Reactivity Equivalence of Control Rods in the Second Charge of the HBWR

    International Nuclear Information System (INIS)

    Weissglas, P.

    1960-11-01

    Full text: Using current methods the reactivity equivalence of 19 31 and 37 centrally located control rods in the second charge of the HBWR has been calculated. An estimate of the available excess reactivity with clean cold core has also been made. Insertion depth was taken as 0, l/3, 2/3 and 3/3 of the core length

  8. 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. Copyright © 2015 Elsevier Ltd. All rights reserved.

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

    International Nuclear Information System (INIS)

    Lemaire, E.; Couet, D.; Molinie, D.; Grandjean, Y.; Radat, M.P.; Guttmann, D.

    1998-01-01

    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)

  10. Sleep-disordered breathing decreases after opioid withdrawal: results of a prospective controlled trial.

    Science.gov (United States)

    Schwarzer, Andreas; Aichinger-Hinterhofer, Marie; Maier, Christoph; Vollert, Jan; Walther, Jörg Werner

    2015-11-01

    An increased cardiovascular event rate in elderly patients under opioid medications was recently reported. One reason for this increase could be the occurrence of nocturnal apnea and hypoxia, as a consequence of sleep-disordered breathing (SDB). Using a controlled study, we prospectively analyzed SDB using polysomnography in a total of 18 patients before and after opioid withdrawal (opioid withdrawal group [OG]) and 14 patients before and after comprehensive pain management (without any strong-acting opioids) who served as the control group (CG). To analyze the differences, unpaired/paired t tests and Mann-Whitney U tests/Wilcoxon rank tests were used. At baseline, the OG presented more nocturnal apneas/hypopneas than the CG with an apnea-hypopnea index (AHI) of 41.4 ± 27.8 vs 21.8 ± 15.9 (P = 0.018). After treatment, the AHI decreased significantly only in the withdrawal group (OG: 16.7 ± 8.9; CG: 20.1 ± 12.9) (P opioid withdrawal and in none of the patients after withdrawal (P opioid intake; these findings may explain the opioid-associated cardiovascular morbidity. Thus, SDB may be a risk at lower opioid doses than hitherto described, and particular caution should be exercised in patients with comorbidities that might make them vulnerable to the consequences of SDB.

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

    International Nuclear Information System (INIS)

    Souza, Rose Mary Gomes do Prado

    2005-01-01

    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)

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

    International Nuclear Information System (INIS)

    Giese, H.

    1991-03-01

    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 B 2 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. Influence of structure improvement of guide tubes and bundles in pressurized water reactor (PWR) on drop of control rods

    International Nuclear Information System (INIS)

    Shen Xiuzhong; Yu Pingan; Yang Guanyue

    1996-01-01

    In order to alleviate the cross hydraulic load on control rod guide tubes and bundles, some protective sleeves are added to those near the upper plenum outlet nozzles (4 symmetric bundles: 02-26, 03-25, 11-29, 12-28). In a 1/4 scale transparent model of the PWR upper plenum of Qinshan Nuclear Power Station, water was chosen as the fluid and hydraulic experiments with improved control rod guide tubes and bundles were carried out. The results were carefully compared with those of the experiments with unimproved control rod guide tubes and bundles. It is concluded that adding protective sleeves to the control rod guide tubes and bundles near the outlet nozzles will help to lighten the hydraulic load on them and make certain of the free movement and rapid dropping of control rods in the tubes and bundles in emergency by order

  14. Comparison of tincture of opium and methadone to control opioid withdrawal in a Thai treatment centre

    Science.gov (United States)

    Jittiwutikarn, Jaroon; Ali, Robert; White, Jason M; Bochner, Felix; Somogyi, Andrew A; Foster, David J R

    2004-01-01

    Aims To evaluate the effectiveness of oral tincture of opium (TOP) and methadone to control opioid withdrawal in patients in northern Thailand. Methods Open label, parallel group study in an inpatient facility compared 15 former heroin users receiving methadone 5–20 mg 12 hourly with 15 former opium smokers receiving TOP (3.33–10 mg morphine equivalents 12 hourly). At 0, 1, 3 and 8 h, blood, withdrawal scores and subjective opioid effects were collected. Results There was a reciprocal association between withdrawal scores/direct subjective opioid effects and plasma (R)-methadone, but not plasma morphine, concentrations. Withdrawal scores at the time of dosing were higher in the TOP patients (9.1 ± 3) than in the methadone patients (4.5 ± 4.6) and in the TOP patients were significantly (P = 0.001) attenuated at 3 and 8 h. Conclusions At the doses used, TOP was inferior to methadone in suppressing withdrawal. It could prove to be a cost effective and valuable drug, but only after dose size and frequency are further investigated. PMID:15521902

  15. Gabapentin's acute effect on mood profile -- a controlled study on patients with alcohol withdrawal.

    Science.gov (United States)

    Bonnet, Udo; Specka, Michael; Leweke, F Markus; Nyhuis, Peter; Banger, Markus

    2007-03-30

    Delayed beneficial effects of gabapentin on mood were frequently reported in various patient populations. This is the first controlled study which addressed acute effects of gabapentin on mood. Analysis of the German version of Profile of Mood States (POMS) throughout a randomised placebo-controlled, double-blinded study of gabapentin on acute alcohol withdrawal [Bonnet, U., Banger, M., Leweke, F.M., Specka, M., Müller, B.W., Hashemi, T., Nyhuis, P.W., Kutscher, S., Burtscheidt, W., Gastpar, M. 2003. Treatment of acute alcohol withdrawal with gabapentin -- results from a controlled two-center trial. J Clin Psychopharmacol 23, 514-519]. In addition, subjective severity of alcohol withdrawal was determined by the Essen Self-Assessment of Alcohol Withdrawal Scale (ESA) to control effects of concurrent withdrawal on POMS. Ratings were performed at intake (baseline), day 1 (study medication 400 mg q.i.d.), day 2 (study medication 400 mg q.i.d.) and day 7 (no study medication). Analyses could be performed on 46 out of 59 randomised subjects. Within the first two days of the study, a significant stronger increase in the POMS-vigour subscore occurred in the gabapentin group. A subgroup analysis suggests that gabapentin's effect on vigour largely results from a stronger improvement of vigour in a small group of 11 patients with co-morbid mild depression (according to ICD-10: dysthymia or depressive adjustment disorder). There were no significant differences between the treatment groups regarding the other POMS-subscores (dejection, fatigue, anger) ruling out an overall fast effect on mood. Moreover, ESA-measures were not significantly altered indicating a missing effect of 400 mg gabapentin q.i.d. on acute alcohol withdrawal itself. After tapering off study medication, no more significant differences between gabapentin and placebo group were observed on vigour, strongly suggesting that the initial effect results from a pharmacological gabapentin action. Gabapentin

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

    International Nuclear Information System (INIS)

    Jacquart, G.

    1993-12-01

    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

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

    International Nuclear Information System (INIS)

    Martinez, J.M.

    1979-01-01

    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 [fr

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

    International Nuclear Information System (INIS)

    Angele, Kristian; Odemark, Ylva; Cehlin, Mathias; Hemström, Bengt; Högström, Carl-Maikel; Henriksson, Mats; Tinoco, Hernan; Lindqvist, Hans

    2011-01-01

    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

  19. Numerical experiment designs. Study of vibratory behaviour of PWR'S control rod clusters

    International Nuclear Information System (INIS)

    Bosselut, D.; Soulier, B.; Regnier, G.

    1997-01-01

    The application of Experiment Design method to Finite Element Model (FEM) calculations is an original way of performing parametric studies. It has been used at EDF to simulate on a large parametric domain the vibrations of PWR's control rod cluster and to analyse the rod wear process. In the first part the FEM and the location of excitation sources are described. The calculated values are: rod displacement in the guiding cards, shock forces on the guiding cards and the wear power produced. In the second part, the computed Experiment Domain is described. This method approaches the response surface by a second degree polynomial. The retained model is composed for every parameters of all linear, quadratic and interaction terms (26 coefficients). In all, 34 polynomials have been built to approach the effective shock forces and the mean wear power at each of the 17 guiding points. In the third part the building of the computer Experiment Design is detailed: by Doehlert design adaptation to take into account a qualitative parameter, design optimization by adding four well chosen experiments and finally, design extension by passing from 4 to 6 parameters. In the last part, all the information deduced from application of this method are presented. The influence of parameters on calculated effective shock forces has been determined along rods and response surface have been easily approximated. The systematism and closeness of Experiment Design technique is underlined. Easy simulation of all the response domain by polynomial approach allows comparison with experimental results. (authors)

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

    International Nuclear Information System (INIS)

    Zwingelstein, G.; Deat, M.; Le Guillou, G.

    1979-01-01

    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

  1. Design and experimental research of hydraulic control rod drive system for 5 MW THR

    International Nuclear Information System (INIS)

    Wu Yuanqiang; Hu Yuedong; Cheng Yunsheng; Yang Nianzu; Liu Chengying; Zhang Fulu

    1990-01-01

    Hydraulic control rod drive system (HCRDS) is a new device of control rod drive which is different from usual electric-magnetic mechanic drive used for water power nuclear reactor. The coolant(water) is used as actuating medium, pumped by canned-pump, then injected into hydraulic step cylinders which are set in the reactor vessel. The outer tube of the step cylinder will be moving step by step by controling of the flow, driving the neutron absorber. 5MW THR is the first reactor in the world in which the hydraulic step drive system was used. Using of the drive system is for getting better safety, reliability and lower cost. In this paper, the design and experimental research of the system are introduced

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

    International Nuclear Information System (INIS)

    Mazufri, C.M; Mazzi, R.O

    2000-01-01

    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

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

    International Nuclear Information System (INIS)

    Huh, Hyung; Cho, Yeong Garp; Kim, Jong In

    2013-01-01

    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

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

    International Nuclear Information System (INIS)

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

    2016-01-01

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

  5. Aging considerations for PWR [pressurized water reactor] control rod drive mechanisms and reactor internals

    International Nuclear Information System (INIS)

    Ware, A.G.

    1988-01-01

    This paper describes age-related degradation mechanisms affecting life extension of pressurized water reactor control rod drive mechanisms and reactor internals. The major sources of age-related degradation for control rod drive mechanisms are thermal transients such as plant heatups and cooldowns, latchings and unlatchings, long-term aging effects on electrical insulation, and the high temperature corrosive environment. Flow induced loads, the high-temperature corrosive environment, radiation exposure, and high tensile stresses in bolts all contribute to aging related degradation of reactor internals. Another problem has been wear and fretting of instrument guide tubes. The paper also discusses age-related failures that have occurred to date in pressurized water reactors

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

    International Nuclear Information System (INIS)

    Loewenhielm, G.

    1978-03-01

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

  7. Insertion and drop of control rod in assembly simulations and parametric analysis

    International Nuclear Information System (INIS)

    Bosselut, D.; Andriambololona, H.; Longatte, E.; PAUTHENET, J.

    2005-01-01

    We present in this paper some results obtained with a numerical model of control rod behaviour. Drop time and insertion force are studied by means of a large parametric analysis. Fuel assembly deflection is the main parameter of this study. Several loads are taken into account in control rod displacement calculation: fluid load, gravity and friction force against the guide. Simulations are compared to measurements carried out in a full-scale prototype. The analysis of drop calculation shows that the friction cannot be neglected when fuel assembly is bowed. The main result for insertion force calculation is the important sensitivity to data entry. Determining exact deflection all along assembly with precision is needed. Otherwise, approximated friction force is obtained. (author)

  8. Parametric experiments on control rod heterogeneity in the framework of the Balzac program

    International Nuclear Information System (INIS)

    Kim, Y.C.; Palmiotti, G.; Salvatores, M.; Soule, R.

    1986-09-01

    The calculation of the so-called heterogeneity effect of control rod reactivity worth, is still to be considered one of the major source of uncertainty in assessing a power reactor control rod worth. The reason for the uncertainty is related both to the fact that at present no specific calculation tool can handle with sufficient precision this effect and that there is limited experimental validation of the approximate available methods. The present note will briefly recall the nature of the problem and its implications, in particular in the framework of the Super Phenix start-up experiments, and will indicate an experimental program on Masurca, aimed to a systematic study of the heterogeneity effect, in the frame of the Balzac program

  9. The BWR core simulator COSIMA with 2 group nodal flux expansion and control rod history

    International Nuclear Information System (INIS)

    Hoejerup, C.F.

    1989-08-01

    The boiling water simulator NOTAM has been modified and improved in several aspects: - The ''1 1/2'' energy group TRILUX nodal flux solution method has been exchanged with a 2 group modal expansion method. - Control rod ''history'' has been introduced. - Precalculated instrument factors have been introduced. The paper describes these improvements, which were considered sufficiently large to justify a new name to the programme: COSIMA. (author)

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

    International Nuclear Information System (INIS)

    Nikonov, S. P.; Velkov, K.; Pautz, A.

    2010-01-01

    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)

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

    International Nuclear Information System (INIS)

    Cho, Byungoh.

    1990-01-01

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

  12. Physico-chemical characterization of aerosols produced by a PWR control rods vaporization

    International Nuclear Information System (INIS)

    Rabu, B.; Pagano, C.; Tourasse, M.; Gros d'Aillon, L.; Boucenna, A.; Boulaud, D.; Dubourg, R.

    2000-01-01

    During a PWR type reactor accident, the aerosols produced by the vaporization of the control rods condition the released fission products evolution, for instance, the iodine or the tellurium. The EMAIC experiment has to characterize the aerosols emitted during the core degradation. The IPSN and EDF finances this program, realized at the CEA Grenoble. The results should allow the simulation of the aerosols source resulting from the vaporization to introduce in the ASTEC code, serious accident codes system. (A.L.B.)

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

    International Nuclear Information System (INIS)

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

    1992-01-01

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

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

    International Nuclear Information System (INIS)

    Biondo, C.D.; Carloni, J.G.; Aba, J.A.

    1987-01-01

    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)

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

    International Nuclear Information System (INIS)

    Sun, Jongoh; Kim, Gyeongho; Yoo, Yeonsik; Cho, Yeonggarp; Kim, Jong In

    2014-01-01

    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

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

    International Nuclear Information System (INIS)

    Yu, J. Y.; Choi, S.; Kim, J. H.; Lee, D. J.

    2007-01-01

    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

  17. Control of a high temperature supercritical pressure light water cooled and moderated reactor with water rods

    International Nuclear Information System (INIS)

    Ishiwatari, Yuki; Oka, Yoshiaki; Koshizuka, Seiichi

    2003-01-01

    The plant system of a supercritical pressure light water reactor (SCR) is once-through direct cycle. The whole coolant from the feedwater pumps is driven to the turbines. The core flow rate is less than 1/7 of that of a boiling water reactor. In the present design of the high temperature thermal reactor (SCLWR-H), the fuel assemblies contain many water rods in which the coolant flows downward. The stepwise responses of the SCLWR-H are analyzed against perturbations without a control system. Based on these analyses, a control system of the SCLWR-H is designed. The pressure is controlled by the turbine control valves. The main steam temperature is controlled by the feedwater pumps. The reactor power is controlled by the control rods. The control parameters are optimized by the test calculations to satisfy the criteria of both fast convergence and stability. The reactor is controlled stably with the designed control systems against various perturbations, such as setpoint change of the pressure, the main steam temperature and the core power, decrease in the feedwater temperature, and decrease in the feedwater flow rate. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Chul Gyo; Park, Cheol; Lee, Choong Sung; Kim, Hark Sung; Chae, Hee Taek [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2007-07-01

    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.

  19. 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)

    Tavolara, Ivan; Verburgh, Pierre; Menager, Antoine

    2010-01-01

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

  20. 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).

  1. Analysis of dose rates received around the storage pool for irradiated control rods in a BWR nuclear power plant.

    Science.gov (United States)

    Ródenas, J; Abarca, A; Gallardo, S

    2011-08-01

    BWR control rods are activated by neutron reactions in the reactor. The dose produced by this activity can affect workers in the area surrounding the storage pool, where activated rods are stored. Monte Carlo (MC) models for neutron activation and dose assessment around the storage pool have been developed and validated. In this work, the MC models are applied to verify the expected reduction of dose when the irradiated control rod is hanged in an inverted position into the pool. 2010 Elsevier Ltd. All rights reserved.

  2. Control rod guide tube wear in operating reactors; operating experience report. Technical report December 1977-December 1979

    International Nuclear Information System (INIS)

    Riggs, R.

    1980-04-01

    Evidence of control rod guide tube wear has been observed in operating pressurized water reactors. The cause of this wear is identified as flow-induced vibration of the control rods. This report describes the measures being taken by both the industry and the NRC to deal with this matter. The staff also presents its technical positions and requirements to support continued operation of the plants as of December 1979 pending completion of this generic effort

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

    International Nuclear Information System (INIS)

    Franco C, A.

    1997-01-01

    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)

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

    International Nuclear Information System (INIS)

    Sohn, Jae Min; Oh, Jong Myung; Park, Sung Jae; Choi, Myung Hwan; Cho, Man Soon; Kang, Young Hwan; Kim, Bong Goo

    2008-01-01

    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 centre 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 during 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 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

  5. Opium tincture versus methadone syrup in management of acute raw opium withdrawal: A randomized, double-blind, controlled trial.

    Science.gov (United States)

    Tabassomi, Farzaneh; Zarghami, Mehran; Shiran, Mohammad-Reza; Farnia, Samaneh; Davoodi, Mohsen

    2016-01-01

    The aim of this study was to evaluate the effectiveness of opium tincture versus methadone syrup in the management of acute withdrawal syndrome in opium dependent patients during the detoxification period. In this double-blind randomized controlled study, a total of 74 adult male raw opium dependent patients were treated with opium tincture or methadone syrup 2 times daily for 5 consecutive days. Detoxification was initiated by tapered dose reductions to reach abstinence. At the end of the 10th day, the medications were discontinued. The Objective Opioid Withdrawal Scale was used to assess withdrawal symptoms every day. Significant decreases on the Objective Opioid Withdrawal Scale were found for both treatment methods during the study period (p Opium tincture can be considered as a potential substitute for methadone syrup for suppression of raw opium withdrawal symptoms, with minimal adverse effects.

  6. Efficacy of extended-release tramadol for treatment of prescription opioid withdrawal: A two-phase randomized controlled trial*

    Science.gov (United States)

    Lofwall, Michelle R.; Babalonis, Shanna; Nuzzo, Paul A.; Siegel, Anthony; Campbell, Charles; Walsh, Sharon L.

    2013-01-01

    Background Tramadol is an atypical analgesic with monoamine and modest mu opioid agonist activity. The purpose of this study was to evaluate: 1) the efficacy of extended-release (ER) tramadol in treating prescription opioid withdrawal and 2) whether cessation of ER tramadol produces opioid withdrawal. Methods Prescription opioid users with current opioid dependence and observed withdrawal participated in this inpatient, two-phase double blind, randomized placebo-controlled trial. In Phase 1 (days 1-7), participants were randomly assigned to matched oral placebo or ER tramadol (200 or 600 mg daily). In Phase 2 (days 8-13), all participants underwent double blind crossover to placebo. Breakthrough withdrawal medications were available for all subjects. Enrollment continued until 12 completers/group was achieved. Results Use of breakthrough withdrawal medication differed significantly (popioid withdrawal. Mild opioid withdrawal occurred after cessation of treatment with 600 mg tramadol. These data support the continued investigation of tramadol as a treatment for opioid withdrawal. PMID:23755929

  7. Efficacy of extended-release tramadol for treatment of prescription opioid withdrawal: a two-phase randomized controlled trial.

    Science.gov (United States)

    Lofwall, Michelle R; Babalonis, Shanna; Nuzzo, Paul A; Siegel, Anthony; Campbell, Charles; Walsh, Sharon L

    2013-11-01

    Tramadol is an atypical analgesic with monoamine and modest mu opioid agonist activity. The purpose of this study was to evaluate: (1) the efficacy of extended-release (ER) tramadol in treating prescription opioid withdrawal and (2) whether cessation of ER tramadol produces opioid withdrawal. Prescription opioid users with current opioid dependence and observed withdrawal participated in this inpatient, two-phase double blind, randomized placebo-controlled trial. In Phase 1 (days 1-7), participants were randomly assigned to matched oral placebo or ER tramadol (200 or 600 mg daily). In Phase 2 (days 8-13), all participants underwent double blind crossover to placebo. Breakthrough withdrawal medications were available for all subjects. Enrollment continued until 12 completers/group was achieved. Use of breakthrough withdrawal medication differed significantly (popioid withdrawal. Mild opioid withdrawal occurred after cessation of treatment with 600 mg tramadol. These data support the continued investigation of tramadol as a treatment for opioid withdrawal. Copyright © 2013 Elsevier Ireland Ltd. All rights reserved.

  8. Performance criteria of control rod absorbers of new VVER reactors and a possibility to increase their lifetime

    International Nuclear Information System (INIS)

    Zakharov, A.V.; Risovany, V.D.; Vasilchenko, I.N.; Kurakin, K.Y.; Kushmanov, S.A.; Makhin, V.M.

    2015-01-01

    At present, the lifetime of control rod absorbers of the existing VVER-100 and constructed VVER-1200 is designed to be ten years; they are operated as control rods for three years and as scram rods for seven years. The examination of spent absorbers did not show any degradation in their performance so their lifetime specification might be too conservative. The criteria of the control rod absorbers performance, together with an adequate computational model, can be a methodological basis to develop a system to manage the VVER-1000 control rod absorbers lifetime to fully use their design and material capacities. The criteria for control rod absorbers performance can be split into 5 groups: 1) criteria for physical efficiency, 2) criteria for dynamic characteristics, 3) criteria for radiation resistance, 4) criteria for thermo-mechanical resistance and 5) criteria for corrosion resistance. There is a need in an analytical model to define the neutron-physical and thermo-hydraulic operating conditions as well as thermo-mechanical state of an absorbing element. At present, the key components of the model have been developed (neutron physics and thermo-hydraulics) and the thermo-mechanical model is under development. The analytical models have been used for the justification of a design of absorbing elements combining dysprosium hafnate pellets at the bottom and boron carbide at the top

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

    International Nuclear Information System (INIS)

    Eto, Motokuni; Ishiyama, Shintaro; Fukaya, Kiyoshi; Saito, Tamotsu; Ishihara, Masahiro; Hanawa, Satoshi.

    1998-01-01

    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)

  10. 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.

  11. Conceptual Design Study on Electromagnets of Control Rod Drive Mechanism of a SFR

    International Nuclear Information System (INIS)

    Lee, Jaehan; Koo, Gyeonghoi

    2013-01-01

    The prototype SFR has six primary control rod assemblies(CRAs) and three secondary shutdown assemblies. The primary control system is used for power control, burnup compensation and reactor shutdown in response to demands from the plant control or protection systems. This paper describes the design concept of primary control rod drive mechanism shortly, and performs the parametric design studies for the electromagnet device of the drive mechanism to maximize CRA gripping force. The electromagnetic core usually confines and guides the magnetic field. The major parameters influenced on the electromagnetic force are the geometry and arrangement of the electromagnet and armature for a given coil specification. A typical equation calculating the electromagnetic force for a solenoid type is represented in equation. The first one is the increasing of the flux cross section area (Α c , Α g ) in magnetic field connecting of air gap, armature and electromagnets. Secondly, the reducing of the path lengths (l c , l g ) of the armature and electromagnet makes the magnetic flux (Β) resistance to be low. An electromagnet field analyses are performed for the initial design values of the electromagnet device. The gripping force is about 3 times of CRA weight when one coil is power on. The parametric studies on air gap, core sizes configuring of the electromagnet cores are performed to maximize the electromagnetic force

  12. 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)

  13. Microstructural engineering applied to the controlled cooling of steel wire rod: Part I. Experimental design and heat transfer

    Science.gov (United States)

    Campbell, P. C.; Hawbolt, E. B.; Brimacombe, J. K.

    1991-11-01

    The goal of this study was to develop a mathematical model which incorporates heat flow, phase transformation kinetics, and property-structure-composition relationships to predict the mechanical properties of steel rod being control cooled under industrial conditions. Thus, the principles of microstructural engineering have been brought to bear on this interdisciplinary problem by combining computer modeling with laboratory measurements of heat flow, austenite decomposition kinetics, microstructure and mechanical properties, and industrial trials to determine heat transfer and obtain rod samples under known conditions. Owing to the length and diversity of the study, it is reported in three parts,[8191]the first of which is concerned with the heat flow measurements. A relatively simple and reliable technique, involving a preheated steel rod instrumented with a thermocouple secured at its centerline, has been devised to determine the cooling rate in different regions of the moving bed of rod loops on an operating Stelmor line. The measured thermal response of the rod has been analyzed by two transient conduction models (lumped and distributed parameter, respectively) to yield overall heat-transfer coefficients for radiation and convection. The adequacy of the technique has been checked by cooling instrumented rods under well-defined, air crossflow conditions in the laboratory and comparing measured heat-transfer coefficients to values predicted from well-established equations. The industrial thermal measurements have permitted the characterization of a coefficient to account for radiative interaction among adjacent rod loops near the edge and at the center of the bed.

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

    International Nuclear Information System (INIS)

    Solbrig, C.W.; Stevens, W.W.

    1990-01-01

    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

  15. Development of a 3-D simulation analysis system for PWR control rod drive mechanism

    International Nuclear Information System (INIS)

    Tanaka, Akio; Futahashi, Kensuke; Takanabe, Kiyoshi; Kurimura, Chikara; Kato, Jungo; Hara, Hidekiyo

    2008-01-01

    A 3-D virtual analysis system to analyze the motion of control rod drive mechanism (CRDM) was developed. The analysis system consists of a 3-D model established as per the actual dimensions and interfaces of CRDM parts and a routine to calculate the forces acting on the mechanism, and was verified by mock-up test using the same equipment as the actual product. The analysis system is useful for functional evaluation in maintenance or to factor out root causes in the case of malfunction of CRDM

  16. Probabilistic safety analysis for control rod drive system of ET-RR-1

    International Nuclear Information System (INIS)

    Nasr, M.; Nasser, O.

    1988-01-01

    The International Atomic Energy Agency (IAEA) co-ordinated a Research programme on Probabilistic Safety Analysis (PSA) for research reactors; with the participation of several countries. In the framework of this project (Project Int. 9/063) the Egyptian Atomic Energy Authority decided to perform a PSA study on the ET-RR-1 (Egypt Thermal Research Reactor). The study is conducted in collaboration between the nuclear regulatory and safety centre (NRSC) and the reactor department of the nuclear research centre at Inchass. The present work is a part of the PSA study on ET-RR- it is concerning a probabilistic safety analysis of the control rod drive mechanism

  17. Calibration of the G2/G3 control rods - history and methods

    International Nuclear Information System (INIS)

    Fenart, H.

    1965-01-01

    This report describes the various control rod calibrations which have been carried out on G2 and G3 reactors since the inauguration of these reactors. The 'historical' section is followed by a concise analysis of the various possible calibration methods, together with an example of the application of each of the methods used on G2 and G3. Finally some indications are given concerning the accuracy to be expected from these calibrations, as well as the best method applicable to these reactors. (author) [fr

  18. Dealing with control rod guide tube support pin cracking in French PWRs

    International Nuclear Information System (INIS)

    Guicherd, L.

    1984-01-01

    Cracking and failure of control rod guide tube support pins has been encountered at a number of PWRs around the world. To deal with the problem, the French embarked on an extremely ambitious backfitting programme, involving the installation of replacement pins at all their operating 900MWe units. This highly successful programme, which will be completed in 1985, has been carried out with very low occupational doses and, in the last two years, has required no extensions to annual refuelling outage periods at the plants concerned. The French approach has involved a number of innovations, which should be of considerable interest to other PWR owners worldwide. (author)

  19. The construction design of ball bearings used in the control rod driving mechanisms of PWRs

    International Nuclear Information System (INIS)

    Leng Chengmu; Huang Chongming; Chen Jianting.

    1986-01-01

    According to the operation conditions of ball bearings used in the control rod driving mechanisms of PWRs, this paper has analysed and discussed the problems that must be taken into account in the construction design of this ball bearing. It includes: a discussion about the reasonable selection of construction parameters of the bearing, deduction of the relationship between bearing clearance and contact angle, and the emphasis on the significance of assembling accuracy and torque measurement in the assurance of operational performance of the bearings. These experiences may be somewhat valuable for the design and application of this kind of ball bearing

  20. The Eclipse system for surveying the guide tubes of control rod clusters

    International Nuclear Information System (INIS)

    Pace, Y.M.

    2008-01-01

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

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

    International Nuclear Information System (INIS)

    Leme, Francisco Louzano

    2003-01-01

    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)

  2. An economic analysis of BWR control rod blade management strategies. Final report

    International Nuclear Information System (INIS)

    Welsh, J.

    1995-12-01

    Nuclear power plants have available a number of alternative courses of action that can contribute to the reduction of personnel exposure to radiation. Possible actions at boiling water reactor (BWR) plants include accelerating the replacement of high-cobalt control rod blades (CRB) or the blades' high-cobalt pins and rollers with low or non-cobalt substitutes. To help utilities understand the exposure reduction and the economic costs and benefits associated with management alternatives, such as accelerated replacement of blades, pins and rollers, EPRI has initiated a project called Cost/Benefit Software for Analyses of Radiation Control Measures (RP1935-32). Through this project EPRI will incorporate engineering-economic techniques into a series of analytical tools that will provide useful insights about alternative exposure reduction options. Prototype software has been developed in an Excel worksheet to analyze issues associated with BWR control rod blade management options. The CRB replacement problem framework and analysis methodology incorporated into the software tool will help plant managers consider explicitly key engineering and economic issues that are relevant to exposure reduction decisions. This tool generates results that can help plant managers make decisions that are fiscally wise by showing all the cost and benefit implications associated with a management action under consideration. This report describes the general analytical approach for evaluating exposure reduction alternatives. The methodology used to analyze blade and pin and roller replacement alternatives, and the results of a case study application of the methodology and the software prototype at Commonwealth Edison

  3. Rapid-L Operator-Free Fast Reactor Concept Without Any Control Rods

    International Nuclear Information System (INIS)

    Kambe, Mitsuru; Tsunoda, Hirokazu; Mishima, Kaichiro; Iwamura, Takamichi

    2003-01-01

    The 200-kW(electric) uranium-nitride-fueled lithium-cooled fast reactor concept 'RAPID-L' to achieve highly automated reactor operation has been demonstrated. RAPID-L is designed for a lunar base power system. It is one of the variants of the RAPID (Refueling by All Pins Integrated Design) fast reactor concept, which enables quick and simplified refueling. The essential feature of the RAPID concept is that the reactor core consists of an integrated fuel assembly instead of conventional fuel subassemblies. In this small-size reactor core, 2700 fuel pins are integrated and encased in a fuel cartridge. Refueling is conducted by replacing a fuel cartridge. The reactor can be operated without refueling for up to 10 yr.Unique challenges in reactivity control systems design have been addressed in the RAPID-L concept. The reactor has no control rod but involves the following innovative reactivity control systems: lithium expansion modules (LEM) for inherent reactivity feedback, lithium injection modules (LIM) for inherent ultimate shutdown, and lithium release modules (LRM) for automated reactor startup. All these systems adopt 6 Li as a liquid poison instead of B 4 C rods. In combination with LEMs, LIMs, and LRMs, RAPID-L can be operated without an operator. This reactor concept is also applicable to the terrestrial fast reactors. In this paper, the RAPID-L reactor concept and its transient characteristics are presented

  4. RAPID-L and RAPID operator free fast reactor concepts without any control rods

    International Nuclear Information System (INIS)

    Kambe, Mitsuru; Tsunoda, Hirokazu; Nakazima, Kiyoshi; Iwamura, Takamichi

    2003-01-01

    The 200 kWe uranium-nitride fueled lithium-cooled fast reactor concept RAPID-L' for lunar base power system, and the 1000 kWe U-Pu-Zr metal fueled sodium-cooled fast reactor concept 'RAPID' for terrestrial power system have been demonstrated. These reactors are characterized by RAPID (Refueling by All Pins Integrated Design) refueling concept, which enable quick and simplified refueling. The essential feature of RAPID concept is that the reactor core consists of an integrated fuel assembly (IFA) instead of conventional fuel subassemblies. In this small size reactor core, all the fuel pins are integrated altogether and encased in a fuel cartridge. Refueling is conducted by replacing an IFA. The reactor can be operated without refueling for up to 10 years. Unique challenges in reactivity control systems design have been addressed in these reactors. They have no control rod, but involve the following innovative reactivity control systems: Lithium Expansion Modules (LEM) for inherent reactivity feedback, Lithium Injection Modules (LIM) for inherent ultimate shutdown, and Lithium Release Modules (LRM) for automated reactor startup. All these systems adopt lithium-6 as a liquid poison instead of B4C rods. In combination with LEMs, LIMs and LRMs, RAPID-L and RAPID can be operated without operator. In this paper, design characteristics of RAPID-L and RAPID reactor concepts are discussed. (author)

  5. Device for relocating fuel elements and control rods in a core reactor

    International Nuclear Information System (INIS)

    Hoffmeister, B.; Schwarz, W.

    1976-01-01

    A device is described for changing the location of fuel elements and control rods in a core reactor with a guiding mast which is mounted on a movable working platform for rotation about a vertical axis and which is provided with a gripping body for grasping the fuel elements and control rods, said gripping body being suspended on a lifting mechanism and being displaceable in vertical direction within the guiding mast while being prevented from rotating relative to the guiding mast. The device furthermore comprises winding means for storing supply lines which are introduced at the top into the guiding mast and the gripper body. The device also includes power operated means for actuating the lifting mechanism. The winding means and lifting mechanism rotate with the mast and are so arranged in the mast that any water dripping therefrom runs down the inside of the mast. Supply conduits leading to the mast are connected thereby by slack loops which permit 360 0 of rotation of the mast. 11 claims, 6 drawing figures

  6. Aging mechanisms in the Westinghouse PWR [Pressurized Water Reactor] Control Rod Drive system

    International Nuclear Information System (INIS)

    Gunther, W.; Sullivan, K.

    1991-01-01

    An aging assessment of the Westinghouse Pressurized Water Reactor (PWR) Control Rod System (CRD) has been completed as part of the US NRC's Nuclear Plant Aging Research, (NPAR) Program. This study examined the design, construction, maintenance, and operation of the system to determine its potential for degradation as the plant ages. Selected results from this study are presented in this paper. The operating experience data were evaluated to identify the predominant failure modes, causes, and effects. From our evaluation of the data, coupled with an assessment of the materials of construction and the operating environment, we conclude that the Westinghouse CRD system is subject to degradation which, if unchecked, could affect its safety function as a plant ages. Ways to detect and mitigate the effects of aging are included in this paper. The current maintenance for the control rod drive system at fifteen Westinghouse PWRs was obtained through a survey conducted in cooperation with EPRI and NUMARC. The results of the survey indicate that some plants have modified the system, replaced components, or expanded preventive maintenance. Several of these activities have effectively addressed the aging issue. 2 refs., 2 figs., 2 tabs

  7. Seismic proving test of PWR core internals (inserting function of control rod during earthquakes)

    International Nuclear Information System (INIS)

    Kawakami, S.; Akiyama, H.; Shibata, H.; Watabe, M.; Ichikawa, T.

    1989-01-01

    A series of seismic reliability proving tests of nuclear power plant facilities has been carried out by the Nuclear Power Engineering Test Center (NUPEC), using a large-scale, high-performance vibration table at Todatsu Engineering Laboratory, sponsored by the Ministry of International Trade and Industry (MITI). In 1985, a seismic proving test of PWR reactor core internals was conducted on a full scale test model. The results showed that the test components proved to have the safe and reliable function of control rod cluster insertion during the seismic proving test and to have the structural soundness against earthquake. Subsequently, detailed analyses and evaluation of these test results were performed, and analysis methods for evaluating the strength to withstand earthquakes were established. After that, seismic analyses and evaluations of actual reactor core internals were performed by these analysis methods, and the safety and reliability of PWR reactor core internals were confirmed. This paper mainly focuses on the function of control rod cluster insertion during an earthquake

  8. Improved control rod drive handling equipment for BWRs [boiling-water reactors]: Final report

    International Nuclear Information System (INIS)

    Turner, A.P.L.; Gorman, J.A.

    1987-08-01

    Improved equipment for removing and replacing control rod drives (CRDs) in BWR plants has been designed, built and tested. Control rod drives must be removed from the reactor periodically for servicing. Removal and replacement of CRDs using equipment originally supplied with the plant has long been recognized as one of the more difficult and highest radiation exposure maintenance operations that must be performed at BWR plants. The improved equipment was used for the first time at Quad Cities, Unit 2, during a Fall 1986 outage. The trial of the equipment was highly successful, and it was shown that the new equipment significantly improves CRD handling operations. The new equipment significantly simplifies the sequence of operations required to lower a CRD from its housing, upend it to a horizontal orientation, and transport it out of the reactor containment. All operations of the new equipment are performed from the undervessel equipment handling platform, thus, eliminating the requirement for a person to work on the lower level of the undervessel gallery which is often highly contaminated. Typically, one less person is required to operate the equipment than were used with the older equipment. The new equipment incorporates a number of redundant and fail safe features that improve operations and reduce the chances for accidents

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

    Directory of Open Access Journals (Sweden)

    A. Kaliatka

    2008-01-01

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

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

    International Nuclear Information System (INIS)

    Sainz M, E.

    1996-01-01

    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)

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

    International Nuclear Information System (INIS)

    Sohn, Jae Min; Oh, Jong Myung; Park, Sung Jae; Choi, Myung Hwan; Cho, Man Soon; Kang, Young Hwan; Kim, Bong Goo

    2008-01-01

    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 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

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

    Science.gov (United States)

    Sun, Jingchang; Zhao, Ting; Ma, Zhangwei; Li, Ming; Chang, Cheng; Liang, Hongwei; Bian, Jiming; Li, Chengren

    2015-09-01

    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.

  13. High patient-parent satisfaction with magnetically controlled growing rod treatment in EOS

    DEFF Research Database (Denmark)

    Skov, Simon Toftgaard; Bünger, Cody

    Background: Magnetic controlled growing rod (MCGR) use is increasing worldwide and was by some considered a revolution in EOS management. Now, reports on MCGR-failures and complications similar to traditional growing rods are being reported but the treatment is still considered to be a major......-lengthening procedure(s). The parents answered a 7 question non-validated questionnaire before lengthening at the day of the procedure. The answers were in a 0-10 numeric scale score. Results: Thirteen patients, median age 11 years (range 6-18) at MCGR treatment. Around the procedure, the physical strain on the patient...... was rated to a mean 0.2 (range 0-1), the psychological strain on the patient was rated to a mean 1.7 (range 0-7), pain was rated to a mean 0.8 (range 0-2), and the anxiety level of the parents was rated to a mean 0.8 (range 0-5) (0 none, 10 highest imaginable). The overall patient/parent satisfaction...

  14. Degradation and oxidation of B 4C control rod segments at high temperatures

    Science.gov (United States)

    Steinbrück, M.

    2010-05-01

    Extensive series of test were performed of the degradation of boron carbide absorber rods and the oxidation of the resultant absorber melts. Various types of control rod segments made of commercial materials used in French 1300 MW PWRs were investigated in the temperature range between 800 °C and 1700 °C in a steam atmosphere. The gaseous reaction products were analyzed quantitatively by mass spectroscopy for evaluation of the oxidation rates. Extensive post-test examinations were performed by light microscopy, scanning electron microscopy as well as EDX and Auger spectroscopy. Rapid melt formation due to eutectic interactions of stainless steel (cladding tube) and B 4C, on the one hand, and steel and Zircaloy-4 (guide tube), on the other hand, was observed at temperatures above 1250 °C. Complex multi-component, multi-phase melts were produced. ZrO 2 oxide scale on the outside kept the melt within the guide tube, thus preventing its early relocation and oxidation. Rapid oxidation of the absorber melts and remaining boron carbide pellets took place after failure of the protective oxide shell above 1450 °C. Only very little methane was produced in these tests which is of interest in fission product gas chemistry because of the production of organic iodine.

  15. Interpretation of out of line control rod experiments for 1300 MWE PWR

    Energy Technology Data Exchange (ETDEWEB)

    Leroy, J.L.; Garcia-Fernandez, L.

    1988-01-01

    The present note summarizes the studies we performed recently in order to search a 2D reconstruction procedure for the 1300 MWE PWR power shape, starting from data coming out from thermocouples placed on several fuel assemblies. In classical PWR design, only a few assemblies are equipped with measurement devices, so that it is necessary to interpolate among measure points in order to obtain a complete coverage of the core. A mathematical approach based on the splitting of the power into a reference steady state nominal shape and some ''influence'' and harmonic functions was chosen. The reference steady state power shape, which corresponds to the full power operating mode, is obtained via direct mobile chamber measurements. The perturbations due to the control rod movements are accounted for by specific ''influence'' functions: moreover, harmonics are used to reconstruct the minor effects due to xenon tilts, rod out of line positions and all actual mechanical and thermohydraulic inhomogeneities. The weighting coefficients of the functions are evaluated by a least square method, starting from the distribution of the deviations among the measurements and the reference values.

  16. Safety assessment for the above ground storage of Cadmium Safety and Control Rods at the Solid Waste Management Facility

    International Nuclear Information System (INIS)

    Shaw, K.W.

    1993-11-01

    The mission of the Savannah River Site is changing from radioisotope production to waste management and environmental restoration. As such, Reactor Engineering has recently developed a plan to transfer the safety and control rods from the C, K, L, and P reactor disassembly basin areas to the Transuranic (TRU) Waste Storage Pads for long-term, retrievable storage. The TRU pads are located within the Solid Waste Management Facilities at the Savannah River Site. An Unreviewed Safety Question (USQ) Safety Evaluation has been performed for the proposed disassembly basin operations phase of the Cadmium Safety and Control Rod Project. The USQ screening identified a required change to the authorization basis; however, the Proposed Activity does not involve a positive USQ Safety Evaluation. A Hazard Assessment for the Cadmium Safety and Control Rod Project determined that the above-ground storage of the cadmium rods results in no change in hazard level at the TRU pads. A Safety Assessment that specifically addresses the storage (at the TRU pads) phase of the Cadmium Safety and Control Rod Project has been performed. Results of the Safety Assessment support the conclusion that a positive USQ is not involved as a result of the Proposed Activity

  17. Current status and Future Works in Dynamic Control Rod Worth Measurement Method in KOREA

    International Nuclear Information System (INIS)

    Lee, Eun-ki; Shin, Ho-chul

    2015-01-01

    Actually all ex-core detector signal data processing and extracting the final conclusion about the single control bank was performed by the 3rd generation of Digital Reactivity Computer System (DRCS). From 2006 to 2016, about 250 control bank worths were measured and the difference between measured and calculated worth of individual bank were several % and total rod worth differences of each cycle were less than 3%. However there were a few odd cases showing the individual difference greater than 15% which is the criteria. And some OPR1000 nuclear power plants built recently use fission chambers instead of traditional uncompensated ion chambers. To consider those conditions, any modification of the DCRM method and DRCS were requested. In this paper, short description about DCRM method, current status of DCRM modification and future works are discussed. About 10 cases among about 250 control bank worth measurement with original DCRM method with UIC signals result in very heavy fluctuation on reactivity curve and the difference approaches 15%, the individual limit. The electrometer signals processing method were the main cause. To overcome this problem, a modified reactivity computer system using own current treatment logic was designed. It shows the reactivity fluctuation can be reduced dramatically in case of UIC signal. Recently the modified DRCS coupled INVESE code applied Westinghouse 2 Loop plant and gave very good results except a control bank whose measured rod worth was shown of 18% difference from the estimated value. The results were same for three repeat tests. All possible causes are examined from computer codes to detail data acquisition system

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-07-01

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

  19. A control absorber rod in Phenix. Comparison of calculated and measured worths

    International Nuclear Information System (INIS)

    Humbert, G.; Petiot, R.; Coulon, P.

    1982-09-01

    For SPX1 design purposes, a so called ''SAC'' (complementary shut down system) rod has been set up at the center of the Phenix core. This gave the opportunity of making a design rod worth calculation check on this simple configuration. Calculations methods are shortly described and an experiment to calculation comparison is made. The type of agreement obtained is coherent with the uncertainty now applied to design rod worth calculation

  20. Adjusts of control rod cross sections and its utilization in power distribution calculations for Angra-1 reactor

    International Nuclear Information System (INIS)

    Pina, C.M. de

    1981-01-01

    One of the most important part in neutronics calculations is the study of core behavior with inserted control rods. The first stage of this calculations consists in generating equivalent microscopic cross sections for the basic cells containing fuel or absorbed material. The cross sections will be then adjusted. The choice of parameters that help in those adjustments, were obtained by the comparisons of data coming from the control rod supercell calculations with the Hammer and Citation computer codes. The effect of those adjustments in core integral parameters was evaluated; in this work only the core power two-dimensional distribution calculations with the D bank completely inserted, is studied. (E.G.) [pt

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

    International Nuclear Information System (INIS)

    Ortiz S, J.J.; Requena R, I.

    2003-01-01

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

  2. Study on electromagnetism force of CARR control rod drive mechanism experimental machine

    International Nuclear Information System (INIS)

    Zhu Xuewei; Zhen Jianxiao; Wang Yulin; Jia Yueguang; Yang Kun; Yin Haozhe

    2015-01-01

    With the aim of acquiring electromagnetic force and electromagnetic field distributions of control rod drive mechanism (CRDM) in China Advanced Research Reactor (CARR), the force analysis on the CRDM was taken. Manufacturing the experimental machine, the electromagnetic force experiment was taken on it. The electromagnetic field and electromagnetic force simulation analyses of experimental machine were taken, working out distribution data of electromagnetic force and magnetic induction intensity distribution curve, and the effects of permanent magnetic field on electromagnetic field and structure parameters on electromagnetic force. The simulation value is accord with experiment value, the research results provide a reference to electromagnetic force study on CRDM in CARR, and also provide a reference to design of the same type CRDM. (authors)

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

    International Nuclear Information System (INIS)

    Ansari, S.A.; Ayazuddin, S.K.

    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.)

  4. Stress and fatigue analysis for lower joint of control rod drive mechanisms seal house

    International Nuclear Information System (INIS)

    Shao Xuejiao; Zhang Liping; Du Juan; Xie Hai

    2013-01-01

    Two kinds of seal houses for control rod drive mechanisms which have different thickness of the lower seal ring was analyzed for its stress and fatigue by finite element method. In the fatigue computation, all the transitions were grouped into several groups, and then the elastoplastic strain correction factor was modified by analyzing thermal and mechanical load separately referring the rules of RCC-M 2002. The results show that the structure with thicker seal ring behaves more safely than the other one except in the second condition. Meanwhile, the amplify of the primary and secondary stress as well as fatigue usage factor can be reduced by regrouping the transients. The precision of fatigue usage factor can be elevated using modified K e when the amplify of the primary and secondary stress is large to some extent produced by both thermal and mechanical loads. (authors)

  5. Analysis of high burnup fuel behavior under control rod ejection accident in Korea standard nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chan Bok; Lee, Chung Chan; Kim, Oh Hwan; Kim, Jong Jin [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1996-07-01

    Test results of high burnup fuel behavior under RIA(reactivity insertion accident) indicated that fuel might fail at the fuel enthalpy lower than that in the current fuel failure criteria was derived by the conservative assumptions and analysis of fuel failure mechanisms, and applied to the analysis of control rod ejection accident in the 1,000 MWe Korea standard PWR. Except that three dimensional core analysis was performed instead of conventional zero dimensional analysis, all the other conservative assumptions were kept. Analysis results showed that less than on percent of the fuel rods in the core has failed which was much less than the conventional fuel failure fraction, 9.8 %, even though a newly derived fuel failure criteria -Fuel failure occurs at the power level lower than that in the current fuel failure criteria. - was applied, since transient fuel rod power level was significantly decreased by analyzing the transient fuel rod power level was significantly decreased by analyzing the transient core three dimensionally. Therefore, it can be said that results of the radiological consequence analysis for the control rod ejection accident in the FSAR where fuel failure fraction was assumed 9.8 % is still bounding. 18 tabs., 48 figs., 39 refs. (Author).

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

    International Nuclear Information System (INIS)

    Rajan Babu, V.; Veerasamy, R.; Patri, Sudheer; Ignatius Sundar Raj, S.; Kumar Krovvidi, S.C.S.P.; Dash, S.K.; Meikandamurthy, C.; Rajan, K.K.; Puthiyavinayagam, P.; Chellapandi, P.; Vaidyanathan, G.; Chetal, S.C.

    2010-01-01

    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 CO 2 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.

  7. Rodding Surgery

    Science.gov (United States)

    ... usually undertaken as a scheduled elective procedure. An optimal age for a first rodding surgery has not ... which may prevent or postpone the need for replacement. The smallest diameter expanding rods are still too ...

  8. Confirming the Multidimensionality of Psychologically Controlling Parenting among Chinese-American Mothers: Love Withdrawal, Guilt Induction, and Shaming

    Science.gov (United States)

    Yu, Jing; Cheah, Charissa S. L.; Hart, Craig H.; Sun, Shuyan; Olsen, Joseph A.

    2015-01-01

    Despite the theoretical conceptualization of parental psychological control as a multidimensional construct, the majority of previous studies have examined psychological control as a unidimensional scale. Moreover, the conceptualization of shaming and its associations with love withdrawal and guilt induction are unclear. The current study aimed to…

  9. Dynamic Rod Worth Measurement

    International Nuclear Information System (INIS)

    Chao, Y.A.; Chapman, D.M.; Hill, D.J.; Grobmyer, L.R.

    2000-01-01

    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. 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)

  11. A test research on ventilative well-distributivity under normal temperature for a control rod drive mechanism (Continuous article)

    International Nuclear Information System (INIS)

    Zhu Longxing

    1989-01-01

    A test for cooling of the control rod drive mechnism under normal temperature is described. The relationship between the unbalanced cofficient and the frictional resistance and wind velocity is found by comparing the ventilation in plate top structure of reactor with that in global top structure of reactor

  12. Experience feedback on rod cluster control assemblies; Ligne de commande de grappe rex des expertises des differents composants

    Energy Technology Data Exchange (ETDEWEB)

    Bourgoin, J. [Electricite de France (EDF), 37 - Avoine (France). Groupe des Laboratoires; Gouthier, A. [Electricite de France (EDF), 93 - Saint Denis (France). Dept. Maintenance

    1998-07-01

    aking into account the safety function provided by Rod Cluster Control Assemblies, metallurgical evaluations on drive line components give an essential experience feedback. This paper presents the EDF assessment and evidences the variety of cases: design, operating and maintenance. The comprehension of degradation mechanisms: wear, fatigue, swelling, creep,... contributes to apply the suitable corrective actions. (authors)

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

    International Nuclear Information System (INIS)

    1997-01-01

    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)

  14. Traditional growing rod versus magnetically controlled growing rod for treatment of early onset scoliosis: Cost analysis from implantation till skeletal maturity.

    Science.gov (United States)

    Wong, Carlos King Ho; Cheung, Jason Pui Yin; Cheung, Prudence Wing Hang; Lam, Cindy Lo Kuen; Cheung, Kenneth Man Chee

    2017-01-01

    To compare the yearly cost involved per patient in the use of magnetically controlled growing rod (MCGR) and traditional growing rods (TGRs) in the treatment of early onset scoliosis (EOS) and to assess the overall cost burden of MCGR with reference to patient and health-care infrastructure. For a hypothetical case of a 5-year-old girl with a diagnosis of EOS, a decision-tree model using TreeAge Software was developed to simulate annual health state transitions and compare the 8-year accumulative direct, indirect, and total cost among the four groups: (1) dual MCGRs with exchange every 2 years, (2) dual MCGRs with exchange every 3 years, (3) TGR with surgical distraction every year, and (4) TGR with surgical distraction every 6 months. Base-case values and ranges of clinical parameters reflecting complication rate after each type of surgical distraction were determined from a review of literature and expert opinion. Government gazette and expert opinion provided cost estimation of growing rods, surgeries, surgical complications, and routine follow-up. Microsimulation of 1000 individuals was conducted to test the variation in total direct costs (in 2016 Hong Kong dollars (HKD)) between individuals, and estimated the standard deviations of total direct costs for each group. Over the projected treatment period, indirect costs incurred by patients and family were higher for the MCGR as compared to the TGR. However, the total costs incurred by MCGR groups (group 1: HKD164k; group 2: HKD138k) were lower than those incurred by TGR groups (group 3: HKD191k; group 4: HKD290k). Although the accumulative costs of three groups (TGR with distraction every year and MCGR replacing every 2 and 3 years) were approaching each other in the first 2 years after initial implantation, at year 3 the accumulative cost of MCGR exchange every 2 years was HKD36k more than the yearly TGR surgery due to the cost of implant exchange. The cost incurred by both the MCGR groups was less than that

  15. Identification model of an accidental drop of a control rod in PWR reactors using thermocouple readings and radial basis function neural networks

    International Nuclear Information System (INIS)

    Souza, T.J.; Medeiros, J.A.C.C.; Gonçalves, A.C.

    2017-01-01

    Highlights: • An alternative model capable of identifying the control rod that has accidentally dropped. • The identification model is based in readings of the thermocouples. • Radial basis function neural network is applied to predict the temperatures in control rod positions. - Abstract: The accidental dropping of a control rod may cause the reactor to operate unsafely. In this type of event, there is a distortion in the distribution of power and temperature in the core may exceed operating limits reactor safe. This work aims to develop an alternative model capable of identifying, at any time of the cycle, the control rod that has accidentally dropped at the core of a PWR reactor, using the readings of the thermocouples in order to minimize possible losses. The model assumes that in a possible drop of a control rod, the largest temperature change occurs in the position where the control rod is inserted. Considering the fact that there are no temperature gauges in all control rod positions, the proposed model uses radial basis function (RBF) neural networks to make a reconstruction of temperatures in these positions from the measurements of the thermocouples at the time of the accidental drop. The study found that the predictions of the temperatures made by the RBF neural networks showed good results, which enables the identification of the control rod dropped accidentally in the core, by simple inference of the fuel assembly of lowest temperature among temperatures reconstructed.

  16. Fault tolerant, multiplexed control rod position detection and indication system for nuclear power plants

    International Nuclear Information System (INIS)

    Dufek, W.L.; Jelovich, J.J.; Neuner, J.A.

    1977-01-01

    The majority of Westinghouse nuclear plants placed in service thus far have incorporated a Rod Position Indication system based upon an analog design philosophy. This system, while meeting all functional and accuracy requirements, has proven somewhat cumbersome, particularly in the area of initial field calibration and maintenance. This paper describes a new Digital Rod Position Indication system (DRPI) developed for use with pressurized water reactors. The system is based upon a digital design philosophy and meets all previous design constraints and environmental requirements. Further, fault tolerance, improved accuracy, interference from adjacent rods and the elimination of adjustments and calibration has been provided

  17. Characterization of the withdrawal phase in a swine controlled intestinal donation after circulatory death model.

    Science.gov (United States)

    Guo, Mingxiao; Li, Linlin; Lu, Chunlei

    2014-10-04

    Transplantation of donation after cardiac death (DCD) intestine has higher rates of organ failure and complications. Fortunately, this is less grievous in a subclass of DCD called controlled (CDCD), those with irreversible but incomplete brain injury. The aim of the paper is to establish a CDCD porcine model which is closely mimicking human CDCD scenario, and investigate the physiologic changes from withdrawal of ventilatory support to circulatory arrest. Ten domestic crossbred pigs were anesthetized and ventilated with room air. Once all baseline data was taken, atracurium besilate (0.9 mg/kg, 3×ED95) was administered and the ventilator was discontinued while the animal was under deep anesthesia to establish the porcine CDCD model. Meanwhile, heparin (150~200 U/kg) was administered after discontinuation of the ventilator. The time to death and the changes of arterial blood gases and hemodynamic parameters were monitored every 5 minutes until circulatory arrest. In addition, histopathology, ultrastructures (via electron microscope) and expression of tight junction proteins of intestinal mucosa were observed at the baseline and the time of death. The mean time to death was approximately (21.8±3.12 min. Within 5 minutes of removal of the ventilator, there was a hyperdynamic period. Systolic blood pressure and heart rate quickly increased to 118.5±10.4 mmHg and 108.2±4.94 bpm, respectively. Blood pressure and heart rate then reduced rapidly until circulatory arrest. Moreover, the PaO2 quickly dropped to 17.4±3.13 mmHg, the blood gases throughout the apneic time showed a rapid hypercapnia and acidosis. In addition, warm ischemia damaged intestinal mucosa and reduced TJ proteins expression. A new swine CDCD model, simulating three stages of "withdrawal of ventilation, systemic anticoagulation and determination of death", which closely mimics the human DCD scenario and can thus be used in studies related to organ transplantation, was successfully established.

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

    Energy Technology Data Exchange (ETDEWEB)

    Ortensi, J.; Brian Boer; Abderrafi M. Ougouag

    2010-10-01

    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 core

  19. Analysis of Control Rod Drop Accidents for the Canadian SCWR Using Coupled 3-Dimensional Neutron Kinetics and Thermal Hydraulics

    Directory of Open Access Journals (Sweden)

    Frederic Salaun

    2018-01-01

    Full Text Available The Canadian Supercritical Water-cooled Reactor (SCWR, a GEN IV reactor design, is a hybrid design of the well-established CANDU™ and Boiling Water Reactor with water above its thermodynamic critical point. Given the batch fueled design, control rods are used to manage the reactivity throughout the fuel cycle. This paper examines the consequences of a control rod drop accident (CRDA for the Canadian SCWR. The asymmetry generated by the dropped rod requires an accurate 3-dimensional neutron kinetics calculation coupled to a detailed thermal-hydraulic model. Before simulating the CRDAs, the proper implementation of the 3D reactivity feedback was verified and various sensitivity studies were performed. This work demonstrates that the proposed safety systems for the SCWR core are capable of terminating the CRDA sequence prior to exceeding maximum sheath and centerline temperatures. In one instance involving a rod on the periphery of the core, the proposed trip setpoint (115% FP was not exceeded and a new steady state was reached. Therefore it is recommended that the design also include provisions for a high-log rate and/or local Neutron Overpower Protection (NOP trips, similar to existing CANDU designs such that reactor shutdown can be assured for such spatial anomalies.

  20. WWER safety analysis in case of simultaneous positive reactivity introduction by control rods withdrawal and pure condensate injection

    International Nuclear Information System (INIS)

    Kuchin, A.; Ovdiienko, I.; Khalimonchuk, V.

    2011-01-01

    Results of calculation assessment of positive reactivity insertion rate are presented under condition of simultaneous effect on reactivity of two provided by WWER project reactivity operation systems - extract of regulation group and decrease of boron acid concentration. Showed that maximal positive reactivity insertion rate is achieved at hot zero power and it's too less than limit value 0.07β eff /sec defined by normative document 'Nuclear safety regulation of NPP with pressurized water reactors'. For given reasons concluded that necessity of application of the extended protection PZ-2 actuation time is absent on interval of pure condensate transportation for full exclusion of possibility of positive reactivity insertion simultaneously by two different operation systems. (Authors)

  1. Melatonin for sedative withdrawal in older patients with primary insomnia: a randomized double-blind placebo-controlled trial

    Science.gov (United States)

    Lähteenmäki, Ritva; Puustinen, Juha; Vahlberg, Tero; Lyles, Alan; Neuvonen, Pertti J; Partinen, Markku; Räihä, Ismo; Kivelä, Sirkka-Liisa

    2014-01-01

    Aim We compared the efficacy of melatonin and placebo as adjuvants in the withdrawal of patients from long term temazepam, zopiclone or zolpidem (here ‘BZD’) use. Methods A double-blind, placebo-controlled, randomized trial was conducted in a primary health care outpatient clinic. Ninety-two men or women (≥55 years) with primary insomnia and chronic BZD use received controlled release melatonin 2 mg (CRM) (n = 46) or placebo (n = 46) during the 1 month withdrawal from BZDs. Psychosocial support was provided. Follow-up continued for up to 6 months. Successful BZD withdrawal by the end of 1 month was confirmed by BZD plasma determinations, while reduction in BZD use and abstinence continuing for 6 months were noted. Results There were two drop-outs on CRM and one on placebo. After a 1 month withdrawal, 31 participants (67%; 95% CI 54, 81) on CRM and 39 (85%; 74, 95) on placebo had withdrawn completely (intention-to-treat analysis between groups, P = 0.051; per protocol P = 0.043). Reduction in BZD use was similar or even more rare in the CRM than in the placebo group (P = 0.052 per protocol). After 6 months, 14 participants in the CRM group and 20 in the placebo group remained non-users of BZD (NS between groups). BZD doses were higher in the CRM than in the placebo group at the end of the 6 month follow-up (P = 0.025). Withdrawal symptoms did not differ between the groups. Conclusions Gradual dose reduction of BZDs combined with CRM or placebo, and psychosocial support produced high short term and moderate long term BZD abstinence. CRM showed no withdrawal benefit compared with placebo. PMID:24286360

  2. Research on the electromagnetic structure of movable coil electromagnet drive mechanism for reactor control rod

    International Nuclear Information System (INIS)

    Zhang Jige; Yian Huijie; Wu Yuanqiang; Wu Xinxin; Yu Suyuan; He Shuyan

    2007-01-01

    The movable coil electromagnet drive mechanism (MCEDM) is a new drive scheme for the reactor control rod, and it has a simple structure, good security and reliability property, etc. MCEDM with an air cooled structure has been used in the land research reactor. In order to apply MCEDM to the mobile reactor, experimental and theoretical study on the electromagnet with an oil-water cooled structure and a single magnetic flux circuit (called the type A electro-magnet) has been completed. It is proven by the experiment and theory that the oil-water cooled structure is an excellent measure to increase the coil current of MCEDM. Moreover, a type B electromagnet with an oil-water cooled structure and double magnetic flux circuits is designed to further increase the magnetic force of MCEDM. The analysis of finite element method shows that the type B electromagnet could double the saturation current of type A electro-magnet and the magnetic force of type B electromagnet is greater than that of the type A electromagnet. Moreover, it is proven that the dynamic property of type B electromagnet is better than type A electromagnet. (author)

  3. Sensitivity analysis of an Advanced Gas-cooled Reactor control rod model

    International Nuclear Information System (INIS)

    Scott, M.; Green, P.L.; O’Driscoll, D.; Worden, K.; Sims, N.D.

    2016-01-01

    Highlights: • A model was made of the AGR control rod mechanism. • The aim was to better understand the performance when shutting down the reactor. • The model showed good agreement with test data. • Sensitivity analysis was carried out. • The results demonstrated the robustness of the system. - Abstract: A model has been made of the primary shutdown system of an Advanced Gas-cooled Reactor nuclear power station. The aim of this paper is to explore the use of sensitivity analysis techniques on this model. The two motivations for performing sensitivity analysis are to quantify how much individual uncertain parameters are responsible for the model output uncertainty, and to make predictions about what could happen if one or several parameters were to change. Global sensitivity analysis techniques were used based on Gaussian process emulation; the software package GEM-SA was used to calculate the main effects, the main effect index and the total sensitivity index for each parameter and these were compared to local sensitivity analysis results. The results suggest that the system performance is resistant to adverse changes in several parameters at once.

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    Czajkowski, C.J.; Schuster, M.H.; Roberts, T.C.; Milian, L.W.

    1989-08-01

    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 (K I ) specimens were tested on an unirradiated compositionally and dimensionally similar (to HFBR follower) 6061 T-6 material with K max 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

  6. Experimental and numerical study of the MYRRHA control rod system dynamics

    International Nuclear Information System (INIS)

    Kennedy, G.; Lamberts, D.; Van Tichelen, K.; Profir, M.; Moreau, V.

    2017-01-01

    This paper presents an experimental and numerical investigation of the buoyancy driven MYRRHA control rod (CR) insertion during an emergency SCRAM. The study aimed to support the MYRRHA reactor design and characterise the hydrodynamic behaviour of the CR system while demonstrating the proof-of-principle. A full-scale mock-up test section of the MYRRHA CR was constructed to test the hydrodynamics in Lead Bismuth Eutectic over a wide range of operating conditions, to provide experimental data for the qualification of the CR system. A numerical CFD model of the CR test section was also setup in STAR-CCM+. The simulations make use of the recently developed overset mesh method to simulate the dynamic two-way coupling between the moving CR bundle and the fluid domain. The numerical methodology and post-test simulation results are validated against the experimental results. The steady state hydraulic results and the transient insertion results from both the experimental and numerical efforts are presented. The influence of the global process conditions on the CR insertion time are presented as well. This investigation successfully demonstrates the CR insertion proof-of-principle during a SCRAM. (author)

  7. Piecewise nonlinear dynamic characteristics study of the control rod drive mechanism

    International Nuclear Information System (INIS)

    Shen Xiaoyao; Wang Feng

    2011-01-01

    Piecewise nonlinear dynamics of the control rod mechanism (CRDM), one of the critical components in PWR nuclear power plants, are studied for its lifting process in this paper. Firstly, equations of the electric circuit and the magnetic circuit are set up. Then based on the dynamic lifting process analysis of CRDM, its motion procedure is divided into three stages, and the coupled magnetic-electric-mechanical equation for each stage is derived. By combining the analytical solution method and the numerical simulation method, the piecewise nonlinear governing equations are solved. Finally, parameters which can illustrate the dynamic characteristics of CRDM, such as the magnetic force, the coil current, the armature displacement, the armature velocity and the acceleration are obtained and corresponding curves with the time are drawn and analyzed. The analysis results are confirmed by the test which proves the validity of our method. Work in this paper can be used for design and analysis as well as the site fault diagnosis of CRDM. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Cancino P, G.

    2016-10-01

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

  9. Sick leave for follow-up control in thyroid cancer patients: comparison between stimulation with Thyrogen and thyroid hormone withdrawal.

    Science.gov (United States)

    Borget, I; Corone, C; Nocaudie, M; Allyn, M; Iacobelli, S; Schlumberger, M; De Pouvourville, G

    2007-05-01

    The clinical benefits of recombinant human thyroid-stimulating hormone (rhTSH; Thyrogen) are well established as an alternative stimulation procedure to thyroid hormone withdrawal in the diagnostic follow-up of thyroid cancer patients. By avoiding periods of hypothyroidism, patients do not suffer from a decreased quality of life and keep their ability to work. This study compared the frequency, the duration and the cost of sick leave for follow-up control between rhTSH and withdrawal. The study population consisted of patients with thyroid carcinoma first treated by thyroidectomy and radioiodine ablation. Patients were recruited at their control visit between October 2004 and May 2006 in three hospitals, both prospectively and retrospectively. Collection data consisted of patient information, job characteristics and duration of sick leave during the month before and the month after control. The valuation of sick leave used the friction cost method. Among the 306 patients included, 292 (95%) completed the entire questionnaire. The mean age was 46.7 years. Among the 194 active patients, patients treated with rhTSH, when compared with patients treated by withdrawal, were less likely to require sick leave (11 vs 33%; P=0.001). The mean duration of sick leave was shorter (3.1 vs 11.2 days; P=0.002) and indirect costs due to absenteeism accounted for 454 Euro +/- 1673 vs 1537 Euro +/- 2899 for withdrawal stimulation. For active patients, rhTSH treatment reduced the length and the cost of sick leave by 8.1 days and 1083 Euro per control respectively, when compared with withdrawal treatment.

  10. Development of Mathematical Model and Analysis Code for Estimating Drop Behavior of the Control Rod Assembly in the Sodium Cooled Fast Reactor

    International Nuclear Information System (INIS)

    Oh, Se-Hong; Kang, SeungHoon; Choi, Choengryul; Yoon, Kyung Ho; Cheon, Jin Sik

    2016-01-01

    On receiving the scram signal, the control rod assemblies are released to fall into the reactor core by its weight. Thus drop time and falling velocity of the control rod assembly must be estimated for the safety evaluation. There are three typical ways to estimate the drop behavior of the control rod assembly in scram action: Experimental, numerical and theoretical methods. But experimental and numerical(CFD) method require a lot of cost and time. Thus, these methods are difficult to apply to the initial design process. In this study, mathematical model and theoretical analysis code have been developed in order to estimate drop behavior of the control rod assembly to provide the underlying data for the design optimization. Mathematical model and theoretical analysis code have been developed in order to estimate drop behavior of the control rod assembly to provide the underlying data for the design optimization. A simplified control rod assembly model is considered to minimize the uncertainty in the development process. And the hydraulic circuit analysis technique is adopted to evaluate the internal/external flow distribution of the control rod assembly. Finally, the theoretical analysis code(named as HEXCON) has been developed based on the mathematical model. To verify the reliability of the developed code, CFD analysis has been conducted. And a calculation using the developed analysis code was carried out under the same condition, and both results were compared

  11. Controlling chaos based on a novel intelligent integral terminal sliding mode control in a rod-type plasma torch

    Science.gov (United States)

    Safa, Khari; Zahra, Rahmani; Behrooz, Rezaie

    2016-05-01

    An integral terminal sliding mode controller is proposed in order to control chaos in a rod-type plasma torch system. In this method, a new sliding surface is defined based on a combination of the conventional sliding surface in terminal sliding mode control and a nonlinear function of the integral of the system states. It is assumed that the dynamics of a chaotic system are unknown and also the system is exposed to disturbance and unstructured uncertainty. To achieve a chattering-free and high-speed response for such an unknown system, an adaptive neuro-fuzzy inference system is utilized in the next step to approximate the unknown part of the nonlinear dynamics. Then, the proposed integral terminal sliding mode controller stabilizes the approximated system based on Lyapunov’s stability theory. In addition, a Bee algorithm is used to select the coefficients of integral terminal sliding mode controller to improve the performance of the proposed method. Simulation results demonstrate the improvement in the response speed, chattering rejection, transient response, and robustness against uncertainties.

  12. Validation the methodology calculate critical position of control rods to the critical facility IPEN/MB-01

    International Nuclear Information System (INIS)

    Lopez Aldama, D.; Rodriguez Gual, R.

    1998-01-01

    Presently work intends to validate the models and programs used in the Nuclear Technology Center for calculating the critical position of control rods by means of the analysis of the measurements performed at the critical facility IPEN/MB-01. The lattice calculations were carried out with the WIMS/D4 code and for the global calculations the diffusion code SNAP-3D was used

  13. NDE of Possible Service-Induced PWSCC in Control Rod Drive Mechanism Housings Removed from Service

    International Nuclear Information System (INIS)

    Cumblidge, Stephen E.; Doctor, Steven R.; Schuster, George J.; Harris, Robert V.; Crawford, Susan L.

    2006-01-01

    Studies being conducted at the Pacific Northwest National Laboratory (PNNL) in Richland, Washington are being performed to assess the effectiveness of nondestructive examination (NDE) techniques on removed-from-service control rod drive mechanism (CRDM) nozzles and the associated J-groove attachment welds. This work is being performed to provide information to the United States Nuclear Regulatory Commission (US NRC) on the effectiveness of NDE techniques such as ultrasonic testing (UT), eddy current testing (ET), and visual testing (VT) 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. The basic NDE measurements follow standard industry techniques for conducting in-service inspections of CRDM nozzles and the crown of the J-groove welds and buttering. In addition, laboratory-based NDE methods were employed to conduct inspections of the CRDM assemblies, with particular emphasis on the J-groove weld and buttering. This paper describes the NDE measurements that were employed on the two CRDMs to detect and characterize the indications and the analysis of these indications. The two CRDM assemblies were removed from service from the North Anna 2 vessel head, including the CRDM nozzle, the J-groove weld, buttering, and a portion of the ferritic head material. One nozzle contains suspected PWSCC, based on in-service inspection data; the second contains evidence suggesting through-wall leakage, although this was unconfirmed. A destructive test plan is being developed to directly characterize the indications found using nondestructive testing. The results of this destructive testing will be included when the destructive testing is completed.

  14. Varenicline for opioid withdrawal in patients with chronic pain: a randomized, single-blinded, placebo controlled pilot trial.

    Science.gov (United States)

    Hooten, W Michael; Warner, David O

    2015-03-01

    The objectives of this randomized, single-blinded, placebo-controlled pilot trial were to investigate the effects of varenicline on opioid withdrawal among chronic pain patients undergoing opioid detoxification in an interdisciplinary pain program and the feasibility of varenicline use in this population. Twenty-one patients were recruited (varenicline=10, placebo=11), and 7 patients in the varenicline and 11 in the placebo group completed the study. Opioid withdrawal was quantified using the Clinical Opiate Withdrawal Scale, and varenicline-related adverse effects were assessed. Opioid withdrawal scores tended to decrease over the course of opioid tapering in those receiving varenicline and increase in those receiving placebo. Varenicline was well-tolerated in this population, with no adverse drug effects (including nausea) observed and no effect on improvements in pain severity and depression. This randomized pilot study provides preliminary data for future trials of varenicline in opioid-dependent adults with chronic pain undergoing medically directed opioid detoxification. Copyright © 2014 Elsevier Ltd. All rights reserved.

  15. Progress on B4C control rod modeling in RELAP/SCDAPSIM with application to quench and Phebus

    International Nuclear Information System (INIS)

    Kawahara, Keisuke; Hohorst, Judith K.; Allison, Chris M.

    2014-01-01

    The RELAP/SCDAPSIM code is designed to predict the behavior of reactor systems during normal and accident conditions. RELAP/SCDAPSIM/MOD3.5 is an experimental version of the code with the most advanced fuel and severe accident behavior models and correlations. It includes modeling improvements that were specifically added to support (a) the ongoing experimental severe accident programs in Europe and Japan and (b) the analysis and assessment activities related to the accident at the Fukushima Daiichi NPS. One of the improved models describes the behavior of cylindrical B 4 C control rods used in selected PWR designs and in integral experiments used to assess the heating and melting of PWR, BWR, and VVER assemblies. It replaces an older model that was originally developed by the US Nuclear Regulatory Commission in the mid- 1980's. It includes a combination of new and improved models and correlations to more accurately describe (a) eutectic reactions between Zircaloy, B 4 C, and stainless steel, (b) oxidation for B 4 C, Zircaloy, and stainless steel, and (c) the effects of the gap between the Zircaloy guide tube and the stainless steel sheath surrounding B 4 C pellets used in many control rod designs. This paper will discuss the development of the new model and validation of the model using the PHEBUS B 4 C test, FPT-3, and the KIT quench experiments with a central B 4 C control rod. (authors)

  16. Postirradiation examination of JOYO MK-II control rod (CRM601). Irradiation performance of shroud type absorber pin

    International Nuclear Information System (INIS)

    Tanaka, Kosuke; Kikuchi, Shin; Katsuyama, Kozo; Nagamine, Tsuyoshi; Mitsugi, Takeshi; Uto, Manabu; Tatebe, Kazuaki; Onose, Shoji; Maruyama, Tadashi

    1998-10-01

    This paper describes the results of postirradiation examination and analysis by CORAL code for irradiation performance of CRM601 control rod, which was the 6th reloaded control rod with shroud type absorber pins for use in JOYO MK-II core. The detailed visual examination indicated that there was no cladding breach in absorber pins. However, sodium ingress from the vent tube was observed in four absorber pins among seven pins. While a remarkable oval deformation occurred in cladding tube of helium bonded absorber pins, a little or no diametral change was observed in the absorber pins in which sodium ingress took place. From metallurgical observations and the analysis by CORAL code, it was estimated that the shroud tube installed in helium bonded absorber pins were irradiated at 720degC, and those in sodium bonded absorber pins were irradiated at 420degC. It was confirmed that diametral change of cladding depended on the initial gap between shroud and cladding tube. The results of present investigation indicate that it is desirable to use the materials with low thermal expansion coefficient for shroud tubes, and that sodium bonded absorber pins were advantageous for obtaining long life control rods. (author)

  17. Maximum/minimum asymmetric rod detection

    International Nuclear Information System (INIS)

    Huston, J.T.

    1990-01-01

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

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

    International Nuclear Information System (INIS)

    Boehm, H.H.; Foerch, H.

    1984-01-01

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

  19. Amantadine as Augmentation in Managing Opioid Withdrawal with Clonidine: a randomized controlled trial.

    Directory of Open Access Journals (Sweden)

    Shahrokh Amiri

    2014-09-01

    Full Text Available Withdrawal symptoms are a main reason of continuous use of opioid. This study compares the efficacy of augmentation of amantadine with clonidine in decreasing opioid withdrawal symptoms.This double-blind randomized clinical trial was carried out in the detoxification and rehabilitation inpatient ward at Razi Hospital, Tabriz, Iran during 2012. The patients were randomly assigned to receive clonidine or clonidine plus amantadine; and withdrawal symptoms were evaluated in the admission day and 24, 48, and 72 hours later. Data were analyzed using SPSS by the 2*2 repeated analyses of variances (ANOVA.From the total of 69 participants, 30 patients completed the trial in each group. The severity of symptoms, however, had an increasing trend in both groups. Analysis of variance of the symptom severity score (by The Clinical Opiate Withdrawal Scale revealed a significant group-time interaction, and the patients who were receiving amantadine experienced milder symptoms.Treatment of opioid withdrawal symptoms with amantadine and clonidine would result in a better outcome compared with clonidine alone.

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

    International Nuclear Information System (INIS)

    Rosa, R.; Palomba, M.; Sepielli, M.

    2008-01-01

    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

  1. 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)

    1995-09-01

    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

  2. Study on Reactor Power Transient Characteristics (Reactor Training Experiments) –Control Rod Reactivity Calibration by Positive Period Method and other Experiment–

    OpenAIRE

    尾崎, 禎彦; Ozaki, Yoshihiko; 砂川, 武義; Sunagawa, Takeyoshi

    2014-01-01

    In this paper, it is reported about some experiments that have been carried out in the reactor training that targets sophomore of the department of applied nuclear engineering, FUT. Reactor of Kinki University Atomic Energy Research Institute (UTR-KINKI) was used for reactor training. When each critical state was achieved at different reactor output respectively in reactor operating, it was confirmed that the control rod position at that time does not change. Further, control rod reactivity c...

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

    Energy Technology Data Exchange (ETDEWEB)

    Mankamo, T. [Avaplan Oy, Espoo (Finland)

    1996-12-01

    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.

  4. Improved point-kinetics model for the BWR control rod drop accident

    International Nuclear Information System (INIS)

    Neogy, P.; Wakabayashi, T.; Carew, J.F.

    1985-01-01

    A simple prescription to account for spatial feedback weighting effects in RDA (rod drop accident) point-kinetics analyses has been derived and tested. The point-kinetics feedback model is linear in the core peaking factor, F/sub Q/, and in the core average void fraction and fuel temperature. Comparison with detailed spatial kinetics analyses indicates that the improved point-kinetics model provides an accurate description of the BWR RDA

  5. Rod drive and latching mechanism

    International Nuclear Information System (INIS)

    Veronesi, L.; Sherwood, D.G.

    1982-01-01

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

  6. 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)

    Cebula, Artur; Taler, Jan

    2014-01-01

    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

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

    International Nuclear Information System (INIS)

    Ekberg, K.

    1980-03-01

    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)

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

    International Nuclear Information System (INIS)

    Kaoutar, M.

    1986-09-01

    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 [fr

  9. Withdrawal Notice.

    Science.gov (United States)

    2018-01-01

    Zhi-Feng Xiang & Li-Xin Deng (2014): Complete mitochondrial genome of white Bengal tiger, Panthera tigris tigris, DOI: 10.3109/19401736.2013.873912. This article, published online ahead of print on 17 January 2014, has been withdrawn by the authors due to an error identified in the data. This updated withdrawal notice replaces the original withdrawal notice, which still included the published article in full.

  10. SIMS imaging analyses of in-reactor irradiated boron carbide control rod samples

    Science.gov (United States)

    Gebhardt, O.; Gavillet, D.

    2000-06-01

    The relative radial burnup of the isotope 10B in an absorber rod has been determined by cross-sectional imaging analysis using a shielded secondary ion mass-spectrometer (SIMS). Two-dimensional secondary ion images of the specimen cross-section were recorded. The count rates of the burnable isotope 10B were normalised with the count rates of the non-burnable isotope 11B. The relative radial burnup of the isotope 10B was calculated and graphically plotted in the form of 2D-images and vector diagrams.

  11. Using nonlinearity and spatiotemporal property modulation to control effective structural properties: dynamic rods

    DEFF Research Database (Denmark)

    Thomsen, Jon Juel; Blekhman, Iliya I.

    2007-01-01

    What are the effective properties of a generally nonlinear material or structure, whose local properties are modulated in both space and time? It has been suggested to use spatiotemporal modulation of structural properties to create materials and structures with adjustable effective properties......, and exemplified. Then simple approximate analytical expressions are derived for the effective wave speed and natural frequencies for one-dimensional wave propagation in a nonlinear elastic rod, where the spatiotemporal modulation is imposed as a high-frequency standing wave, supposed to be given. Finally the more...

  12. Development of external coupling for calculation of the control rod worth in terms of burn-up for a WWER-1000 nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Noori-Kalkhoran, Omid, E-mail: o_noori@yahoo.com [Reactor Research School, Nuclear Science and Technology Research Institute (NSTRI), Tehran (Iran, Islamic Republic of); Yarizadeh-Beneh, Mehdi [Faculty of Engineering, Shahid Beheshti University, Tehran (Iran, Islamic Republic of); Ahangari, Rohollah [Reactor Research School, Nuclear Science and Technology Research Institute (NSTRI), Tehran (Iran, Islamic Republic of)

    2016-08-15

    Highlights: • Calculation of control rod worth in term of burn-up. • Calculation of differential and integral control rod worth. • Developing an external couple. • Modification of thermal-hydraulic profiles in calculations. - Abstract: One of the main problems relating to operation of a nuclear reactor is its safety and controlling system. The most widely used control systems for thermal reactors are neutron absorbent rods. In this study a code based method has been developed for calculation of integral and differential control rod worth in terms of burn-up for a WWER-1000 nuclear reactor. External coupling of WIMSD-5B, PARCS V2.7 and COBRA-EN has been used for this purpose. WIMSD-5B has been used for cell calculation and handling burn-up of the core in various days. PARCS V2.7 has been used for neutronic calculation of core and critical boron concentration search. Thermal-hydraulic calculation has been performed by COBRA-EN. An external coupling algorithm has been developed by MATLAB to couple and transfer suitable data between these codes in each step. Steady-State Power Picking Factors (PPFs) of the core and control rod worth for different control rod groups have been calculated from Beginning Of Cycle (BOC) to 289.7 Effective Full Power Days (EFPDs) in some steps. Results have been compared with the results of Bushehr Nuclear Power Plant (BNPP) Final Safety Analysis Report (FSAR). The results show a good agreement and confirm the ability of developed coupling in calculation of control rod worth in terms of burn-up.

  13. 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)

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  15. RAPID-L Highly Automated Fast Reactor Concept Without Any Control Rods (1) Reactor concept and plant dynamics analyses

    International Nuclear Information System (INIS)

    Kambe, Mitsuru; Tsunoda, Hirokazu; Mishima, Kaichiro; Iwamura, Takamichi

    2002-01-01

    The 200 kWe uranium-nitride fueled lithium cooled fast reactor concept 'RAPID-L' to achieve highly automated reactor operation has been demonstrated. RAPID-L is designed for Lunar base power system. It is one of the variants of RAPID (Refueling by All Pins Integrated Design), fast reactor concept, which enable quick and simplified refueling. The essential feature of RAPID concept is that the reactor core consists of an integrated fuel assembly instead of conventional fuel subassemblies. In this small size reactor core, 2700 fuel pins are integrated altogether and encased in a fuel cartridge. Refueling is conducted by replacing a fuel cartridge. The reactor can be operated without refueling for up to 10 years. Unique challenges in reactivity control systems design have been attempted in RAPID-L concept. The reactor has no control rod, but involves the following innovative reactivity control systems: Lithium Expansion Modules (LEM) for inherent reactivity feedback, Lithium Injection Modules (LIM) for inherent ultimate shutdown, and Lithium Release Modules (LRM) for automated reactor startup. All these systems adopt lithium-6 as a liquid poison instead of B 4 C rods. In combination with LEMs, LIMs and LRMs, RAPID-L can be operated without operator. This is the first reactor concept ever established in the world. This reactor concept is also applicable to the terrestrial fast reactors. In this paper, RAPID-L reactor concept and its transient characteristics are presented. (authors)

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

    Directory of Open Access Journals (Sweden)

    Falko Böttger-Hiller

    2013-01-01

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

  17. Fuel rods

    International Nuclear Information System (INIS)

    Hattori, Shinji; Kajiwara, Koichi.

    1980-01-01

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

  18. 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)

    Cieslav, C.; Peteuil, M.

    1985-01-01

    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

  19. Control rod drop transient analysis with the coupled parallel code pCTF-PARCSv2.7

    International Nuclear Information System (INIS)

    Ramos, Enrique; Roman, Jose E.; Abarca, Agustín; Miró, Rafael; Bermejo, Juan A.

    2016-01-01

    Highlights: • An MPI parallel version of the thermal–hydraulic subchannel code COBRA-TF has been developed. • The parallel code has been coupled to the 3D neutron diffusion code PARCSv2.7. • The new codes are validated with a control rod drop transient. - Abstract: In order to reduce the response time when simulating large reactors in detail, a parallel version of the thermal–hydraulic subchannel code COBRA-TF (CTF) has been developed using the standard Message Passing Interface (MPI). The parallelization is oriented to reactor cells, so it is best suited for models consisting of many cells. The generation of the Jacobian matrix is parallelized, in such a way that each processor is in charge of generating the data associated with a subset of cells. Also, the solution of the linear system of equations is done in parallel, using the PETSc toolkit. With the goal of creating a powerful tool to simulate the reactor core behavior during asymmetrical transients, the 3D neutron diffusion code PARCSv2.7 (PARCS) has been coupled with the parallel version of CTF (pCTF) using the Parallel Virtual Machine (PVM) technology. In order to validate the correctness of the parallel coupled code, a control rod drop transient has been simulated comparing the results with the real experimental measures acquired during an NPP real test.

  20. Treatment of acute alcohol withdrawal with gabapentin: results from a controlled two-center trial.

    Science.gov (United States)

    Bonnet, Udo; Banger, Markus; Leweke, F Markus; Specka, Michael; Müller, Bernhard W; Hashemi, Thilo; Nyhuis, Peter W; Kutscher, Sven; Burtscheidt, Wilhelm; Gastpar, Markus

    2003-10-01

    A few case reports and data from animal experiments point to a possible efficacy of gabapentin (GP) in the treatment of alcohol withdrawal syndrome (AWS). Because of ethical considerations, the efficacy of GP in acute AWS was tested in an add-on fashion to clomethiazole (CLO). Given that the symptom-triggered amount of CLO required to limit AWS within the first 24 hours is related to the severity of AWS, we tested this amount of CLO during placebo (P) or GP (400 mg qid) under double blind, randomized conditions. Sixty-one patients (P = 29/GP = 32) suffering from alcohol dependence (ICD-10) and without any other psychiatric condition or psychotropic medication were included. The groups were not significantly different in baseline characteristics (eg, demographic data, severity of AWS). Both ITT and completer analyses revealed no significant differences between the groups considering the primary effectiveness measure: amount of CLO required in the first 24 hours (P = 6.1 +/- 5.4/GP = 6.2 +/- 4.7 capsules). In addition, premature discontinuations (P = 3/GP = 2) and decreases in Mainz Alcohol Withdrawal Scores were not significantly different in the first 48 hours of AWS (secondary effectiveness measures). Tolerability of combined CLO/GP was studied throughout the whole treatment comprising a 5-day lasting reduction part subsequent to the first 48 hours. Throughout the whole 7-day treatment a total of 5 and 2 patients dropped out and 6 and 5 patients reported adverse clinical events in the P and GP groups, respectively. All together, GP (400 mg qid) was no better than P in saving initial consumption of CLO or decreasing initial Mainz Alcohol Withdrawal Scores suggesting that GP was ineffective in the management of acute AWS in this model. The combination of GP and CLO was safe.

  1. 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)

  2. 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)

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

    1996-01-01

    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)

  3. 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)

    Rahgoshay, M.; Rahmani, Y.

    2007-01-01

    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)

  4. Visual inspections of the neutron absorber control rods of the IEA-R1 reactor; Inspecoes visuais nas barras absorvedoras de neutrons do reator IEA-R1

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Jose Eduardo R. da; Terremoto, Luis A.A.; Castanheira, Myrthes; Zeituni, Carlos A.; Damy, Margaret de A. [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil). Centro de Engenharia Nuclear]. E-mail: jersilva@net.ipen.br

    2002-07-01

    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)

  5. Rod behaviour under base load, load follow and frequency control operation: CYRANO 2 code predictions versus experimental results

    International Nuclear Information System (INIS)

    Gautier, B.; Raybaud, A.

    1984-01-01

    The French PWR reactors are now currently operating under load follow and frequency control. In order to demonstrate that these operating conditions were not able to increase the fuel failure rate, fuel rod behaviour calculations have been performed by E.D.F. with CYRANO 2 code. In parallel with these theoretical calculations, code predictions have been compared to experimental results. The paper presents some of the comparisons performed on 17x17 fuel irradiated in FESSENHEIM 2 up to 30 GWd/tU under base load operation and in the CAP reactor under load follow and frequency control conditions. It is shown that experimental results can be predicted with a reasonable accuracy by CYRANO 2 code. The experimental work was carried out under joint R and D programs by EDF, FRAGEMA, CEA, and WESTINGHOUSE (CAP program by French partners only). (author)

  6. Calculation of neutron activation of control rods of a nuclear reactor, using MCNP5; Calculo de activacion neutronica de barras de control de un reactor nuclear, utilizando MCNP5

    Energy Technology Data Exchange (ETDEWEB)

    Pena V, J.D.

    2016-07-01

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

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

    International Nuclear Information System (INIS)

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

    2000-09-01

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

  8. 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.

  9. Alcohol withdrawal

    Science.gov (United States)

    ... so they can monitor you for symptoms of alcohol withdrawal. Prevention Reduce or avoid alcohol. If you have a drinking problem, you should ... team. 02-05-18: Editorial update. Alcoholism and Alcohol Abuse Read more ... HealthCare Commission (www.urac.org). URAC's accreditation program is an independent audit to verify that A. ...

  10. Analytical support for the B4C control rod test QUENCH-07

    International Nuclear Information System (INIS)

    Homann, C.; Hering, W.; Fernandez Benitez, J.A.; Ortega Bernardo, M.

    2003-04-01

    Degradation of B 4 C absorber rods during a beyond design accident in a nuclear power reactor may be a safety concern. Among others, the integral test QUENCH-07 is performed in the FZK QUENCH facility and supported by analytical work within the Euratom Fifth Framework Programme on Nuclear Fission Safety to get a more profound database. Since the test differed substantially from previous QUENCH tests, much more work had to be done for pretest calculations than usual to guarantee the safety of the facility and to derive the test protocol. Several institutions shared in this work with different computer code systems, as used for nuclear reactor safety analyses. Due to this effort, problems could be identified and solved, leading to several modifications of the originally planned test conduct, until a feasible test protocol could be derived and recommended. All calculations showed the same trends. Especially the high temperatures and hence the small safety margin for the facility were a concern. In this report, contributions of various authors, engaged in this work, are presented. The test QUENCH-07 and the related computational support by the engaged institutions were co-financed by the European Community under the Euratom Fifth Framework Programme on Nuclear Fission Safety 1998 - 2002 (COLOSS Project, contract No. FIKS-CT-1999-00002). (orig.)

  11. Analytical support for the B{sub 4}C control rod test QUENCH-07

    Energy Technology Data Exchange (ETDEWEB)

    Homann, C.; Hering, W. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Inst. fuer Reaktorsicherheit]|[Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Programm Nukleare Sicherheitsforschung; Birchley, J. [Paul Scherrer Inst. (Switzerland); Fernandez Benitez, J.A.; Ortega Bernardo, M. [Univ. Politecnica de Madrid (Spain)

    2003-04-01

    Degradation of B{sub 4}C absorber rods during a beyond design accident in a nuclear power reactor may be a safety concern. Among others, the integral test QUENCH-07 is performed in the FZK QUENCH facility and supported by analytical work within the Euratom Fifth Framework Programme on Nuclear Fission Safety to get a more profound database. Since the test differed substantially from previous QUENCH tests, much more work had to be done for pretest calculations than usual to guarantee the safety of the facility and to derive the test protocol. Several institutions shared in this work with different computer code systems, as used for nuclear reactor safety analyses. Due to this effort, problems could be identified and solved, leading to several modifications of the originally planned test conduct, until a feasible test protocol could be derived and recommended. All calculations showed the same trends. Especially the high temperatures and hence the small safety margin for the facility were a concern. In this report, contributions of various authors, engaged in this work, are presented. The test QUENCH-07 and the related computational support by the engaged institutions were co-financed by the European Community under the Euratom Fifth Framework Programme on Nuclear Fission Safety 1998 - 2002 (COLOSS Project, contract No. FIKS-CT-1999-00002). (orig.)

  12. Impact of mesh points number on the accuracy of deterministic calculations of control rods worth for Tehran research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Boustani, Ehsan [Nuclear Science and Technology Research Institute (NSTRI), Tehran (Iran, Islamic Republic of); Amirkabir University of Technology, Tehran (Iran, Islamic Republic of). Energy Engineering and Physics Dept.; Khakshournia, Samad [Amirkabir University of Technology, Tehran (Iran, Islamic Republic of). Energy Engineering and Physics Dept.

    2016-12-15

    In this paper two different computational approaches, a deterministic and a stochastic one, were used for calculation of the control rods worth of the Tehran research reactor. For the deterministic approach the MTRPC package composed of the WIMS code and diffusion code CITVAP was used, while for the stochastic one the Monte Carlo code MCNPX was applied. On comparing our results obtained by the Monte Carlo approach and those previously reported in the Safety Analysis Report (SAR) of Tehran research reactor produced by the deterministic approach large discrepancies were seen. To uncover the root cause of these discrepancies, some efforts were made and finally was discerned that the number of spatial mesh points in the deterministic approach was the critical cause of these discrepancies. Therefore, the mesh optimization was performed for different regions of the core such that the results of deterministic approach based on the optimized mesh points have a good agreement with those obtained by the Monte Carlo approach.

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

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

    DEFF Research Database (Denmark)

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

    2013-01-01

    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......-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...... near a mirror, not only allows an extraction of calibrated ensemble-averaged rates, but for the first time also to quantify the full inhomogeneous dispersion of radiative and non radiative decay rates across thousands of nanocrystals. We apply the technique to novel ultrastable CdSe/CdS dot...

  15. Heterostructure of Au nanocluster tipping on a ZnS quantum rod: controlled synthesis and novel luminescence.

    Science.gov (United States)

    Tian, Yang; Wang, Ligang; Yu, Shanshan; Zhou, Weiwei

    2015-08-14

    Heterostructures of metal nanoparticles and semiconductors are widely studied for their unique properties. However, few reports are available on the heterostructure of metal nanoclusters and semiconductors. In the present study, a heterostructure, in which gold nanoclusters selectively locate at ZnS quantum rod (QR) tips, was fabricated using a two-step solvothermal route. The composition, intrinsic crystallography, and junction of the prepared heterostructure were thoroughly investigated, and it was observed to exhibit novel luminescent behaviours. By comparison with the individual components of ZnS QRs and gold clusters, the resultant heterostructure shows an enhanced exciton emission and complete depression of defect emission for the ZnS component, and a pronounced red emission for the gold nanocluster component. The mechanism of these properties and the charge transfer between gold nanoclusters and ZnS QRs were also explored. The size and location of gold in the heterostructure were also controlled during synthesis to study their effects on the luminescence.

  16. Impact of mesh points number on the accuracy of deterministic calculations of control rods worth for Tehran research reactor

    International Nuclear Information System (INIS)

    Boustani, Ehsan; Amirkabir University of Technology, Tehran; Khakshournia, Samad

    2016-01-01

    In this paper two different computational approaches, a deterministic and a stochastic one, were used for calculation of the control rods worth of the Tehran research reactor. For the deterministic approach the MTRPC package composed of the WIMS code and diffusion code CITVAP was used, while for the stochastic one the Monte Carlo code MCNPX was applied. On comparing our results obtained by the Monte Carlo approach and those previously reported in the Safety Analysis Report (SAR) of Tehran research reactor produced by the deterministic approach large discrepancies were seen. To uncover the root cause of these discrepancies, some efforts were made and finally was discerned that the number of spatial mesh points in the deterministic approach was the critical cause of these discrepancies. Therefore, the mesh optimization was performed for different regions of the core such that the results of deterministic approach based on the optimized mesh points have a good agreement with those obtained by the Monte Carlo approach.

  17. Calculation of static characteristics of linear step motors for control rod drives of nuclear reactors - an approximate approach

    International Nuclear Information System (INIS)

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

    1993-01-01

    This paper describes an approximate method for calculating the static characteristics of linear step motors (LSM), being developed for control rod drives (CRD) in large nuclear reactors. The static characteristic of such an LSM which is given by the variation of electromagnetic force with armature displacement determines the motor performance in its standing and dynamic modes. The approximate method of calculation of these characteristics is based on the permeance analysis method applied to the phase magnetic circuit of LSM. This is a simple, fast and efficient analytical approach which gives satisfactory results for small stator currents and weak iron saturation, typical to the standing mode of operation of LSM. The method is validated by comparing theoretical results with experimental ones. (Author)

  18. Numerical simulation of fluid flow and forced convection heat transfer around a circular cylinder with control rods located in equilateral triangular arrangement

    Energy Technology Data Exchange (ETDEWEB)

    Harimi, Somayeh; Marjani, Azam [Dept. of Chemistry, Arak Branch, Islamic Azad University, Arak (Iran, Islamic Republic of); Moradi, Sadegh [Dept. of Chemical Engineering, Arak University, Arak (Iran, Islamic Republic of)

    2016-09-15

    Numerical study of forced convection heat transfer and fluid flow in laminar flow regime for a circular cylinder attached by three control rods is performed using the overset grid method. The aim of this work is evaluation of the control rods performance placed in equilateral triangular arrangements in suppressing vortex induced vibration of a primary cylinder. The influence of the dimensionless parameters including attach angle α, spacing ratio G/D, and Reynolds number on the hydrodynamic forces of the primary cylinder is also investigated. The unsteady flow at Reynolds number of 200 and Prandtl numbers of 0.7 and 7.0 is considered. In order to discretize the governing equations, a finite volume code based on the SIMPLEC algorithm is employed. Moreover, the local and mean Nusselt numbers are presented to illustrate the heat transfer characteristics of the primary cylinder and surrounding rods.

  19. Numerical simulation of fluid flow and forced convection heat transfer around a circular cylinder with control rods located in equilateral triangular arrangement

    International Nuclear Information System (INIS)

    Harimi, Somayeh; Marjani, Azam; Moradi, Sadegh

    2016-01-01

    Numerical study of forced convection heat transfer and fluid flow in laminar flow regime for a circular cylinder attached by three control rods is performed using the overset grid method. The aim of this work is evaluation of the control rods performance placed in equilateral triangular arrangements in suppressing vortex induced vibration of a primary cylinder. The influence of the dimensionless parameters including attach angle α, spacing ratio G/D, and Reynolds number on the hydrodynamic forces of the primary cylinder is also investigated. The unsteady flow at Reynolds number of 200 and Prandtl numbers of 0.7 and 7.0 is considered. In order to discretize the governing equations, a finite volume code based on the SIMPLEC algorithm is employed. Moreover, the local and mean Nusselt numbers are presented to illustrate the heat transfer characteristics of the primary cylinder and surrounding rods

  20. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-07-01

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

  2. Expert system for assisting the repair operations on the control racks of the control rods assembly in a 900 MW PWR type reactor

    International Nuclear Information System (INIS)

    Monnier, B.; Doutre, J.L.; Franco, A.

    1990-01-01

    The expert system presented was developed for assisting the repair operations on the control equipment of the control rod assembly in a PWR type reactor. The expert system allows the representation of expert knowledge and diagnostic reasoning. The objective of the expert system is to achieve the most precise diagnostic and localizing of the breakdown elements, by processing the data acquired during breakdown. The development steps, the structure and the applications of the expert system are summarized. The expert system operates in an IBM PC equipped with a AMAIA 8 Mo card. A time schedule of 18 months is predicted [fr

  3. Opiate and opioid withdrawal

    Science.gov (United States)

    ... medlineplus.gov/ency/article/000949.htm Opiate and opioid withdrawal To use the sharing features on this page, ... or withdrawing from opiates. Alternative Names Withdrawal from opioids; Dopesickness; Substance use - opiate withdrawal; Substance abuse - opiate withdrawal; Drug abuse - opiate withdrawal; ...

  4. A Phase 3 Placebo-Controlled, Double Blind, Multi-Site Trial of the alpha-2-adrenergic Agonist, Lofexidine, for Opioid Withdrawal

    Science.gov (United States)

    Yu, Elmer; Miotto, Karen; Akerele, Evaristo; Montgomery, Ann; Elkashef, Ahmed; Walsh, Robert; Montoya, Ivan; Fischman, Marian W.; Collins, Joseph; McSherry, Frances; Boardman, Kathy; Davies, David K.; O’Brien, Charles P.; Ling, Walter; Kleber, Herbert; Herman, Barbara H.

    2008-01-01

    Context Lofexidine is an alpha-2-A noradrenergic receptor agonist that is approved in the United Kingdom for the treatment of opioid withdrawal symptoms. Lofexidine has been reported to have more significant effects on decreasing opioid withdrawal symptoms with less hypotension than clonidine. Objective To demonstrate that lofexidine is well tolerated and effective in the alleviation of observationally-defined opioid withdrawal symptoms in opioid dependent individuals undergoing medically supervised opioid detoxification as compared to placebo. Design An inpatient, Phase 3, placebo-controlled, double blind, randomized multi-site trial with three phases: (1) Opioid Agonist Stabilization Phase (days 1–3), (2) Detoxification/Medication or Placebo Phase (days 4–8), and (3) Post Detoxification/Medication Phase (days 9–11). Subjects Sixty-eight opioid dependent subjects were enrolled at three sites with 35 randomized to lofexidine and 33 to placebo. Main Outcome Measure Modified Himmelsbach Opiate Withdrawal Scale (MHOWS) on study day 5 (2nd opioid detoxification treatment day). Results Due to significant findings, the study was terminated early. On the study day 5 MHOWS, subjects treated with lofexidine had significantly lower scores (equating to fewer/less severe withdrawal symptoms) than placebo subjects (Least squares means 19.5 ± 2.1 versus 30.9 ± 2.7; p=0.0019). Lofexidine subjects had significantly better retention in treatment than placebo subjects (38.2% versus 15.2%; Log rank test p=0.01). Conclusions Lofexidine is well tolerated and more efficacious than placebo for reducing opioid withdrawal symptoms in inpatients undergoing medically supervised opioid detoxification. Trial Registration trial registry name A Phase 3 Placebo-Controlled, Double-Blind Multi-Site Trial of Lofexidine for Opiate Withdrawal, registration number NCT00032942, URL for the registry http://clinicaltrials.gov/ct/show/NCT00032942?order=4. PMID:18508207

  5. Evaluation of the reduction of boron-10 in the control rods in the BWR of the Laguna Verde Central, through steady state calculations

    International Nuclear Information System (INIS)

    Montes T, J.L.; Perusquia, R.; Hernandez, J.L.; Ramirez S, J.R.

    2003-01-01

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

  6. Application of B{sub 4}C/Al{sub 2}O{sub 3} Burnable Absorber Rod to Control Excess Reactivity of SMR

    Energy Technology Data Exchange (ETDEWEB)

    Muth, Boravy; Hah, C. J. [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

    2016-10-15

    Soluble boron in a nuclear reactor coolant is one of the methods to control excess reactivity of the reactor. However, the use of soluble boron also causes some negative effects such as corrosion, more-positive tendency of Moderator Temperature Coefficient (MTC) and the requirement of Chemical Volume Control System (CVCS). One of the conceptual design features of SMR having been developed in Korea is soluble boron- free reactor to eliminate those drawbacks. Control rods and Burnable Absorber (BA) rods can be other methods than soluble to control excess reactivity. WABA (Wet Annular Burnable Absorber) and PYREX are such type. The other type is IFBA (Integral Fuel Burnable Absorber) in which fuel pellet surface is coated with BA. This paper compares nuclear characteristics of three types of BA as well as SLOBA in terms of k-infinite vs. burnup and explain design basis of SLOBA. This paper also presents the application of SLOBA rods to control long-term excess reactivity of SMR. The SMR loaded with SLOBA rods has been developed for the past few years in Korean. It is named as Bandi-50 with design features of 180 MWth, 37 FAs, fuel assembly height of 200 cm. Soluble-boron-free is one of nuclear design requirements of Bandi-50 and is achieved by controlling excess reactivity of the SMR using BAs and control rods only. To achieve this design requirement, LP is carefully determined in such way that CBC should be as low as possible. Fuel assembly cross-sections are generated by CASMO-3, and core depletion calculations are performed by MASTER.

  7. PIE data of lower guide tubes of control rod and MK-II fuel assemblies. Face to face distance data of spacer pad of lower guide tubes

    Energy Technology Data Exchange (ETDEWEB)

    Kikuchi, Shin [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1998-11-01

    This paper is arranged PIE data in order to provide data sheet for evaluation of core distortion. In this paper, there are two kinds of PIE data; one is face to face distance data of the spacer pad part of lower guide tubes of control rod. The other is the axial profile of fuel assemblies bowing. (author)

  8. Control rod calibration methods for fast breeder reactors applied to Phenix; Les methodes d'etalonnage des barres de commande des reacteurs a neutrons rapides application a Phenix

    Energy Technology Data Exchange (ETDEWEB)

    Lecourt, G

    1998-06-18

    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.)

  9. A phase III, randomized, multi-center, double blind, placebo controlled study of safety and efficacy of lofexidine for relief of symptoms in individuals undergoing inpatient opioid withdrawal.

    Science.gov (United States)

    Gorodetzky, Charles W; Walsh, Sharon L; Martin, Peter R; Saxon, Andrew J; Gullo, Kristen L; Biswas, Kousick

    2017-07-01

    Lofexidine is an alpha-2-adrenergic receptor agonist approved in the United Kingdom (UK) for the treatment of opioid withdrawal symptoms. Lofexidine has demonstrated better efficacy than placebo for reducing opioid withdrawal symptoms in patients undergoing opioid withdrawal with less reported hypotension than clonidine. Designed as an FDA registration trial, this 8-day, randomized, double-blind, placebo-controlled, parallel-group study in 264 patients dependent on short-acting opioids evaluated the efficacy of lofexidine hydrochloride in reducing withdrawal symptoms in patients undergoing opioid withdrawal. The primary efficacy measures were SOWS-Gossop on Day 3 and time-to-dropout. Secondary endpoints included the proportion of participants who were completers; area under the 5-day SOWS-Gossop - time curve (i.e., AUC 1-5 ), and daily mean SOWS-Gossop, OOWS-Handelsman, MCGI (subject and rater), and VAS-E scores. Participants received lofexidine HCl 3.2mg daily in four divided doses or matching placebo on Days 1-5, followed by 2days of placebo. Lofexidine significantly decreased mean Day 3 SOWS scores compared to placebo, 6.32 versus 8.67, respectively, p=0.0212. Fewer lofexidine patients were early terminators compared to placebo (59 versus 80, respectively); and non-completers in the lofexidine group remained in the study longer than those assigned to placebo (p=0.0034). Secondary endpoints consistently favored lofexidine. Lofexidine was well tolerated in this trial. Lofexidine significantly decreased SOWS scores compared to placebo and demonstrated better retention rates in participants undergoing opioid withdrawal. Lofexidine potentially offers a useful non-opioid alternative to treat opioid withdrawal symptoms. Copyright © 2017. Published by Elsevier B.V.

  10. A Proof-of-Concept Randomized Controlled Study of Gabapentin: Effects on Cannabis Use, Withdrawal and Executive Function Deficits in Cannabis-Dependent Adults

    Science.gov (United States)

    Mason, Barbara J; Crean, Rebecca; Goodell, Vivian; Light, John M; Quello, Susan; Shadan, Farhad; Buffkins, Kimberly; Kyle, Mark; Adusumalli, Murali; Begovic, Adnan; Rao, Santosh

    2012-01-01

    There are no FDA-approved pharmacotherapies for cannabis dependence. Cannabis is the most widely used illicit drug in the world, and patients seeking treatment for primary cannabis dependence represent 25% of all substance use admissions. We conducted a phase IIa proof-of-concept pilot study to examine the safety and efficacy of a calcium channel/GABA modulating drug, gabapentin, for the treatment of cannabis dependence. A 12-week, randomized, double-blind, placebo-controlled clinical trial was conducted in 50 unpaid treatment-seeking male and female outpatients, aged 18–65 years, diagnosed with current cannabis dependence. Subjects received either gabapentin (1200 mg/day) or matched placebo. Manual-guided, abstinence-oriented individual counseling was provided weekly to all participants. Cannabis use was measured by weekly urine toxicology and by self-report using the Timeline Followback Interview. Cannabis withdrawal symptoms were assessed using the Marijuana Withdrawal Checklist. Executive function was measured using subtests from the Delis–Kaplan Executive Function System. Relative to placebo, gabapentin significantly reduced cannabis use as measured both by urine toxicology (p=0.001) and by the Timeline Followback Interview (p=0.004), and significantly decreased withdrawal symptoms as measured by the Marijuana Withdrawal Checklist (pbrain stress systems that are dysregulated in drug dependence and withdrawal. PMID:22373942

  11. A proof-of-concept randomized controlled study of gabapentin: effects on cannabis use, withdrawal and executive function deficits in cannabis-dependent adults.

    Science.gov (United States)

    Mason, Barbara J; Crean, Rebecca; Goodell, Vivian; Light, John M; Quello, Susan; Shadan, Farhad; Buffkins, Kimberly; Kyle, Mark; Adusumalli, Murali; Begovic, Adnan; Rao, Santosh

    2012-06-01

    There are no FDA-approved pharmacotherapies for cannabis dependence. Cannabis is the most widely used illicit drug in the world, and patients seeking treatment for primary cannabis dependence represent 25% of all substance use admissions. We conducted a phase IIa proof-of-concept pilot study to examine the safety and efficacy of a calcium channel/GABA modulating drug, gabapentin, for the treatment of cannabis dependence. A 12-week, randomized, double-blind, placebo-controlled clinical trial was conducted in 50 unpaid treatment-seeking male and female outpatients, aged 18-65 years, diagnosed with current cannabis dependence. Subjects received either gabapentin (1200 mg/day) or matched placebo. Manual-guided, abstinence-oriented individual counseling was provided weekly to all participants. Cannabis use was measured by weekly urine toxicology and by self-report using the Timeline Followback Interview. Cannabis withdrawal symptoms were assessed using the Marijuana Withdrawal Checklist. Executive function was measured using subtests from the Delis-Kaplan Executive Function System. Relative to placebo, gabapentin significantly reduced cannabis use as measured both by urine toxicology (p=0.001) and by the Timeline Followback Interview (p=0.004), and significantly decreased withdrawal symptoms as measured by the Marijuana Withdrawal Checklist (pbrain stress systems that are dysregulated in drug dependence and withdrawal.

  12. A programme to reduce down-time and site radiation dose through improvements in control rod design

    International Nuclear Information System (INIS)

    Izatt, J.A.; Scobie, J.

    1983-01-01

    In many nuclear facilities a high proportion of the total radiation dose to staff may be ascribed to very few routine operations. Often this critical operation will require significant shut-down time, both to perform the function itself and to allow sufficient decay time to reduce radiation levels to acceptable values. This situation arises on the UTR-300 reactor at S.U.R.R.C., where over the twenty years of operation it has been found necessary to carry out preventive, and on some occasions remedial, maintenance at least once per year on the in-core elements of the control rod assemblies. This had led inevitably to reduced availability of the reactor and to significant overall radiation doses to the few long-serving staff members who carry out this work. The source of the problem seems to lie in the basic design of the assemblies, which may have been adequate for the original 10 KW versions of this reactor, but has proven to be deficient in several respects at the present 300 KW operating level. In an attempt to reduce down-time and radiation dose to staff a radical re-design of the control assemblies was undertaken in-house, and a conversion programme has been underway over the last two years. Details of the design improvements are presented with a discussion of the operating experience to date. (author)

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

    International Nuclear Information System (INIS)

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

    1988-08-01

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

  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)

    Reyes F, M. del C.

    2015-01-01

    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

  16. Fuel rods

    International Nuclear Information System (INIS)

    Fukushima, Kimichika.

    1984-01-01

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

  17. ELECTROMAGNETIC APPARATUS FOR MOVING A ROD

    Science.gov (United States)

    Young, J.N.

    1958-04-22

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

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

    International Nuclear Information System (INIS)

    Ortiz S, J.J.; Montes T, J.L.; Requena R, I.

    2003-01-01

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

  19. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-07-01

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

  1. Opiate exposure state controls dopamine D3 receptor and cdk5/calcineurin signaling in the basolateral amygdala during reward and withdrawal aversion memory formation.

    Science.gov (United States)

    Rosen, Laura G; Rushlow, Walter J; Laviolette, Steven R

    2017-10-03

    The dopamine (DA) D3 receptor (D3R) is highly expressed in the basolateral nucleus of the amygdala (BLA), a neural region critical for processing opiate-related reward and withdrawal aversion-related memories. Functionally, D3R transmission is linked to downstream Cdk5 and calcineurin signaling, both of which regulate D3R activity states and play critical roles in memory-related synaptic plasticity. Previous evidence links D3R transmission to opiate-related memory processing, however little is known regarding how chronic opiate exposure may alter D3R-dependent memory mechanisms. Using conditioned place preference (CPP) and withdrawal aversion (conditioned place aversion; CPA) procedures in rats, combined with molecular analyses of BLA protein expression, we examined the effects of chronic opiate exposure on the functional role of intra-BLA D3R transmission during the acquisition of opiate reward or withdrawal aversion memories. Remarkably, we report that the state of opiate exposure during behavioural conditioning (opiate-naïve/non-dependent vs. chronically exposed and in withdrawal) controlled the functional role of intra-BLA D3R transmission during the acquisition of both opiate reward memories and withdrawal-aversion associative memories. Thus, whereas intra-BLA D3R blockade had no effect on opiate reward memory formation in the non-dependent state, blockade of intra-BLA D3R transmission prevented the formation of opiate reward and withdrawal aversion memory in the chronically exposed state. This switch in the functional role of D3R transmission corresponded to significant increases in Cdk5 phosphorylation and total expression levels of calcineurin, and a corresponding decrease in intra-BLA D3R expression. Inhibition of either intra-BLA Cdk5 or calcineurin reversed these effects, switching intra-BLA associative memory formation back to a D3R-independent mechanism. Copyright © 2017 Elsevier Inc. All rights reserved.

  2. A prospective randomised control study: reduction of children's pain expectation using a picture book during blood withdrawal.

    Science.gov (United States)

    Zieger, B; Praskova, M; Busse, E; Barth, M

    2013-05-01

    Blood drawings are very painful and stressful for children. In a prospective control group study we investigated if using a picture book could reduce the children's pain expectation. In addition, the children's pain experience and the observed pain behaviour was monitored. Block-randomization were used and 120 children at the age of 6-12 years who were visiting the general pediatric and coagulation outpatient clinics were included in this study. Pain expectation and experience were assessed with the Face-Pain-Scale-Revised and the pain behavior with the Faces-Legs-Activity-Cry-Consolability Scale. Multivariate covariance analysis was used for data analysis. The results showed that with statistical controlling the influence of the primary pain expectation (baseline) the pain expectation before blood withdrawal was reduced significantly (p=0.001) and effectively (ES=0.56) using the picture book. Children who received no local anaesthesia reported that they felt less pain during blood drawing after reading the picture book. The few children with local anaesthesia reported no benefit from the picture book. The observed use of local anaesthesia was very heterogeneous. The results recommend the usage of this picture book in everyday practice, if the use of local anaesthesia could not be used in an appropriate way. © Georg Thieme Verlag KG Stuttgart · New York.

  3. Withdrawal Method (Coitus Interruptus)

    Science.gov (United States)

    Withdrawal method (coitus interruptus) Overview The withdrawal method of contraception, also known as coitus interruptus, is the practice of withdrawing the penis from the vagina and away from a woman's external ...

  4. Methadone Detoxification Versus Traditional Gradual Decrease in the Consumed Amount of Refined Opium Dross (Shireh: The Preferred Method for Controlling Withdrawal Syndrome

    Directory of Open Access Journals (Sweden)

    Mohammadreza Farsinejad

    2012-08-01

    Full Text Available Background: The aim of this study was to compare the effectiveness of methadone detoxification with traditional method of gradual decrement in the abused amount of the refined opium dross (Shireh to control withdrawal syndrome in Shireh-addicted patients. Methods: In this study, two groups of Shireh addicts were compared. The first group was treated by methadone and the second group by gradual decrement in the amount of consumed Shireh. Those experiencing the adverse effects of the treatment were excluded from the study. Methadone dose was calculated based on the amount of the Shireh consumed and detoxification was performed during a 21-day period. In the second group, the amount of the consumed Shireh was gradually decreased within 21 days and some of the withdrawal symptoms were selected as indicators for patient evaluation. Results: A total of 35 patients (16 versus 19 patients in the first and second groups were evaluated. Their mean age was 43 ± 4 years and all were male. A statistically significant difference was found between these two groups in terms of severity and duration of withdrawal symptoms within the first five days and their duration after the 21st day of the onset of detoxification (P< 0.05. Conclusion: In comparison with methadone detoxification, traditional method of gradually decreasing the consumed amount of Shireh controls the severity and duration of withdrawal symptoms better in the course of detoxification.

  5. CONTROL SYSTEM

    Science.gov (United States)

    Shannon, R.H.; Williamson, H.E.

    1962-10-30

    A boiling water type nuclear reactor power system having improved means of control is described. These means include provisions for either heating the coolant-moderator prior to entry into the reactor or shunting the coolantmoderator around the heating means in response to the demand from the heat engine. These provisions are in addition to means for withdrawing the control rods from the reactor. (AEC)

  6. FUEL ROD ASSEMBLY

    Science.gov (United States)

    Hutter, E.

    1959-09-01

    A cluster of nuclear fuel rods aod a tubular casing through which a coolant flows in heat-change contact with the ruel rods are described. The casting is of trefoil section and carries the fuel rods, each of which has two fin engaging the serrated fins of the other two fuel rods, whereby the fuel rods are held in the casing and are interlocked against relative longitudinal movement.

  7. Moles and Mole Control on British Farms, Amenities and Gardens after Strychnine Withdrawal

    Directory of Open Access Journals (Sweden)

    Sandra E. Baker

    2016-06-01

    Full Text Available Moles are considered pests in Britain, but this issue has been little studied. Lower welfare standards have been tolerated for moles than for most other managed wild mammal species, as use of both the controversial poison, strychnine, and unregulated traps have been permitted. Strychnine was withdrawn in 2006 and there were fears that mole populations would increase as a result. In 2007, we conducted a comprehensive, nationwide survey of land manager perceptions, opinions and behaviour regarding moles and mole control on farms, amenities and domestic gardens in Britain. We surveyed 2150 land managers (achieving a 59% response rate and ground-truthed 29 responses. Moles were reported to be present on most farms and amenities, and 13% of gardens, and were more common in lighter soils. Where present, moles were usually considered pests, this being more likely in Wales, Scotland and northern England, on livestock and mixed farms, and on large, high-value amenities, e.g., racecourses and golf courses. Mole control followed similar patterns to mole presence. More control may occur than is economically, and therefore potentially ethically, justified. Control should be more carefully considered and, where necessary, more effectively targeted. Kill-trapping was the favoured recent and future method on farms and amenities, even if strychnine was to be reintroduced; however, because mole traps are currently unregulated, some might not meet current welfare standards if tested. We found no evidence for an increase in moles since a farm questionnaire survey conducted in 1992; this could have wider implications for future wildlife management policy changes.

  8. Automatic safety rod for reactors. [LMFBR

    Science.gov (United States)

    Germer, J.H.

    1982-03-23

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

  9. Monte Carlo Simulations Suggest Current Chlortetracycline Drug-Residue Based Withdrawal Periods Would Not Control Antimicrobial Resistance Dissemination from Feedlot to Slaughterhouse

    Directory of Open Access Journals (Sweden)

    Casey L. Cazer

    2017-09-01

    Full Text Available Antimicrobial use in beef cattle can increase antimicrobial resistance prevalence in their enteric bacteria, including potential pathogens such as Escherichia coli. These bacteria can contaminate animal products at slaughterhouses and cause food-borne illness, which can be difficult to treat if it is due to antimicrobial resistant bacteria. One potential intervention to reduce the dissemination of resistant bacteria from feedlot to consumer is to impose a withdrawal period after antimicrobial use, similar to the current withdrawal period designed to prevent drug residues in edible animal meat. We investigated tetracycline resistance in generic E. coli in the bovine large intestine during and after antimicrobial treatment by building a mathematical model of oral chlortetracycline pharmacokinetics-pharmacodynamics and E. coli population dynamics. We tracked three E. coli subpopulations (susceptible, intermediate, and resistant during and after treatment with each of three United States chlortetracycline indications (liver abscess reduction, disease control, disease treatment. We compared the proportion of resistant E. coli before antimicrobial use to that at several time points after treatment and found a greater proportion of resistant enteric E. coli after the current withdrawal periods than prior to treatment. In order for the proportion of resistant E. coli in the median beef steer to return to the pre-treatment level, withdrawal periods of 15 days after liver abscess reduction dosing (70 mg daily, 31 days after disease control dosing (350 mg daily, and 36 days after disease treatment dosing (22 mg/kg bodyweight for 5 days are required in this model. These antimicrobial resistance withdrawal periods would be substantially longer than the current U.S. withdrawals of 0–2 days or Canadian withdrawals of 5–10 days. One published field study found similar time periods necessary to reduce the proportion of resistant E. coli following

  10. Development of a parallel processing couple for calculations of control rod worth in terms of burn-up in a WWER-1000 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Noori-Kalkhoran, Omid; Ahangari, R. [Nuclear Science and Technology Research Institute (NSTRI), Tehran (Iran, Islamic Republic of). Reactor Research school; Shirani, A.S. [Shahid Beheshti Univ., Tehran (Iran, Islamic Republic of). Faculty of Engineering

    2017-03-15

    In this study a code based method has been developed for calculation of integral and differential control rod worth in terms of burn-up for a WWER-1000 reactor. Parallel processing of WIMSD-5B, PARCS V2.7 and COBRA-EN has been used for this purpose. WIMSD-5B has been used for cell calculation and handling burn-up of core at different days. PARCS V2.7?has been used for neutronic calculation of core and critical boron concentration search. Thermal-hydraulic calculation has been performed by COBRA-EN. A Parallel processing algorithm has been developed by MATLAB to couple and transfer suitable data between these codes in each step. Steady-State Power Picking Factors (PPFs) of the core and Control rod worth have been calculated from Beginning Of Cycle (BOC) to 289.7 Effective full Power Days (EFPDs) in some steps. Results have been compared with Bushehr Nuclear Power Plant (BNPP) Final Safety Analysis Report (FSAR) results. The results show great similarity and confirm the ability of developed coupling in calculation of control rod worth in terms of burn-up.

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

    International Nuclear Information System (INIS)

    Bosselut, D.; Soulier, B.

    1997-03-01

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

  12. Determination of curve 1/M profile as a function of control rod bank position; Determinacao do perfil da curva 1/M em funcao da posicao dos bancos de barras de controle

    Energy Technology Data Exchange (ETDEWEB)

    Pereira, Valmir; Martinez, Aquilino Senra; Silva, Fernando Carvalho da [Universidade Federal, Rio de Janeiro, RJ (Brazil). Coordenacao dos Programas de Pos-graduacao de Engenharia. Programa de Engenharia Nuclear

    2002-07-01

    Determination of the subcritical multiplication curve profile (1/M) as a function of control rod bank position is of paramount importance to the development of a system which allows to foresee and also anticipate determination of criticality of a PWR reactor core. This work aims at determining this profile. For that, the 3D- two group-diffusion equations for a subcritical PWR reactor core with external neutron source is solved for different control rod bank positions. Results obtained are compared with the results from the corresponding eigenvalue problem, in order to verify how the external neutron source interferes with the reactor criticality search. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-07-01

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

  14. Nuclear reactor with scrammable part length rod

    International Nuclear Information System (INIS)

    Bevilacqua, F.

    1979-01-01

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

  15. Atomoxetine treatment for nicotine withdrawal: a pilot double-blind, placebo-controlled, fixed-dose study in adult smokers

    Directory of Open Access Journals (Sweden)

    Silverstone Peter H

    2012-03-01

    Full Text Available Abstract Background Many effective treatments for nicotine addiction inhibit noradrenaline reuptake. Three recent studies have suggested that another noradrenaline reuptake inhibitor, atomoxetine, may reduce smoking behaviors. Methods The present double-blind, placebo-controlled, fixed-dose study was carried out over 21 days during which administration of 40 mg atomoxetine was compared to placebo in 17 individuals. Of these, nine were randomized to atomoxetine and eight to placebo. Baseline and weekly measurements were made using the Cigarette Dependence Scale (CDS, Cigarette Withdrawal Scale (CWS, Questionnaire of Smoking Urges (QSU, reported number of cigarettes smoked, and salivary cotinine levels. Results The study results showed that all those on placebo completed the study. In marked contrast, of the nine individuals who started on atomoxetine, five dropped out due to side effects. In a completer analysis there were statistically significant differences at 14 and 21 days in several measures between the atomoxetine and placebo groups, including CDS, CWS, QSU, number of cigarettes smoked (decreasing to less than two per day in the treatment group who completed the study, and a trend towards lower mean salivary cotinine levels. However, these differences were not seen in a last observation carried forward (LOCF analysis. Conclusions In summary, this is the first study to examine the use of atomoxetine in non-psychiatric adult smokers for a period of more than 7 days, and the findings suggest that atomoxetine might be a useful treatment for nicotine addiction. However, the dose used in the current study was too high to be tolerated by many adults, and a dose-finding study is required to determine the most appropriate dose for future studies of this potential treatment for smoking cessation.

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

    International Nuclear Information System (INIS)

    Mankamo, Tuomas

    2006-11-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Mankamo, Tuomas [Avaplan Oy, Espoo (Finland)

    2006-11-15

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

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

    International Nuclear Information System (INIS)

    Aly, Omar Fernandes

    2006-01-01

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

  19. Nuclear fuel rod loading apparatus

    International Nuclear Information System (INIS)

    King, H.B.

    1981-01-01

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

  20. Cellular Internalization of Rod-Like Nanoparticles with Various Surface Patterns: Novel Entry Pathway and Controllable Uptake Capacity.

    Science.gov (United States)

    Xue, Jiaxiao; Guan, Zhou; Lin, Jiaping; Cai, Chunhua; Zhang, Wenjie; Jiang, Xinquan

    2017-06-01

    The cellular internalization of rod-like nanoparticles (NPs) is investigated in a combined experimental and simulation study. These rod-like nanoparticles with smooth, abacus-like (i.e., beads-on-wires), and helical surface patterns are prepared by the cooperative self-assembly of poly(γ-benzyl-l-glutamate)-block-poly(ethylene glycol) (PBLG-b-PEG) block copolymers and PBLG homopolymers. All three types of NPs can be internalized via endocytosis. Helical NPs exhibit the best endocytic efficacy, followed by smooth NPs and abacus-like NPs. Coarse-grained molecular dynamics simulations are used to examine the endocytic efficiency of these NPs. The NPs with helical and abacus-like surfaces can be endocytosed via novel "standing up" (tip entry) and "gyroscope-like" (precession) pathways, respectively, which are distinct from the pathway of traditional NPs with smooth surfaces. This finding indicates that the cellular internalization capacity and pathways can be regulated by introducing stripe patterns (helical and abacus-like) onto the surface of rod-like NPs. The results of this study may lead to novel applications of biomaterials, such as advanced drug delivery systems. © 2017 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  1. Tie rod insertion test

    CERN Multimedia

    B. LEVESY

    2002-01-01

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

  2. Nuclear fuel rod end plug weld inspection

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  3. Design and realization experience of Advanced Control Rod Group and Individual Control System (GIC) for VVER-1000 reactors

    International Nuclear Information System (INIS)

    Cerny, V.; Novy, L.; Janour, J.; Ris, M.; Zidek, P.

    1997-01-01

    During the reactor refueling outage of unit 1 of the South Ukrainian nuclear power plant in mid-1996, full replacement of the reactor's group and individual control (GIC) system was performed. The main functions of the GIC system are briefly characterized. The structure of the advanced GIC system is described and shown by means of a diagram. The criteria used in deciding on the upgrading strategy are discussed in some detail. The implementation of the replacement is also dealt with, as is the testing and commissioning of the system. (A.K.)

  4. Fuel rod leak detector

    International Nuclear Information System (INIS)

    Womack, R.E.

    1978-01-01

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

  5. Effects of nicotine withdrawal on panic-like response to breath holding: a placebo-controlled, double-blind, crossover patch study.

    Science.gov (United States)

    Cosci, Fiammetta; Bertoli, Giuly; Abrams, Kenneth

    2013-12-01

    Cigarette smoking may increase the likelihood of developing panic disorder. Periods of nicotine withdrawal, in particular, may promote panic in individuals high in anxiety sensitivity. We examined the importance of nicotine withdrawal in the occurrence of smoking and panic. We utilized a placebo-controlled, double-blind, randomized, crossover design. Fifty smokers underwent a breath-holding (BH) challenge after the transdermal administration of nicotine on one test day and a placebo on another test day. Physiological and psychological variables were assessed at baseline as well as directly before and after the challenges. Nicotine abstinence induced a decrease in heart rate and systolic blood pressure (BP) before the BH procedure (heart rate: 78.80 ± 11.43 under nicotine, 70.88 ± 10.83 under placebo; systolic BP: 124.90 ± 11.34 under nicotine, 121.18 ± 13.44 under placebo) and shorter BH duration relative to the nicotine patch condition. Nicotine abstinence did not, though, increase fear reactivity to the challenge. The findings for heart rate and BP are consistent with the stimulant properties of nicotine. The reduced capacity to maintain apnea under placebo might be due to carbon dioxide (CO2 ) hypersensitivity during periods of nicotine abstinence. The negative findings regarding fear reactivity might be due to BH being a relatively weak anxiogen. Future researchers are encouraged to employ CO2 -inhalation procedures to study the relationship between nicotine withdrawal and panic. © 2013 Wiley Periodicals, Inc.

  6. [Cost] effectiveness of withdrawal of fall-risk increasing drugs versus conservative treatment in older fallers: design of a multicenter randomized controlled trial (IMPROveFALL-study).

    Science.gov (United States)

    Hartholt, Klaas A; Boyé, Nicole D A; Van der Velde, Nathalie; Van Lieshout, Esther M M; Polinder, Suzanne; De Vries, Oscar J; Kerver, Albert J H; Ziere, Gijsbertus; Bruijninckx, Milko M M; De Vries, Mark R; Mattace-Raso, Francesco U S; Uitterlinden, André G; Van Beeck, Ed F; Lips, Paul; Patka, Peter; Van der Cammen, Tischa J M

    2011-08-21

    Fall incidents represent an increasing public health problem in aging societies worldwide. A major risk factor for falls is the use of fall-risk increasing drugs. The primary aim of the study is to compare the effect of a structured medication assessment including the withdrawal of fall-risk increasing drugs on the number of new falls versus 'care as usual' in older adults presenting at the Emergency Department after a fall. A prospective, multi-center, randomized controlled trial will be conducted in hospitals in the Netherlands. Persons aged ≥65 years who visit the Emergency Department due to a fall are invited to participate in this trial. All patients receive a full geriatric assessment at the research outpatient clinic. Patients are randomized between a structured medication assessment including withdrawal of fall-risk increasing drugs and 'care as usual'. A 3-monthly falls calendar is used for assessing the number of falls, fallers and associated injuries over a one-year follow-up period. Measurements will be at three, six, nine, and twelve months and include functional outcome, healthcare consumption, socio-demographic characteristics, and clinical information. After twelve months a second visit to the research outpatient clinic will be performed, and adherence to the new medication regimen in the intervention group will be measured. The primary outcome will be the incidence of new falls. Secondary outcome measurements are possible health effects of medication withdrawal, health-related quality of life (Short Form-12 and EuroQol-5D), costs, and cost-effectiveness of the intervention. Data will be analyzed using an intention-to-treat analysis. The successful completion of this trial will provide evidence on the effectiveness of withdrawal of fall-risk increasing drugs in older patients as a method for falls reduction. The trial is registered in the Netherlands Trial Register (NTR1593).

  7. [Cost]effectiveness of withdrawal of fall-risk increasing drugs versus conservative treatment in older fallers: design of a multicenter randomized controlled trial (IMPROveFALL-study

    Directory of Open Access Journals (Sweden)

    Mattace-Raso Francesco US

    2011-08-01

    Full Text Available Background Fall incidents represent an increasing public health problem in aging societies worldwide. A major risk factor for falls is the use of fall-risk increasing drugs. The primary aim of the study is to compare the effect of a structured medication assessment including the withdrawal of fall-risk increasing drugs on the number of new falls versus 'care as usual' in older adults presenting at the Emergency Department after a fall. Methods/Design A prospective, multi-center, randomized controlled trial will be conducted in hospitals in the Netherlands. Persons aged ≥65 years who visit the Emergency Department due to a fall are invited to participate in this trial. All patients receive a full geriatric assessment at the research outpatient clinic. Patients are randomized between a structured medication assessment including withdrawal of fall-risk increasing drugs and 'care as usual'. A 3-monthly falls calendar is used for assessing the number of falls, fallers and associated injuries over a one-year follow-up period. Measurements will be at three, six, nine, and twelve months and include functional outcome, healthcare consumption, socio-demographic characteristics, and clinical information. After twelve months a second visit to the research outpatient clinic will be performed, and adherence to the new medication regimen in the intervention group will be measured. The primary outcome will be the incidence of new falls. Secondary outcome measurements are possible health effects of medication withdrawal, health-related quality of life (Short Form-12 and EuroQol-5D, costs, and cost-effectiveness of the intervention. Data will be analyzed using an intention-to-treat analysis. Discussion The successful completion of this trial will provide evidence on the effectiveness of withdrawal of fall-risk increasing drugs in older patients as a method for falls reduction. Trial Registration The trial is registered in the Netherlands Trial Register (NTR1593

  8. Compacting spent fuel rods

    International Nuclear Information System (INIS)

    Wachter, W.J.

    1988-01-01

    A method and apparatus for compacting spent fuel rods comprises transferring the rods from a nuclear fuel rod assembly into a different nuclear fuel rod container having a smaller cross section than the assembly. The individual rods are moved from a fuel assembly and through a transition funnel by movable grippers at opposite ends of the funnel. One movable gripper reciprocates between gripping and release positions in a gap between the fuel assembly and the transition funnel. All of the fuel rods are withdrawn concurrently and are merged towards one another into a tighter array within the transition funnel and emerge as a bundle. A movable and a stationary bundle gripper are provided between the funnel and the storage container to advance the bundle of fuel rods into the container. (author)

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

    International Nuclear Information System (INIS)

    Zhang Fan; Chen Wenzhen; Yu Lei

    2008-01-01

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

  10. Nuclear fuel rods

    International Nuclear Information System (INIS)

    Wada, Toyoji.

    1979-01-01

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

  11. Lab-in-a-drop: controlled self-assembly of CdSe/ZnS quantum dots and quantum rods into polycrystalline nanostructures with desired optical properties

    International Nuclear Information System (INIS)

    Sukhanova, Alyona; Volkov, Yuri; Rogach, Andrey L; Baranov, Alexander V; Susha, Andrei S; Klinov, Dmitriy; Oleinikov, Vladimir; Cohen, Jacques H M; Nabiev, Igor

    2007-01-01

    Among the different nanometre-scale building blocks, colloidal nanocrystals are of special interest in construction of ordered assemblies to be used in optoelectronics, photonics and biosensing. It is important that the nanocrystal properties essential to allow the arrangement process, including their size, shape, surface protection, stabilization and charge, can be controlled along with the electronic structure of each nanocrystal. Here, we describe an operation of the 'lab-in-a-drop', droplets of the aqueous solutions of the water-solubilized CdSe/ZnS core/shell nanocrystal quantum dots and quantum rods, in which a variety of nanostructures with desired properties may be produced. We show that, upon incubation and controlled evaporation of the solvent from the aqueous droplets of nanocrystals, one may produce either nanowires or polycrystalline dendrites of different morphologies and dimensions, depending on the nanocrystal shape and on the very narrow concentration and temperature specific ranges. Hence, the operation of this 'lab-in-a-drop' is controlled by external parameters providing the fluorescent nanostructures of desired size and morphology. Although a majority of the results presented here were obtained with CdSe/ZnS quantum dots and rods, similar polycrystalline patterns may be produced in the aqueous suspensions of other nanocrystals

  12. Comparison of the worth of control and protection system rods of different design on the basis of the measurements in BN-600 reactor

    International Nuclear Information System (INIS)

    Vasilyev, B.A.; Roslyakov, V.F.; Farakshin, M.R.

    1988-01-01

    The results of the worth measurements of the basic and experimental absorbing rods of BN-600 reactor are presented. The procedure used for the rods worth comparison on the basis of calculated and experimental data interpretation is described here. Basic and experimental rods relative worth is also presented. (author). 5 refs, 3 figs, 2 tabs

  13. Development of ball bearing in high temperature water for in-vessel type control rod drive mechanism of advanced marine reactor

    International Nuclear Information System (INIS)

    Nunokawa, Hiroshi; Yoritsune, Tsutomu; Imayoshi, Shou; Ochiai, Masa-aki; Ishida, Toshihisa

    2001-06-01

    An advanced marine reactor MRX designed by Japan Atomic Energy Research Institute (JAERI) adopts an in-vessel type control rod drive mechanism, which is installed inside the reactor vessel. Since the in-vessel type control rod drive mechanism should work at a severe condition of a high temperature and high pressure water - 310degC and 12 MPa -, the JAERI has developed the components, a ball bearing of which especially is one of key technologies for realization of this type mechanism. The present report describes the development of the ball bearing containing a survey of materials, material screening tests on oxidation in an autoclave and rolling wear by a small facility, a trial fabrication of the full size ball bearing, and endurance test of it in the high temperature water. As a result, it was found from the development that the materials of cobalt alloy for both of the inner and outer races, cermet for the ball, and graphite for the retainer can satisfy the design condition of the ball bearing. (author)

  14. Elastic Wave Control Beyond Band-Gaps: Shaping the Flow of Waves in Plates and Half-Spaces with Subwavelength Resonant Rods

    Directory of Open Access Journals (Sweden)

    Andrea Colombi

    2017-08-01

    Full Text Available In metamaterial science, local resonance and hybridization are key phenomena strongly influencing the dispersion properties; the metasurface discussed in this article created by a cluster of resonators, subwavelength rods, atop an elastic surface being an exemplar with these features. On this metasurface, band-gaps, slow or fast waves, negative refraction, and dynamic anisotropy can all be observed by exploring frequencies and wavenumbers from the Floquet–Bloch problem and by using the Brillouin zone. These extreme characteristics, when appropriately engineered, can be used to design and control the propagation of elastic waves along the metasurface. For the exemplar we consider, two parameters are easily tuned: rod height and cluster periodicity. The height is directly related to the band-gap frequency and, hence, to the slow and fast waves, while the periodicity is related to the appearance of dynamic anisotropy. Playing with these two parameters generates a gallery of metasurface designs to control the propagation of both flexural waves in plates and surface Rayleigh waves for half-spaces. Scalability with respect to the frequency and wavelength of the governing physical laws allows the application of these concepts in very different fields and over a wide range of lengthscales.

  15. Wear of control rod cluster assemblies and of instrumentation thimbles: first results obtained with the vibrateau wear simulator

    International Nuclear Information System (INIS)

    Zbinden, M.; Hersant, D.

    1993-07-01

    Several REP components are affected by a particular sort of damage called impact/sliding wear. This kind of wear, originating from flow induced vibrations, affects loosely supported tubular structures. The main involved components are: - the RCCAs claddings and the guides tubes, - the instrumentation thimbles, - the fuel rods claddings, - the SG tubes. The R and D Division is concerned with studies aiming to understand and to master the phenomena leading to this wear. The MTC Branch is charged of the study of the wear itself. Tests are carried out on wear rigs to understand and to model wear mechanisms. The following work is related to the two first wear tests campaigns on the VIBRATEAU wear simulator: - a reproducibility test series in order to assess the spreading of the experimental results, - a comparative test series on surface treatments used to improve the components war resistance. (authors). 7 figs., 2 tabs., 4 refs

  16. Microstructural engineering applied to the controlled cooling of steel wire rod: Part II. Microstructural evolution and mechanical properties correlations

    Science.gov (United States)

    Campbell, P. C.; Hawbolt, E. B.; Brimacombe, J. K.

    1991-11-01

    In the second part of this paper, the microstructural evolution and mechanical properties of plain-carbon steel rods which have been subjected to known cooling conditions are described. Specifically, the isothermal phase transformation kinetics for the decomposition of austenite into ferrite and pearlite have been determined with a diametral dilatometer and characterized in terms of empirical coefficients in the Avrami equation. The continuous cooling transformation (CCT) start time, fraction ferrite, ferrite grain diameter, and pearlite interlamellar spacing have been quantified and correlated with steel composition and cooling rate. Tensile tests have been conducted to obtain yield strength (YS) and ultimate tensile strength (UTS), which, with literature data, have been related to the microstructure and composition of the steels. These correlations, which apply to both hypoeutectoid and eutectoid steels, have been incorporated in a mathematical model of the Stelmor process, to be described in Part III of this article.[441

  17. Low fluid level in pulse rod shock absorber

    International Nuclear Information System (INIS)

    Aderhold, H.C.

    1974-01-01

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

  18. Why Rods and Cocci

    Indian Academy of Sciences (India)

    experience greater frictional resistance. This hypothesis is supported by the fact that among the flagellated motile bacteria almost all are rod shaped. Only exceptionally few cocci are motile. This hypothesis, however, is not adequate since a large number of species of bacteria are non-motile. A rod shape can confer another ...

  19. Why Rods and Cocci

    Indian Academy of Sciences (India)

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

  20. INFLUENCE OF DYNAMIC MAGNETIZATION TO IMPROVE THE EFFICIENCY OF ELECTROMAGNETIC-ACOUSTIC TRANSFORMATION WITH WAVEGUIDE CONTROL RODS

    Directory of Open Access Journals (Sweden)

    D. V. Zlobin

    2017-01-01

    Full Text Available The disadvantage of the electromagnetic-acoustic (EMA method receiving ultrasonic waves are low efficiency. The traditional way to enhance its effectiveness is increase the bias field. The aim of the study was research the way to improve the efficiency of the EMA transformation, using a time-varying bias field.The researches held with the help of a specially designed installation that allows the magnetization to be performed by a constant and alternating magnetic field (dynamic bias, synchronously with the passage of the received pulse. The object of the study were rods made of different grades of steel with a diameter of 4–6 mm, in which the symmetrical zero mode S0 of the rod wave was excited by the EMA method (in the frequency range of about 40 kHz. A comparative analysis of the amplitudes and form pulses of multiple reflections during static and dynamic reversal of magnetization and with a full cycle of magnetization reversal conducted.The result of the efficiency measurements EMA reception during static and dynamic bias found a significant (up to 5 times increase in the signal amplitude on the receiving transducer. Taking into account that the main contribution to the excitation mechanism and the reception mechanism made the magnetostrictive effect on low frecuncy, it can assumed that using a dynamic bias field is impacting significant on the effective mobility of magnetic domains (that is changes the dynamic magnetic susceptibility of the material. It is established that it is possible to monitor steel at lower values of the bias field, and, consequently, to reduce the mass dimensions of the magnetic system.Thus, in the course of the researchers found of effect of dynamic bias and effect of dynamic bias increase acoustic pulse amplitude of the signal of the received EMA method. Using this method will improve the quality EMA testing by creating more efficient EMA transducer. Taking into account that the value of the detected effect

  1. [Psychotic experiences in the course of alcohol withdrawal symptoms: locus of control among patients with and without delirium and analysis of subjective experiences in delirium].

    Science.gov (United States)

    Kokoszka, Andrzej; Laskowska, Marta; Mikuła, Joanna

    2011-01-01

    The comparison of the locus of control in groups of patients hospitalised due to alcohol withdrawal with and without delirium and analysis of psychotic experiences of patients with delirium. 25 patients with alcohol withdrawal with delirium and 25 without delirium took part in the study. They filled-in the Internal-External (I-E) Locus of Control Scale by Rotter; Multidimensional Health Locus of Control (MHLC) scale; the group with delirium also did the Psychopathological Symptoms Inventory, by Bizoń et al. The mean score in I-E Locus of Control Scale in the group with delirium was more external than in the group without delirium (M = 13.28; SD = 2.762 versus M = 11.64; SD = 2.612; t(48) = -2.157; p = 0.036). Group with delirium had also lower mean score in the dimension of internal control in MHLC, than the group without delirium (M = 24.8; SD = 6.149 versus M = 26.8; SD = 4.648; t(48) = 1.99; p = 0.04). There were no statistically significant differences between the groups in the other subscales. The auditory and visual hallucinations were most common among patients with delirium (84%, 80% respectively, as well as delusions of taking part in not existing events (92%) and persecutory delusions (80%). Psychotic experiences influenced behaviour in nearly 50% of the cases. A more external locus of control may be one of the factors contributing to the development of alcohol delirium. The content of psychotic experiences seems to have impact on the behaviour of many patients with alcohol delirium.

  2. Rod Photoreceptors Detect Rapid Flicker

    Science.gov (United States)

    Conner, J. D.; MacLeod, Donald I. A.

    1977-01-01

    Rod-isolation techniques show that light-adapted human rods detect flicker frequencies as high as 28 hertz, and that the function relating rod critical flicker frequency to stimulus intensity contains two distinct branches. (MLH)

  3. Neonatal opioid withdrawal syndrome.

    Science.gov (United States)

    Sutter, Mary Beth; Leeman, Lawrence; Hsi, Andrew

    2014-06-01

    Neonatal opioid withdrawal syndrome is common due to the current opioid addiction epidemic. Infants born to women covertly abusing prescription opioids may not be identified as at risk until withdrawal signs present. Buprenorphine is a newer treatment for maternal opioid addiction and appears to result in a milder withdrawal syndrome than methadone. Initial treatment is with nonpharmacological measures including decreasing stimuli, however pharmacological treatment is commonly required. Opioid monotherapy is preferred, with phenobarbital or clonidine uncommonly needed as adjunctive therapy. Rooming-in and breastfeeding may decease the severity of withdrawal. Limited evidence is available regarding long-term effects of perinatal opioid exposure. Copyright © 2014 Elsevier Inc. All rights reserved.

  4. Efficacy of long-term milnacipran treatment in patients meeting different thresholds of clinically relevant pain relief: subgroup analysis of a randomized, double-blind, placebo-controlled withdrawal study.

    Science.gov (United States)

    Mease, Philip J; Clauw, Daniel J; Trugman, Joel M; Palmer, Robert H; Wang, Yong

    2014-01-01

    Fibromyalgia patients from a long-term, open-label study of milnacipran (50-200 mg/day) were eligible to participate in a 12-week, randomized, placebo-controlled withdrawal study. The withdrawal study evaluated loss of therapeutic response in patients who achieved ≥50% pain improvements after receiving up to 3.25 years of milnacipran. This post-hoc analysis investigated whether patients who met lower thresholds of pain improvement also experienced worsening of fibromyalgia symptoms upon treatment withdrawal. Among patients who received milnacipran ≥100 mg/day during the long-term study, three subgroups were identified based on percentage of pain reduction at randomization: ≥50% (protocol-defined "responders"; n=150); ≥30% to Fibromyalgia Impact Questionnaire-Revised (FIQR), 36-Item Short-Form Health Survey Physical Component Summary (SF-36 PCS), and Beck Depression Inventory (BDI). In the ≥30 to subgroup, significant worsening in pain was detected after treatment withdrawal. The difference between placebo and milnacipran in mean VAS score changes for this subgroup (+9.0, Psubgroup, no worsening in pain was observed in either treatment arm. However, patients in this subgroup experienced significant worsening in FIQR scores after treatment withdrawal (placebo, +6.9; milnacipran, -2.8; Pfibromyalgia symptoms after treatment withdrawal. These results suggest that the conventional ≥30% pain responder cutoff may be adequate to demonstrate efficacy in randomized withdrawal studies of fibromyalgia. Patients in the subgroup had worsening scores on the FIQR and other multidimensional measures after treatment withdrawal, indicating the importance of identifying and managing the multiple symptoms of fibromyalgia.

  5. Palonosetron and hydroxyzine pre-treatment reduces the objective signs of experimentally-induced acute opioid withdrawal in humans: a double-blinded, randomized, placebo-controlled crossover study.

    Science.gov (United States)

    Erlendson, Matthew J; D'Arcy, Nicole; Encisco, Ellen M; Yu, Jeffrey J; Rincon-Cruz, Lorena; Peltz, Gary; Clark, J David; Chu, Larry F

    2017-01-01

    Treatments for reducing opioid withdrawal are limited and prone to problematic side effects. Laboratory studies, clinical observations, and limited human trial data suggest 5-HT3-receptor antagonists and antihistamines may be effective. This double-blind, crossover, placebo-controlled study employing an acute physical dependence model evaluated whether (i) treatment with a 5-HT3-receptor antagonist (palonosetron) would reduce opioid withdrawal symptoms, and (ii) co-administration of an antihistamine (hydroxyzine) would enhance any treatment effect. At timepoint T = 0, healthy (non-opioid dependent, non-substance abuser) male volunteers (N = 10) were pre-treated with either a) placebo, b) palonosetron IV (0.75 mg), or c) palonosetron IV (0.75 mg) and hydroxyzine PO (100 mg) in a crossover study design. This was followed at T = 30 by intravenous morphine (10 mg/70kg). At T = 165, 10 mg/70kg naloxone IV was given to precipitate opioid withdrawal. The objective opioid withdrawal score (OOWS) and subjective opioid withdrawal score (SOWS) were determined 5 and 15 minutes after naloxone administration (T = 170, 180, respectively). Baseline measurements were recorded at T = -30 and T = -15. Comparison of average baseline OOWS scores with OOWS scores obtained 15 minutes after naloxone was significant (p = 0.0001). Scores from 15 minutes post-naloxone infusion showed significant differences in OOWS scores between treatment groups: placebo, 3.7 ± 2.4; palonosetron, 1.5 ± 0.97; and palonosetron with hydroxyzine, 0.2 ± 0.1333. Pretreatment with palonosetron significantly reduced many signs of experimentally-induced opioid withdrawal. Co-administration with hydroxyzine further reduced opioid withdrawal severity. These results suggest that 5-HT3 receptor antagonists, alone or in combination with an antihistamine, may be useful in the treatment of opioid withdrawal.

  6. Response Surface Methodology Control Rod Position Optimization of a Pressurized Water Reactor Core Considering Both High Safety and Low Energy Dissipation

    Directory of Open Access Journals (Sweden)

    Yi-Ning Zhang

    2017-02-01

    Full Text Available Response Surface Methodology (RSM is introduced to optimize the control rod positions in a pressurized water reactor (PWR core. The widely used 3D-IAEA benchmark problem is selected as the typical PWR core and the neutron flux field is solved. Besides, some additional thermal parameters are assumed to obtain the temperature distribution. Then the total and local entropy production is calculated to evaluate the energy dissipation. Using RSM, three directions of optimization are taken, which aim to determine the minimum of power peak factor Pmax, peak temperature Tmax and total entropy production Stot. These parameters reflect the safety and energy dissipation in the core. Finally, an optimization scheme was obtained, which reduced Pmax, Tmax and Stot by 23%, 8.7% and 16%, respectively. The optimization results are satisfactory.

  7. Effect of relative position of the control rods in the power calibration of the running chains of the RP-10 nuclear reactor

    International Nuclear Information System (INIS)

    Zuniga, Agustin; Arrieta, Rolando

    2014-01-01

    The monitoring of the thermal power levels of an experimental nuclear reactor, such as the RP-10 reactor is performed by current levels reported by the driving chains and their correlation with the power. The calibration curves for each channel will depend on the neutron distribution observed for the corresponding configuration. One way to modify this configuration is positioning the control rods that can be chosen according to the convenience of providing greater neutron flux in a given irradiation position. This article shows this variation, and consequently the current power curves corresponding thereto. We have chosen two relative positions of the control rods (BC1 and BC2), the first is the one that favors the production of radioisotopes (PPR) (Case 1) and the second favors the use of neutron activation (NAA) (Case 2). The used method was based on the record of the current (up camera, M4) by the acquisition system (SAD) of the reactor and the power from the digital reading of nitrogen 16, which was previously calibrated by thermal balance. The results for the nuclear configuration 30, M4 chain up were: Case 1: I M4 (A) = 2.95 x 10 -6 x P (MW) + 1.45 x 10 -6 R 2 = 9.99 x 10 -1 and Case 2: I M4 (A) = 2.39 x 10 -6 x P (MW) + 5.73 x 10 -7 , R 2 = 9.97 x 10 -1 . The dispersions of the data are in the order of 10 % for low power (under 2 MW), decreasing as the power rises to under 3 % (for values greater than 3 MW). (authors).

  8. In vitro formation and activity-dependent plasticity of synapses between Helix neurons involved in the neural control of feeding and withdrawal behaviors.

    Science.gov (United States)

    Fiumara, F; Leitinger, G; Milanese, C; Montarolo, P G; Ghirardi, M

    2005-01-01

    Short-term activity-dependent synaptic plasticity has a fundamental role in short-term memory and information processing in the nervous system. Although the neuronal circuitry controlling different behaviors of land snails of the genus Helix has been characterized in some detail, little is known about the activity-dependent plasticity of synapses between identified neurons regulating specific behavioral acts. In order to study homosynaptic activity-dependent plasticity of behaviorally relevant Helix synapses independently of heterosynaptic influences, we sought to reconstruct them in cell culture. To this aim, we first investigated in culture the factors regulating synapse formation between Helix neurons, and then we studied the short-term plasticity of in vitro-reconstructed monosynaptic connections involved in the neural control of salivary secretion and whole-body withdrawal. We found that independently of extrinsic factors, cell-cell interactions are seemingly sufficient to trigger the formation of electrical and chemical synapses, although mostly inappropriate--in their type or association--with respect to the in vivo synaptic connectivity. The presence of ganglia-derived factors in the culture medium was required for the in vitro reestablishment of the appropriate in vivo-like connectivity, by reducing the occurrence of electrical connections and promoting the formation of chemical excitatory synapses, while apparently not influencing the formation of inhibitory connections. These heat-labile factors modulated electrical and chemical synaptogenesis through distinct protein tyrosine kinase signal transduction pathways. Taking advantage of in vitro-reconstructed synapses, we have found that feeding interneuron-efferent neuron synapses and mechanosensory neuron-withdrawal interneuron synapses display multiple forms of short-term enhancement-like facilitation, augmentation and posttetanic potentiation as well as homosynaptic depression. These forms of plasticity

  9. Direct costs associated with the management of progressive early onset scoliosis: estimations based on gold standard technique or with magnetically controlled growing rods.

    Science.gov (United States)

    Charroin, C; Abelin-Genevois, K; Cunin, V; Berthiller, J; Constant, H; Kohler, R; Aulagner, G; Serrier, H; Armoiry, X

    2014-09-01

    The main disadvantage of the surgical management of early onset scoliosis (EOS) using conventional growing rods is the need for iterative surgical procedures during childhood. The emergence of an innovative device using distraction-based magnetically controlled growing rods (MCGR) provides the opportunity to avoid such surgeries and therefore to improve the patient's quality of life. Despite the high cost of MCGR and considering its potential impact in reducing hospital stays, the use of MCGR could reduce medical resource consumption in a long-term view in comparison to traditional growing rod (TGR). A cost-simulation model was constructed to assess the incremental cost between the two strategies. The cost for each strategy was estimated based on probability of medical resource consumption determined from literature search as well as data from EOS patients treated in our centre. Some medical expenses were also estimated from expert interviews. The time horizon chosen was 4 years as from first surgical implantation. Costs were calculated in the perspective of the French sickness fund (using rates from year 2013) and were discounted by an annual rate of 4%. Sensitivity analyses were conducted to test model strength to various parameters. With a time horizon of 4 years, the estimated direct costs of TGR and MCGR strategies were 49,067 € and 42,752 €, respectively leading to an incremental costs of 6135 € in favour of MCGR strategy. In the first case, costs were mainly represented by hospital stays expenses (83.9%) whereas in the other the cost of MCGR contributed to 59.5% of the total amount. In the univariate sensitivity analysis, the tariffs of hospital stays, the tariffs of the MCG, and the frequency of distraction surgeries were the parameters with the most important impact on incremental cost. MCGR is a recent and promising innovation in the management of severe EOS. Besides improving the quality of life, its use in the treatment of severe EOS is likely to

  10. A randomized, open-label, standard controlled, parallel group study of efficacy and safety of baclofen, and chlordiazepoxide in uncomplicated alcohol withdrawal syndrome.

    Science.gov (United States)

    Girish, K; Vikram Reddy, K; Pandit, Lakshmi V; Pundarikaksha, H P; Vijendra, R; Vasundara, K; Manjunatha, R; Nagraj, Moulya; Shruthi, R

    2016-02-01

    Alcohol withdrawal syndrome (AWS) is a distressing condition, generally controlled by benzodiazepines (BZD's). Baclofen, a gamma-aminobutyric acid-B (GABAB) agonist, has also shown promising results in controlling AWS. As there are few studies comparing the efficacy and tolerability of chlordiazepoxide with baclofen, the present study was taken up. The objective of this study was to compare efficacy and tolerability of baclofen with chlordiazepoxide in uncomplicated AWS. Sixty subjects with uncomplicated AWS were randomized into two groups of 30 each, to receive baclofen (30 mg) or chlordiazepoxide (75 mg) in decremented fixed dose regime for 9 days. Clinical efficacy was assessed by Clinical Institute Withdrawal Assessment for Alcohol-Revised Scale (CIWA-Ar) and tolerability by the nature and severity of adverse events. Lorazepam was used as rescue medication. Secondary efficacy parameters were Clinical Global Impression scores, symptom-free days, and subject satisfaction as assessed by visual analog scale. This study was registered with Clinical Trial Registry-India (CTRI/2013/04/003588), also subsequently registered with WHO's ICTRP clinical trial portal. Both baclofen and chlordiazepoxide showed a consistent reduction in the total CIWA-Ar scores. However, chlordiazepoxide showed a faster and a more effective control of anxiety and agitation requiring lesser lorazepam supplementation, and also showed a better subject satisfaction compared to baclofen. Both the drugs showed good tolerability with mild self-limiting adverse events. The present study demonstrates that baclofen is not as good as chlordiazepoxide in the treatment of uncomplicated AWS. However, baclofen might be considered as an alternative. Copyright © 2016 Chang Gung University. Published by Elsevier B.V. All rights reserved.

  11. Controlled synthesis and photoluminescence properties of In{sub 2}O{sub 3} rods with dodecahedron In{sub 2}O{sub 3} microcrystals on top

    Energy Technology Data Exchange (ETDEWEB)

    Ouacha, Hassan [King Abdullah Institute for Nanotechnology, King Saud University, Riyadh (Saudi Arabia); Hendaoui, Ali [Department of Physics, College of Science and General Studies, Alfaisal University, Riyadh (Saudi Arabia); Kleineberg, Ulf [Faculty of Physics, Ludwig Maximilian University of Munich, Garching (Germany); Albrithen, Hamad; Azzeer, Abdallah [Physics and Astronomy Department, King Saud University, Riyadh (Saudi Arabia)

    2017-10-15

    In{sub 2}O{sub 3} rods with dodecahedron In{sub 2}O{sub 3} microcrystals on top were synthesized in an electrical furnace via Au-catalyzed vapor transport process. A catalyst-assisted selective vapor-solid (VS) growth was proposed to explain the formation of the dodecahedron In{sub 2}O{sub 3} microcrystal, while the self-catalytic VS growth mechanism dominated the subsequent one-dimensional (1D) growth of the In{sub 2}O{sub 3} rod underneath the In{sub 2}O{sub 3} microcrystal. The structural evolution of these structures was carefully examined during the synthesis process by controlling the growth parameters. The morphologies, crystalline structures and surface chemistry were characterized by scanning electron microscopy (SEM), X-ray diffraction technique (XRD), and X-ray photoelectron spectroscopy (XPS), respectively. The photoluminescence (PL) spectrum at room temperature of the as-grown In{sub 2}O{sub 3} structures exhibited both UV and blue luminescence emission under one excitation at 260 nm, which may be related to the existence of oxygen vacancies. The synthesized multifaceted In{sub 2}O{sub 3} microcrystal has shown to contain a large number of vertices and may find many applications in developing three-dimensional (3D) resonators. This work will not only enrich the synthesis science but also will open doors for applications of such structures in optical devices. (copyright 2017 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  12. Power Burst Facility: U(18)O2-CaO-ZrO2 Fuel Rods in Water

    Energy Technology Data Exchange (ETDEWEB)

    Jose Ignacio Marquez Damian; Alexis Weir; Valeria L. Putnam; John D. Bess

    2009-09-01

    The Power Burst Facility (PBF) reactor operated from 1972 to 1985 on the SPERT Area I of the Idaho National Laboratory, then known as Nuclear Reactor Test Station. PBF was designed to provide experimental data to aid in defining thresholds for and modes of failure under postulated accident conditions. PBF reactor startup testing began in 1972. This evaluation focuses on two operational loading tests, chronologically numbered 1 and 2, published in a startup-test report in 1974 [1]. Data for these tests was used by one of the authors to validate a MCNP model for criticality safety purposes [2]. Although specific references to original documents are kept in the text, all the reactor parameters and test specific data presented here was adapted from that report. The tests were performed with operational fuel loadings, a stainless steel in-pile tube (IPT) mockup, a neutron source, four pulse chambers, two fission chambers, and one ion chamber. The reactor's four transition rods (TRs) and control rods (CRs) were present but TR boron was completely withdrawn below the core and CR boron was partially withdrawn above the core. Test configurations differ primarily in the number of shim rods, and consequently the number of fuel rods included in the core. The critical condition was approached by incrementally and uniformly withdrawing CR boron from the core. Based on the analysis of the experimental data and numerical calculations, both experiments are considered acceptable as criticality safety benchmarks.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-07-01

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

  14. Visualization of groundwater withdrawals

    Science.gov (United States)

    Winston, Richard B.; Goode, Daniel J.

    2017-12-21

    Generating an informative display of groundwater withdrawals can sometimes be difficult because the symbols for closely spaced wells can overlap. An alternative method for displaying groundwater withdrawals is to generate a “footprint” of the withdrawals. WellFootprint version 1.0 implements the Footprint algorithm with two optional variations that can speed up the footprint calculation. ModelMuse has been modified in order to generate the input for WellFootprint and to read and graphically display the output from WellFootprint.

  15. Buprenorphine for managing opioid withdrawal.

    Science.gov (United States)

    Gowing, Linda; Ali, Robert; White, Jason M; Mbewe, Dalitso

    2017-02-21

    Managed withdrawal is a necessary step prior to drug-free treatment or as the endpoint of substitution treatment. To assess the effects of buprenorphine versus tapered doses of methadone, alpha 2 -adrenergic agonists, symptomatic medications or placebo, or different buprenorphine regimens for managing opioid withdrawal, in terms of the intensity of the withdrawal syndrome experienced, duration and completion of treatment, and adverse effects. We searched the Cochrane Central Register of Controlled Trials (CENTRAL, Issue 11, 2016), MEDLINE (1946 to December week 1, 2016), Embase (to 22 December 2016), PsycINFO (1806 to December week 3, 2016), and the Web of Science (to 22 December 2016) and handsearched the reference lists of articles. Randomised controlled trials of interventions using buprenorphine to modify the signs and symptoms of withdrawal in participants who were primarily opioid dependent. Comparison interventions involved reducing doses of methadone, alpha 2 -adrenergic agonists (clonidine or lofexidine), symptomatic medications or placebo, and different buprenorphine-based regimens. We used standard methodological procedures expected by Cochrane. We included 27 studies involving 3048 participants. The main comparators were clonidine or lofexidine (14 studies). Six studies compared buprenorphine versus methadone, and seven compared different rates of buprenorphine dose reduction. We assessed 12 studies as being at high risk of bias in at least one of seven domains of methodological quality. Six of these studies compared buprenorphine with clonidine or lofexidine and two with methadone; the other four studies compared different rates of buprenorphine dose reduction.For the comparison of buprenorphine and methadone in tapered doses, meta-analysis was not possible for the outcomes of intensity of withdrawal or adverse effects. However, information reported by the individual studies was suggestive of buprenorphine and methadone having similar capacity to

  16. Withdrawal Method (Coitus Interruptus)

    Science.gov (United States)

    ... fluid and pregnancy Withdrawal method (coitus interruptus) About Advertisement Mayo Clinic does not endorse companies or products. ... a Job Site Map About This Site Twitter Facebook Google YouTube Pinterest Mayo Clinic is a not- ...

  17. Reactor power control device

    International Nuclear Information System (INIS)

    Watanabe, Mitsutaka

    1997-01-01

    Hardware of an analog nuclear instrumentation system is reformed, a function generator is added to a setting calculation circuit of the nuclear instrumentation system, and each of setting lines of the nuclear instrumentation system is set in parallel with an upper limit curve in an operation region defined by a second order or third order equation. Upon transient change of abnormal power elevation during operation, scram signals are generated by power change in the same state as 100% rated operation due to elevation of reactor thermal power. Since the operation limit value relative to transient change due to power elevation can be made substantially equal with the same as that upon rated operation, the operation limit value for partial power operation state can be kept substantially the same level as that upon rated operation. When transition change caused by abnormal control rod withdrawal occurs during operation, a control rod withdrawal inhibition signal can ensure the power elevation width equal with that upon rated power operation, and since the withdrawal inhibition signal is generated in substantially the same withdrawing state, the operation limit value relative to a partial power operation state can be kept at the same level as that during rated operation. (N.H.)

  18. Numerical simulation of 900 MW control rods impact friction vibration and wear; Simulation numerique du comportement vibratoire et de l`usure des crayons de grappe de commande 900 MW

    Energy Technology Data Exchange (ETDEWEB)

    Jacquart, G.

    1993-12-01

    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.

  19. Flow in rod bundles

    International Nuclear Information System (INIS)

    Hazi, G.; Mayer, G.

    2005-01-01

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

  20. Burn-up Credit Criticality Safety Benchmark-Phase II-E. Impact of Isotopic Inventory Changes due to Control Rod Insertions on Reactivity and the End Effect in PWR UO2 Fuel Assemblies

    International Nuclear Information System (INIS)

    Neuber, Jens Christian; Tippl, Wolfgang; Hemptinne, Gwendoline de; Maes, Philippe; Ranta-aho, Anssu; Peneliau, Yannick; Jutier, Ludyvine; Tardy, Marcel; Reiche, Ingo; Kroeger, Helge; Nakata, Tetsuo; Armishaw, Malcom; Miller, Thomas M.

    2015-01-01

    The report describes the final results of the Phase II-E Burn-up Credit Criticality Benchmark conducted by the Expert Group on Burn-up Credit Criticality Safety. The objective of Phase II of the Burn-up Credit Criticality Safety programme is to study the impact of axial burn-up profiles of PWR UO 2 spent fuel assemblies on the reactivity of PWR UO 2 spent fuel assembly configurations. The objective of the Phase II-E benchmark was to study the impact of changes on the spent nuclear fuel isotopic composition due to control rod insertion during depletion on the reactivity and the end effect of spent fuel assemblies with realistic axial burn-up profiles for different control rod insertion depths ranging from 0 cm (no insertion) to full insertion (i.e. to the case that the fuel assemblies were exposed to control rod insertion over their full active length). For this purpose two axial burn-up profiles have been extracted from an AREVA-NP-GmbH-owned 17x17-(24+1) PWR UO 2 spent fuel assembly burn-up profile database. One profile has an average burn-up of 30 MWd/kg U, the other profile is related to an average burn-up of 50 MWd/kg U. Two profiles with different average burn-up values were selected because the shape of the burn-up profile is affected by the average burn-up and the end effect depends on the average burn-up of the fuel. The Phase II-E benchmark exercise complements the Phase II-C and Phase II-D benchmark exercises. In Phase II-D different irradiation histories were analysed using different control rod insertion histories during depletion as well as irradiation histories without control rod insertion. But in all the histories analysed a uniform distribution of the burn-up and hence a uniform distribution of the isotopic composition were assumed; and in all the histories including any usage of control rods full insertion of the control rods was assumed. In Phase II-C the impact of the asymmetry of axial burn-up profiles on the reactivity and the end effect of