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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. Monitoring device for withdrawing control rods

    International Nuclear Information System (INIS)

    Higashigawa, Yuichi.

    1985-01-01

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 The symmetricity of control rods to be selected and withdrawn by an operator is judged by a control rod symmetry monitoring device, while the symmetricity of the withdrawn control rods is judged by a control rod withdrawal state monitoring device.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

  4. 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.)

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

    International Nuclear Information System (INIS)

    Smith, J. E.

    1984-01-01

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

  6. 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.

  7. Uncontrolled withdrawal of a control rod without scram

    International Nuclear Information System (INIS)

    Abou-EL-Maaty, Talal

    2008-01-01

    In the present work the thermal-hydraulics of reactivity-induced transients in low enriched uranium (LEU) core of a typical material test research reactor (MTR) are analyzed using the previous program developed by Khater et al. The analysis was done for uncontrolled withdrawal of a control rod with scram-disabled conditions. Initiating reactivity events with and without the influence of reactivity efficiency curve ('S' curve) were considered. The results of the proposed transients are analyzed and compared with each other. In transient without the 'S' curve influence, a high primary peak power of 406.18 MW is attained and a clad melt down takes place after 1.85 s. In the transient with the 'S' curve influence, a high super prompt-critical situation is produced (1.762$ at 0.895 s) with a very high primary peak power of 801.05 MW at 0.912 s. Also, a fast clad melt down is resulted in the hot channel at 1.088 s and a stable film boiling is established. This study indicates that, compared to the application of linear reactivity curve, the application of the reactivity efficiency curve results in the prediction of higher peaks in power and temperatures (fuel, clad and coolant) with a fast clad melt down

  8. Uncontrolled withdrawal of a control rod without SCRAM

    International Nuclear Information System (INIS)

    Abou-El-Maaty, T.

    2007-01-01

    In the present work, the thermal-hydraulic analysis of reactivity-induced transients in a Low Enriched Uranium (LEU) core of a typical material test research reactor is conducted using the previous program developed Khater et al. The analysis was done for the uncontrolled withdrawal of a control rod under scram-disabled conditions. The initiating event reactivity was considered with and without influence of the reactivity efficiency curve (''S'' curve). The results of the transient calculations are analyzed and compared with each other. In the transient without the ''S'' curve influence, a high primary peak power of 406.18 MW is attained and a clad melt down is occurring after 1.85 s. In the transient with the ''S'' curve influence, a super prompt highly critical situation is produced (1.762 $ at 0.895 s) with a very high primary peak power of 801.05 MW at 0.912 s. A fast clad melt down is resulting in the hot channel at 1.088 s and a stable film boiling is occurring. This study shows that the influence of the reactivity efficiency curve results in higher peaks in power and temperatures (fuel, clad and coolant) with a fast clad melt down than that of a linear assumption. (orig.)

  9. 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

  10. 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

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

    International Nuclear Information System (INIS)

    Yang, Soo Hyung; Chung, Young-Jong; Kim, Hee Cheol

    2005-01-01

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

  12. 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.

  13. 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)

  14. Modeling of Phenix End-of-Life control rod withdrawal benchmark with DYN3D SFR version

    Energy Technology Data Exchange (ETDEWEB)

    Nikitin, Evgeny; Fridman, Emil [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany). Reactor Safety

    2017-06-01

    The reactor dynamics code DYN3D is currently under extension for Sodium cooled Fast Reactor applications. The control rod withdrawal benchmark from the Phenix End-of-Life experiments was selected for verification and validation purposes. This report presents some selected results to demonstrate the feasibility of using DYN3D for steady-state Sodium cooled Fast Reactor analyses.

  15. 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.)

  16. Control rod

    International Nuclear Information System (INIS)

    Kawakami, Kazuo; Shimoshige, Takanori; Nishimura, Akira

    1979-01-01

    Purpose: A control rod has been developed, which provided a plurality of through-holes in the vicinity of the sheath fitting position, in order to flatten burn-up, of fuel rods in positions confronting a control rod. Thereby to facilitate the manufacture of the control rods and prevent fuel rod failures. Constitution: A plurality of through-holes are formed in the vicinity of the sheath fitting position of a central support rod to which a sheath for the control rod is fitted. These through-holes are arranged in the axial direction of the central support rod. Accordingly, burn-up of fuel rods confronting the control rods can be reduced by through-holes and fuel rod failures can be prevented. (Yoshino, Y.)

  17. Safety analysis results for the control rod banks withdrawal event at a full power of the SMART-P

    International Nuclear Information System (INIS)

    Yang, S. H.; Chung, Y. J.; Kim, H. C.; Zee, S. Q.

    2005-01-01

    For the validation of the 330 MWt SMART (System-integrated Modular Advanced ReacTor), a detailed design for the SMART-P has been accomplished by KAERI. In the SMART-P design similar to the SMART design, the soluble boron free design is adapted. This concept results in a larger reactivity worth of the control rod bank compared to that of the commercial pressurized water reactor. Moreover, in the SMART-P design, the control rod banks are fairly well inserted into the core, even at a full power condition. Therefore, accidents related to the reactivity anomalies have been evaluated as crucial events when compared to the other initiating events. In this paper, safety analysis for the control rod banks withdrawal event at a full power of the SMART-P has been accomplished by considering various initial conditions, different withdrawal times of the control rod banks and the reactivity feedback. To perform the safety analysis, the TASS/SMR (Transients And Setpoint Simulation/Small and Medium Reactor) code for a system response and SSF-1 correlation for a CHFR (Critical Heat Flux Ratio) have been used

  18. 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.)

  19. 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)

  20. Analysis of core physics and thermal-hydraulics results of control rod withdrawal experiments in the LOFT facility

    International Nuclear Information System (INIS)

    Varacalle, D.J. Jr.; Chen, T.H.; Harvego, E.A.; Ollikkala, H.

    1983-01-01

    Two anticipated transient experiments simulating an uncontrolled control rod withdrawal event in a pressurized water reactor (PWR) were conducted in the Loss-of-Fluid Test (LOFT) Facility at the Idaho National Engineering Laboratory. The scaled LOFT 50-MW(t) PWR includes most of the principal features of larger commercial PWRs. The experiments tested the ability of reactor analysis codes to accurately calculate core reactor physics and thermal-hydraulic phenomena in an integral reactor system. The initial conditions and scaled operating parameters for the experiments were representative of those expected in a commercial PWR. In both experiments, all four LOFT control rod assemblies were withdrawn at a reactor power of 37.5 MW and a system pressure of 14.8 MPa

  1. 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.)

  2. Control rod drive

    International Nuclear Information System (INIS)

    Hawke, B.C.

    1986-01-01

    A reactor core, one or more control rods, and a control rod drive are described for selectively inserting and withdrawing the one or more control rods into and from the reactor core, which consists of: a support structure secured beneath the reactor core; control rod positioning means supported by the support structure for movably supporting the control rod for movement between a lower position wherein the control rod is located substantially beneath the reactor core and an upper position wherein at least an upper portion of the control rod extends into the reactor core; transmission means; primary drive means connected with the control rod positioning means by the transmission means for positioning the control rod under normal operating conditions; emergency drive means for moving the control rod from the lower position to the upper position under emergency conditions, the emergency drive means including a weight movable between an upper and a lower position, means for movably supporting the weight, and means for transmitting gravitational force exerted on the weight to the control rod positioning means to move the control rod upwardly when the weight is pulled downwardly by gravity; the transmission means connecting the control rod positioning means with the emergency drive means so that the primary drive means effects movement of the weight and the control rod in opposite directions under normal conditions, thus providing counterbalancing to reduce the force required for upward movement of the control rod under normal conditions; and restraint means for restraining the fall of the weight under normal operating conditions and disengaging the primary drive means to release the weight under emergency conditions

  3. Control rod

    International Nuclear Information System (INIS)

    Igarashi, Takao; Yoshimoto, Yuichiro; Sugawara, Satoshi; Fukumoto, Takashi; Endo, Zen-ichiro; Saito, Shozo; Shinpo, Katsutoshi; Nishimura, Akira; Ozawa, Michihiro

    1988-01-01

    Purpose: To provide a sufficient shutdown margin upon reactor shutdown, prevent sheath deformation without decreasing neutron absorbents and prevent impact shocks exerted to structural materials. Constitution: The control rod of the present invention comprises a neutron absorption region, a sheath deformation means attached to the side wall and means for restricting and supporting axial movement of the neutron absorbent rod. Then, the amount of absorptive nuclei chained absorbents in the lower region is reduced than that in the upper region. In this way, effective neutron absorbing performance can be obtained relative to the neutron importance distribution during reactor shutdown. In addition, since the operationability is improved by reducing the weight of the control rod and the absorptive nuclei chained neutron abosrbers are used, mechanical nuclear life of the control rod can be increased. Thus, it is possible to prevent the outward deformation of the sheath, as well as prevent collision between the neutron absorber rod and the structural material on the side of inserting the control rod generated upon reactor scram by a simple structure. (Kamimura, M.)

  4. 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.)

  5. 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.

  6. PANTHER solution to the NEA-NSC 3-D PWR core transient benchmark. Uncontrolled withdrawal of control rods at zero power

    Energy Technology Data Exchange (ETDEWEB)

    Kuijper, J.C.

    1994-10-01

    This report contains the 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.).

  7. 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.)

  8. PANTHER solution to the NEA-NSC 3-D PWR core transient benchmark. Uncontrolled withdrawal of control rods at zero power

    International Nuclear Information System (INIS)

    Kuijper, J.C.

    1994-10-01

    This report contains the 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.)

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

    Energy Technology Data Exchange (ETDEWEB)

    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.).

  10. CONTROL ROD

    Science.gov (United States)

    Walker, D.E.; Matras, S.

    1963-04-30

    This patent shows a method of making a fuel or control rod for a nuclear reactor. Fuel or control material is placed within a tube and plugs of porous metal wool are inserted at both ends. The metal wool is then compacted and the tube compressed around it as by swaging, thereby making the plugs liquid- impervious but gas-pervious. (AEC)

  11. Control rods

    International Nuclear Information System (INIS)

    Koga, Isao; Masuoka, Ryuzo.

    1979-01-01

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

  12. Control rod

    International Nuclear Information System (INIS)

    Takeda, Toshikazu; Inoue, Kotaro.

    1979-01-01

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

  13. 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.)

  14. 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.)

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

    International Nuclear Information System (INIS)

    Kriventsev, Vladimir; Gabrielli, Fabrizio; Rineiski, Andrei

    2014-01-01

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

  16. Control rods

    International Nuclear Information System (INIS)

    Hirukawa, Koji.

    1979-01-01

    Purpose: To ensure the fuel safety by constituting a control rod with a plurality of poison bodies suspended in a cross-like section and shorter length poison bodies made movable and engageable in the gap between each of the above poison bodies thereby maintaining the function of the shorter length poison constant. Constitution: Cross-like supports are secured to the upper and lower parts of a driving shaft journaled in a sheath and poison bodies composed of neutron absorber poisons of a large thermal neutron absorption cross section and neutron absorber poison tubes for containing them are suspended from the supports. A movable cross-like support is mounted slidably at its base to the lower part of the driving shaft and poison bodies shorter than the above poison bodies and composed of neutron absorber poisons having a greater absorption cross section at the neutron energy region higher than thermal neutron region and neutron poison tubes for containing them are suspended to the movable support at the position capable of engaging in the gap between each of the poison bodies. (Kawakami, Y.)

  17. Control rod assembly

    International Nuclear Information System (INIS)

    Takahashi, Akio.

    1982-01-01

    Purpose: To enable reliable insertion and drops of control rods, as well as insure a sufficient flow rate of coolants flowing through the control rods for attaining satisfactory cooling thereof to enable relexation of thermal stress resulted to rectifying mechanisms or the likes. Constitution: To the outer circumference of a control rod contained vertically movably within a control rod guide tube, resistive members are retractably provided in such a way as to project to close the gap between outer circumference of the control rod and the inner surface of the control rod guide tube upon engagement of a gripper of control rod drives, and retract upon release of the engagement of the gripper. Thus, since the resistive members project to provide a greater resistance to the coolants flowing between them and the control rod guide tube in the normal operation where the gripper is engaged to drive the control rod by the control rod drives, a major part of the coolant flowing into the control rod guide tube flows into the control rod. This enables to cool the control rod effectively and make the temperature distribution uniform for the coolant flowing from the upper end of the control rod guide tube to thereby attain the relaxation of the thermal stress resulted in the rectifying mechanisms or the likes. (Moriyama, K.)

  18. 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

  19. 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.)

  20. Benchmark Analyses on the Control Rod Withdrawal Tests Performed During the PHÉNIX End-of-Life Experiments. Report of a Coordinated Research Project 2008–2011

    International Nuclear Information System (INIS)

    2014-06-01

    The IAEA supports Member State activities in advanced fast reactor technology development by providing a major fulcrum for information exchange and collaborative research programmes. The IAEA’s activities in this field are mainly carried out within the framework of the Technical Working Group on Fast Reactors (TWG-FR), which assists in the implementation of corresponding IAEA activities and ensures that all technical activities are in line with the expressed needs of Member States. In the broad range of activities, the IAEA proposes and establishes coordinated research projects (CRPs) aimed at improving Member States’ capabilities in fast reactor design and analysis. An important opportunity to conduct collaborative research activities was provided by the experimental campaign run by the French Alternative Energies and Atomic Energy Commission (CEA, Commissariat à l’énergie atomique et aux énergies alternatives) at the PHÉNIX, a prototype sodium cooled fast reactor. Before the definitive shutdown in 2009, end-of-life tests were conducted to gather additional experience on the operation of sodium cooled reactors. Thanks to the CEA opening the experiments to international cooperation, the IAEA decided in 2007 to launch the CRP entitled Control Rod Withdrawal and Sodium Natural Circulation Tests Performed during the PHÉNIX End-of-Life Experiments. The CRP, together with institutes from seven States, contributed to improving capabilities in sodium cooled fast reactor simulation through code verification and validation, with particular emphasis on temperature and power distribution calculations and the analysis of sodium natural circulation phenomena. The objective of this publication is to document the results and main achievements of the benchmark analyses on the control rod withdrawal test performed within the framework of the PHÉNIX end-of-life experimental campaign

  1. Benchmark Analyses on the Control Rod Withdrawal Tests Performed During the PHÉNIX End-of-Life Experiments. Report of a Coordinated Research Project 2008–2011

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2014-06-15

    The IAEA supports Member State activities in advanced fast reactor technology development by providing a major fulcrum for information exchange and collaborative research programmes. The IAEA’s activities in this field are mainly carried out within the framework of the Technical Working Group on Fast Reactors (TWG-FR), which assists in the implementation of corresponding IAEA activities and ensures that all technical activities are in line with the expressed needs of Member States. In the broad range of activities, the IAEA proposes and establishes coordinated research projects (CRPs) aimed at improving Member States’ capabilities in fast reactor design and analysis. An important opportunity to conduct collaborative research activities was provided by the experimental campaign run by the French Alternative Energies and Atomic Energy Commission (CEA, Commissariat à l’énergie atomique et aux énergies alternatives) at the PHÉNIX, a prototype sodium cooled fast reactor. Before the definitive shutdown in 2009, end-of-life tests were conducted to gather additional experience on the operation of sodium cooled reactors. Thanks to the CEA opening the experiments to international cooperation, the IAEA decided in 2007 to launch the CRP entitled Control Rod Withdrawal and Sodium Natural Circulation Tests Performed during the PHÉNIX End-of-Life Experiments. The CRP, together with institutes from seven States, contributed to improving capabilities in sodium cooled fast reactor simulation through code verification and validation, with particular emphasis on temperature and power distribution calculations and the analysis of sodium natural circulation phenomena. The objective of this publication is to document the results and main achievements of the benchmark analyses on the control rod withdrawal test performed within the framework of the PHÉNIX end-of-life experimental campaign.

  2. Control rod drive mechanism

    International Nuclear Information System (INIS)

    Nakamura, Akira.

    1981-01-01

    Purpose: To ensure the scram operation of a control rod by the reliable detection for the position of control rods. Constitution: A permanent magnet is provided to the lower portion of a connecting rod in engagement with a control rod and a tube having a plurality of lead switches arranged axially therein in a predetermined pitch is disposed outside of the control rod drives. When the control rod moves upwardly in the scram operation, the lead switches are closed successively upon passage of the permanent magnet to operate the electrical circuit provided by way of each of the lead switches. Thus, the position for the control rod during the scram can reliably be determined and the scram characteristic of the control rod can be recognized. (Furukawa, Y.)

  3. Fission reactor control rod

    International Nuclear Information System (INIS)

    Irie, Tomoo.

    1991-01-01

    The present invention concerns a control rod in a PWR type reactor. A control rod has an inner cladding tube and an outer cladding tube disposed coaxially, and a water draining hole is formed at the inside of the inner cladding tube. Neutron absorbers are filled in an annular gap between the outer cladding tube and the inner cladding tube. The water draining hole opens at the lower end thereof to the top end of the control rod and at the upper end thereof to the side of the upper end plug of the control rod. If the control rod is dropped to a control rod guide thimble for reactor scram, coolants from the control rod guide thimble are flown from the lower end of the water draining hole and discharged from the upper end passing through the water draining hole. In this way, water from the control rod guide thimble is removed easily when the control rod is dropped. Further, the discharging amount of water itself is reduced by the provision of the water draining hole. Accordingly, sufficient control rod dropping speed can be attained. (I.N.)

  4. Control rod drives

    International Nuclear Information System (INIS)

    Futatsugi, Masao.

    1980-01-01

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

  5. 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.)

  6. 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.)

  7. 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

  8. Control rod drive

    International Nuclear Information System (INIS)

    Okutani, Tetsuro.

    1988-01-01

    Purpose: To provide a simple and economical control rod drive using a control circuit requiring no pulse circuit. Constitution: Control rods in a BWR type reactor are driven by hydraulic pressure and inserted or withdrawn in the direction of applying the hydraulic pressure. The direction of the hydraulic pressure is controlled by a direction control valve. Since the driving for the control rod is extremely important in view of the operation, a self diagnosis function is disposed for rapid inspection of possible abnormality. In the present invention, two driving contacts are disposed each by one between the both ends of a solenoid valve of the direction control valve for driving the control rod and the driving power source, and diagnosis is conducted by alternately operating them. Therefore, since it is only necessary that the control circuit issues a driving instruction only to one of the two driving contacts, the pulse circuit is no more required. Further, since the control rod driving is conducted upon alignment of the two driving instructions, the reliability of the control rod drive can be improved. (Horiuchi, T.)

  9. 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

  10. Control rod drives

    International Nuclear Information System (INIS)

    Hayakawa, Hiroyasu; Kawamura, Atsuo.

    1979-01-01

    Purpose: To reduce pellet-clad mechanical interactions, as well as improve the fuel safety. Constitution: In the rod drive of a bwr type reactor, an electric motor operated upon intermittent input such as of pulse signals is connected to a control rod. A resolver for converting the rotational angle of the motor to electric signals is connected to the rotational shaft of the motor and the phase difference between the output signal from the resolver and a reference signal is adapted to detect by a comparator. Based on the detection result, the controller is actuated to control a motor for control rod drive so that fine control for the movement of the control rod is made possible. This can reduce the moving distance of the control rod, decrease the thermal stress applied to the control rod and decrease the pellet clad mechanical interaction failures due to thermal expansion between the cladding tube and the pellets caused by abrupt changes in the generated power. (Furukawa, Y.)

  11. Control rod driving mechanism

    International Nuclear Information System (INIS)

    Ooshima, Yoshio.

    1983-01-01

    Purpose: To perform reliable scram operation, even if abnormality should occur in a system instructing scram operation in FBR type reactors. Constitution: An aluminum alloy member to be melt at a predetermined temperature (about 600sup(o)C) is disposed to a connection part between a control rod and a driving mechanism, whereby the control rod is detached from the driving mechanism and gravitationally fallen to the reactor core. (Ikeda, J.)

  12. Detection device for control rod interference

    International Nuclear Information System (INIS)

    Saito, Noboru.

    1984-01-01

    Purpose: To enable to detect the mechanical interference or friction between a control rod and a channel box automatically, simply and rapidly. Constitution: A signal from a gate circuit and a signal from a comparison mechanism are inputted into an AND circuit if a control rod has not been displaced by a predetermined distance within a prescribed time Δt after the output of an insertion or withdrawal signal for the control rod, by which a control-rod-interference signal is outputted from the AND circuit. Accordingly, the interference between the control rod and the channel box can be detected automatically, easily and rapidly. Furthermore, by properly adjusting the prescribed time Δt set by the gate circuit, the degree of the interference can also be detected, whereby the safety and the reliability of the reactor can be improved significantly. (Horiuchi, T.)

  13. Control rod position control device

    International Nuclear Information System (INIS)

    Ubukata, Shinji.

    1997-01-01

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

  14. 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

  15. Control rod drives

    International Nuclear Information System (INIS)

    Asano, Hiromitsu.

    1979-01-01

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

  16. 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

  17. Control rod drive mechanism

    International Nuclear Information System (INIS)

    Mizuno, Katsuyuki.

    1976-01-01

    Object: To restrict the reduction in performance due to stress corrosion cracks by making use of condensate produced in a turbine steam condenser. Structure: Water produced in a turbine steam condenser is forced into a condensed water desalting unit by low pressure condensate pump. The condensate is purified and then forced by a high pressure condensate pump into a feedwater heater for heating before it is returned to the reactor by a feedwater pump. Part of the condensate issuing from the condensate desalting unit is branched from the remaining portion at a point upstream the pump and is withdrawn into a control rod drive water pump after passing through a motordriven bypass valve, an orifice and a condenser water level control valve, is pressurized in the control rod drive water desalting unit and supplied to a control rod drive water pressure system. The control rod is vertically moved by the valve operation of the water pressure system. Since water of high oxygen concentration does not enter during normal operation, it is possible to prevent the stress cracking of the stainless steel apparatus. (Nakamura, S.)

  18. 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

  19. Control rod drives

    International Nuclear Information System (INIS)

    Ikakura, Hiroaki.

    1986-01-01

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

  20. 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.)

  1. Control rod drive mechanism

    International Nuclear Information System (INIS)

    Futatsugi, Masao; Goto, Mikihiko.

    1976-01-01

    Purpose: To provide a control rod drive mechanism using water as an operating source, which prevents a phenomenon for forming two-layers of water in the neighbourhood of a return nozzle in a reactor to limit formation of excessive thermal stress to improve a safety. Constitution: In the control rod drive mechanism of the present invention, a heating device is installed in the neighbourhood of a pressure container for a reactor. This heating device is provided to heat return water in the reactor to a level equal to the temperature of reactor water thereby preventing a phenomenon for forming two-layers of water in the reactor. This limits formation of thermal stress in the return nozzle in the reactor. Accordingly, it is possible to minimize damages in the return nozzle portion and yet a possibility of failure in reactor water. (Kawakami, Y.)

  2. Control rod drives

    International Nuclear Information System (INIS)

    Yamanaka, Toshikatsu.

    1979-01-01

    Purpose: To protect bellows against failures due to negative pressure to prevent the loss of pressure balance caused by the expansion of the bellows upon scram. Constitution: An expansion pipe connected to the control rod drive is driven along a guide pipe to insert a control rod into the reactor core. Expansible bellows are provided at the step between the expansion pipe and the guide pipe. Further, a plurality of bore holes or slits are formed on the side wall of the guide pipe corresponding to the expansion portion of the bellows. In such an arrangement, when the expansion pipe falls rapidly and the bellows are expanded upon scram, the volume between each of the pipes of the bellows and the guide pipe is increased to produce a negative pressure, but the effect of the negative pressure on the bellows can be eliminated by the flowing-in of coolants corresponding to that pressure through the bore holes or the slits. (Furukawa, Y.)

  3. 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)

  4. Control rod drive shaft latch

    International Nuclear Information System (INIS)

    Thorp, A.G. II.

    1976-01-01

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

  5. Analysis of control rod behavior based on numerical simulation

    International Nuclear Information System (INIS)

    Ha, D. G.; Park, J. K.; Park, N. G.; Suh, J. M.; Jeon, K. L.

    2010-01-01

    The main function of a control rod is to control core reactivity change during operation associated with changes in power, coolant temperature, and dissolved boron concentration by the insertion and withdrawal of control rods from the fuel assemblies. In a scram, the control rod assemblies are released from the CRDMs (Control Rod Drive Mechanisms) and, due to gravity, drop rapidly into the fuel assemblies. The control rod insertion time during a scram must be within the time limits established by the overall core safety analysis. To assure the control rod operational functions, the guide thimbles shall not obstruct the insertion and withdrawal of the control rods or cause any damage to the fuel assembly. When fuel assembly bow occurs, it can affect both the operating performance and the core safety. In this study, the drag forces of the control rod are estimated by a numerical simulation to evaluate the guide tube bow effect on control rod withdrawal. The contact condition effects are also considered. A full scale 3D model is developed for the evaluation, and ANSYS - commercial numerical analysis code - is used for this numerical simulation. (authors)

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

    International Nuclear Information System (INIS)

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

    1999-10-01

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

  7. 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.)

  8. 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.)

  9. Control-rod driving mechanism

    International Nuclear Information System (INIS)

    Jodoi, Takashi.

    1976-01-01

    Purpose: To prevent falling of control rods due to malfunction. Constitution: The device of the present invention has a scram function in particular, and uses principally a fluid pressure as a scram accelerating means. The control rod is held by upper and lower holding devices, which are connected by a connecting mechanism. This connecting mechanism is designed to be detachable only at the lower limit of driving stroke of the control rod so that there occurs no erroneous scram resulting from careless disconnection of the connecting mechanism. Further, scramming operation due to own weight of the scram operating portion such as control rod driving shaft may be effected to increase freedom. (Kamimura, M.)

  10. Control rod drives

    International Nuclear Information System (INIS)

    Furumitsu, Yutaka.

    1981-01-01

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

  11. Reconstitutable control rod spider assembly

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  12. 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

  13. Control rod drive

    International Nuclear Information System (INIS)

    Kojima, Akira.

    1989-01-01

    In the control rod drive for a BWR type reactor, etc., according to this invention, the lower limit flow rate is set so as to keep the restriction for stability upon spectral shift operation. The setting condition for keeping the restriction is the lowest pump speed and the lower limit for the automatic control of the flow rate, which are considered to be important in view of the stablility from the actual power state. In view of the above, it is possible to keep the reactor core stably even in a case where such a transient phenomenon occurs that the recycling flow rate has to be run back to the lowest pump speed during spectral shift opeeration or in a case where the load demand is reduced and the flow rate is decreased by an automatic mode as in night operation. Accordingly, in the case of conducting the spectral shift operation according to this invention, the operation region capable of keeping the reactor core state stably during operation can be extended. (I.S.)

  14. 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

  15. 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.)

  16. 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

  17. 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.)

  18. Apparatus for decelerating the dropping speed of a control rod

    International Nuclear Information System (INIS)

    Shirakawa, Toshihisa.

    1975-01-01

    Object: To reduce the dropping speed (i.e. withdrawal) of a control rod of the upward insertion type in a BWR type reactor without reducing the speed of insertion. Structure: A control rod is provided with a flaring lower end so as to constitute a speed limiter which is penerated by vertically extending and upwardly open flow ducts that each have a narrow opening and flare upwardly. Thus, at the time of insertion of the control rod, the resistance offered thereto by the surrounding fluid is reduced to provide increased insertion speed. On the other hand, at the time of withdrawal the resistance offered by the fluid is increased to reduce the dropping speed of the control rod. (Ikeda, J.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-07-15

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

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  1. Control rod calibration including the rod coupling effect

    International Nuclear Information System (INIS)

    Szilard, R.; Nelson, G.W.

    1984-01-01

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

  2. 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.)

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

    International Nuclear Information System (INIS)

    Denizou, J.P.

    1989-01-01

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

  4. 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.)

  5. Advanced gray rod control assembly

    Science.gov (United States)

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

    2013-09-17

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

  6. The BWR Hybrid 4 control rod

    International Nuclear Information System (INIS)

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

    1988-01-01

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

  7. Control rod driving hydraulic device

    International Nuclear Information System (INIS)

    Sugano, Hiroshi.

    1993-01-01

    In a control rod driving hydraulic device for an improved BWR type reactor, a bypass pipeline is disposed being branched from a scram pipeline, and a control orifice and a throttle valve are interposed to the bypass pipeline for restricting pressure. Upon occurrence of scram, about 1/2 of water quantity flowing from an accumulator of a hydraulic control unit to the lower surface of a piston of control rod drives by way of a scram pipeline is controlled by the restricting orifice and the throttle valve, by which the water is discharged to a pump suction pipeline or other pipelines by way of the bypass pipeline. With such procedures, a function capable of simultaneously conducting scram for two control rod drives can be attained by one hydraulic control unit. Further, an excessive peak pressure generated by a water hammer phenomenon in the scram pipeline or the control rod drives upon occurrence of scram can be reduced. Deformation and failure due to the excessive peak pressure can be prevented, as well as vibrations and degradation of performance of relevant portions can be prevented. (N.H.)

  8. 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

  9. Control rod guide tube assemblies

    International Nuclear Information System (INIS)

    Jabsen, F.S.

    1979-01-01

    A nuclear fuel assembly including sleeves telescoped over end portions of control rod guide tubes which bear against internal shoulders of the sleeves. Upper ends of the sleeves protrude beyond a control rod guide tube spider and are locked in place by means of a resilient cellular lattice or lock that is seated in mating grooves in the outer surfaces of the sleeves. A grapple is provided for disengaging the entire lock structure spider and associated washers, springs and a grill from the end of the fuel assembly in order to enable these components to be removed and subsequently replaced on the fuel assembly after inspection and repair. (UK)

  10. NEUTRONIC REACTOR CONTROL ROD DRIVE APPARATUS

    Science.gov (United States)

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

    1959-12-15

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

  11. 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

  12. Measuring device for control rod driving time

    International Nuclear Information System (INIS)

    Tanaka, Kazuhiko; Hanabusa, Masatoshi.

    1993-01-01

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

  13. BWR control rod drive scram pilot valve monitoring system

    International Nuclear Information System (INIS)

    Soden, R.A.; Kelly, V.

    1984-01-01

    The control rod drive system in a Boiling Water Reactor is the most important safety system in the power plant. All components of the system can be verified except the solenoid operated, scram pilot valves without scramming a rod. The pilot valve mechancial works is the weak link to the control rod drive system. These pilot valves control the hydraulic system which applies pressure to the ''insert'' side of the control rod piston and vents the ''withdraw'' side of the piston causing the rods to insert during a scam. The only verification that the valve is operating properly is to scram the rod. The concern for this portion of the system is demonstrated by the high number of redundant components and complete periodic testing of the electrical circuits. The pilot valve can become hung-up through wear, fracture of internal components, mechanical binding, foreign material or chemicals left in the valve during maintenance, etc. If the valve becomes hung-up the electrical tests performed will not indicate this condition and scramming the rod is in jeopardy. Only an attempt to scram a rod will indicate the hung-up valve. While this condition exists the rod is considered inoperative. This paper describes a system developed at a nuclear power plant that monitors the pilot valves on the control rod drive system. This system utilizes pattern recognition to assure proper internal workings of the scram pilot valves to plant operators. The system is totally automatic such that each time the valve is operated on a ''half scram'', a printout is available to the operator along with light indication that each of the 370 valves (on one unit of a BWR) is operating properly. With this monitoring system installed, all components of the control rod drive system including the solenoid pilot valves can be verified as operational without scramming any rods

  14. BWR control rod drive scram pilot valve monitoring program

    International Nuclear Information System (INIS)

    Soden, R.A.; Kelly, V.

    1986-01-01

    The control rod drive system in a Boiling Water Reactor is the most important safety system in the power plant. All components of the system can be verified except the solenoid operated, scram pilot valves without scramming a rod. The pilot valve mechanical works is the weak link to the control rod drive system. These pilot valves control the hydraulic system which applies pressure to the insert side of the control rod piston and vents the withdraw side of the piston causing the rods to insert during a scram. The only verification that the valve is operating properly is to scram the rod. The concern for this portion of the system is demonstrated by the high number of redundant components and complete periodic testing of the electrical circuits. The pilot valve can become hung-up through wear, fracture of internal components, mechanical binding, foreign material or chemicals left in the valve during maintenance, etc. If the valve becomes hung-up the electrical tests performed will not indicate this condition and scramming the rod is in jeopardy. Only an attempt to scram a rod will indicate the hung-up valve. While this condition exists the rod is considered inoperative. This paper describes a system developed at a nuclear power plant that monitors the pilot valves on the control rod drive system. This system utilizes pattern recognition to assure proper internal workings of the scram pilot valves to plant operators. The system is totally automatic such that each time the valve is operated on a half scram, a printout is available to the operator along with light indication that each of the 370 valves (on one unit of a BWR) is operating properly. With this monitoring system installed, all components of the control rod drive system including the solenoid pilot valves can be verified as operational without scramming any rods

  15. Control rod repositioning considerations in core design analysis

    International Nuclear Information System (INIS)

    Armstrong, B.C.; Buechel, R.J.

    1990-01-01

    Control rod repositioning is a method for minimizing rod cluster control assembly (RCCA) wear in the upper internals area where the guide cards interface with the rodlets of the RCCAs. A number of utilities have implemented strategies for rod repositioning, which often has no impact on the nuclear analysis for cases where the control rods are never repositioned into the active fuel. Other strategies involve repositioning the control rods several steps into the active fuel. The impact of this type of repositioning on the axial power shape and consequently the total peaking factor F Q T varies, depending on the method in which the repositioning is implemented at the plant. Operating for long periods with all the control and shutdown rods inserted several steps in the active fuel followed by withdrawing them fully from the core results in a shifting of the power distribution toward the top of the core and must be accounted for in the design analysis. On the other hand, an optional plan for control rod repositioning that considers margins available in related design parameters can be devised that minimizes the effects of the repositioning for the reload. This paper summarizes a rod repositioning strategy implemented for a recent reload and some calculated power shape results for this strategy and other scenarios

  16. Control rod drive for vertical movement

    International Nuclear Information System (INIS)

    Suskov, I.I.; Gorjunov, V.S.; Zajcev, B.I.; Derevjankin, N.E.; Petrov, V.A.; Istomin, S.D.; Kovalencik, D.I.; Archipov, E.A.; Serebrjakov, V.I.; Kacalin, V.S.

    1982-01-01

    The control of the rod repositioning gear unit and the control unit of the profile grab of the control rod drive for the alkali metal-cooled fast breeder reactor is achieved by an electromotor being arranged outside the hermetic drive casing. The guide tube is directly repositioned by the rod repositioning gear unit. Coupling control of the drive with the control rod is done in the lower operative position of the control rod and that because of the interaction of the tie rod arranged on the spring-mounted control rod with the induction transmitter for the lower position of the control rod. In the transfer position the rod is fixed within the guide tube. (orig.)

  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. Snubber assembly for a control rod drive

    International Nuclear Information System (INIS)

    1976-01-01

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

  2. Fuel element cellular grid structure and procedure to insert and withdraw fuel rods from that structure

    International Nuclear Information System (INIS)

    1975-01-01

    A typical embodiment of the invention provides a means for selectively inserting and withdrawing one or more fuel rods from a fuel element cellular grid structure. The transverse stubs on one side of a long, thin bar are turned through 90deg to extend across the gap between mutually perpendicular grid structure plates. The extreme ends of these stubs engage the adhacent portions of the associated plates that form part of the grid cells. Pressing the stubs against the plate portions through the application of appropriate force in a longitudinal direction relative to the bar deflects the engaged plates through a sufficient distance to enable fuel rods to be inserted into or withdrawn from respective cells. After rod insertion, the force applied to the bar is released to enable the plates to relax and engage the fuel rods. The bars are rotated once more through 90deg and withdrawn from the grid structure. A similar procedure is employed to withdraw fuel rods from the grid structure

  3. Fuel followed control rod installation at AFRRI

    International Nuclear Information System (INIS)

    Moore, Mark; Owens, Chris; Forsbacka, Matt

    1992-01-01

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

  4. Control rods in LMFBRs: a physics assessment

    International Nuclear Information System (INIS)

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

    1982-08-01

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

  5. 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)

  6. Rope wind-up type control rod

    International Nuclear Information System (INIS)

    Tsuji, Teruaki; Watanabe, Shigeru.

    1979-01-01

    Purpose: To hold a control rod at a certain position even if the sealed cover of the rod drive mechanism should fail. Constitution: A plurality of friction plates, engaging wheels and a threaded shaft are provided to the wind-up drum for winding up a rope which moves the control rod up and down. While the control rod is adapted to drop by its own weight upon insertion, it is adapted to stop at a predetermined position exactly with no shocks by gradually increasing braking force by the sliding friction caused from the friction plates or the like. A ratch mechanism is provided to the upper portion of the control rod so that the top of the ratch piece may automatically engage the guide passage wall of the control rod upon uncontrolled running of the control rod to prevent further uncontrolled running thereof. (Ikeda, J.)

  7. Accident-tolerant control rod

    International Nuclear Information System (INIS)

    Ohta, Hirokazu; Sawabe, Takashi; Ogata, Takanari

    2013-01-01

    Boron carbide (B 4 C) and hafnium (Hf) metal are used for the neutron absorber materials of control rods in BWRs, and silver-indium-cadmium (Ag-In-Cd) alloy is used in PWRs. These materials are clad with stainless steel. The eutectic point of B 4 C and iron (Fe) is about 1150 deg. C and the melting point of Ag-In-Cd alloy is about 800 deg. C, which are lower than the temperature of zircaloy - steam reaction increases rapidly (∼1200 deg. C). Accordingly, it is possible that the control rods melt and collapse before the reactor core is significantly damaged in the case of severe accidents. Since the neutron absorber would be separated from the fuels, there is a risk of re-criticality, when pure water or seawater is injected for emergency cooling. In order to ensure sub-criticality and extend options of emergency cooling in the course of severe accidents, a concept of accident-tolerant control rod (ACT) has been derived. ACT utilises a new absorber material having the following properties: - higher neutron absorption than current control rod; - higher melting or eutectic temperature than 1200 deg. C where rapid zircaloy oxidation occurs; - high miscibility with molten fuel materials. The candidate of a new absorber material for ATC includes gadolinia (Gd 2 O 3 ), samaria (Sm 2 O 3 ), europia (Eu 2 O 3 ), dysprosia (Dy 2 O 3 ), hafnia (HfO 2 ). The melting point of these materials and the liquefaction temperature with Fe are higher than the rapid zircaloy oxidation temperature. ACT will not collapse before the core melt-down. After the core melt-down, the absorber material will be mixed with molten fuel material. The current absorber materials, such as B 4 C, Hf and Ag-In-Cd, are charged at the tip of ATC in which the neutron flux is high, and a new absorber material is charged in the low-flux region. This design could minimise the degradation of a new absorber material by the neutron absorption and the influence of ATC deployment on reactor control procedure. As a

  8. 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.)

  9. 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.)

  10. 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.)

  11. Control rod guide tube assembly

    International Nuclear Information System (INIS)

    Jabsen, F.S.

    1982-01-01

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

  12. Seismic scrammability of HTTR control rods

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  13. Catawba-1 rod withdrawal event and analysis with SIMULATE-3K

    International Nuclear Information System (INIS)

    Thomas, S.B.; Borkowski, J.A.; Smith, K.S.

    1995-01-01

    On March 23, 1995, Catawba unit 1 was in startup physics testing for cycle 9. During a rod-swap measurement between the reference bank shutdown bank B (SB) and shutdown bank A (SA) at hot zero power (HZP), an erroneous reactivity meter indication led to excessive withdrawal of SB. Before operator action returned the plant to the initial condition, power was measured on the ex-core detectors and the loop thermocouples. The rod motion transient with feedback was analyzed with the SIMULATE-3K code, using both best-estimate and conservative assumptions. The purposes of the analysis were to determine the calculated in-core power and reactivity response during the transient, to assess the implications of actual and postulated operator actions, and to assess the performance of SIMULATE-3K against measured data

  14. 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

  15. Development of a control rod drive

    International Nuclear Information System (INIS)

    1991-01-01

    In the period under review, the computer codes required for transients calculation have been completed, as well as the programs for modelling and testing the hot-gas temperature control by means of combined core rod and reflector rod operation. The specification of requirements to be fulfilled by the rod drive computer and the neutron flux measuring system has been done relying essentially on the data obtained by the transients calculations performed and the resulting informations on operating conditions. The work for optimization of the core rod drive with regard to rod driving speeds and the 'three-point switch' with hysteresis for controlled, automatic core rod operation has been concentrating on the case of specified, normal operation of the reactor. (orig./DG) [de

  16. 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.)

  17. Control rod for HTGR type reactor

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  18. Position control device for a control rod

    International Nuclear Information System (INIS)

    Ono, Takehiko; Kusaka, Shuji.

    1976-01-01

    Purpose: To reliably prevent dangerous operation in the control of the position of the control rod by checking for abnormal pulse motor coil excitation voltage and, at the time of occurrence of abnormality, immediately holding the control rod stationary lest it should be moved to an unsafe position, this being accomplished excitation from a compensating excitation system. Constitution: In an FBR reactor, a circuit for memorizing the correct output states of individual drive signals at arbitrary instants and consequtively producing the memorized results is provided, and the output of the circuit and the actual drive signal are compared at all times to discriminate whether the drive signal being compared is normal or not. When the actual drive signal is abnormal, a series signal varying after a predetermined pattern is shifted to enable replacement of the actual drive signal, so that irrespective of whether the problem drive signal is ''on'' or ''off'', a drive signal of the correct pattern may be supplied to the pulse motor to hold the control rod and prevent it from being moved toward the dangerous side due to its own weight or other causes. (Horiuchi, T.)

  19. RODMOD: a code for control rod positioning

    International Nuclear Information System (INIS)

    Vondy, D.R.; Fowler, T.B.

    1978-11-01

    The report documents a computer code which has been implemented to position control rods according to a prescribed schedule during the calculation of a reactor history. Control rods may be represented explicitly with or without internal black absorber conditions in selected energy groups, or fractional insertion may be done, or both, in a problem. There is provision for control rod follower, movement of materials through a series of zones in a closed loop, and shutdown rod insertion and subsequent removal to allow the reactor history calculation to be continued. This code is incorporated in the system containing the VENTURE diffusion theory neutronics and the BURNER exposure codes for routine use. The implemented automated procedures cause the prescribed control rod insertion schedule to be applied without the access of additional user input data during the calculation of a reactor operating history

  20. Microcomputer system for controlling fuel rod length

    International Nuclear Information System (INIS)

    Meyer, E.R.; Bouldin, D.W.; Bolfing, B.J.

    1979-01-01

    A system is being developed at the Oak Ridge National Laboratory (ORNL) to automatically measure and control the length of fuel rods for use in a high temperature gas-cooled reactor (HTGR). The system utilizes an LSI-11 microcomputer for monitoring fuel rod length and for adjusting the primary factor affecting length. Preliminary results indicate that the automated system can maintain fuel rod length within the specified limits of 1.940 +- 0.040 in. This system provides quality control documentation and eliminates the dependence of the current fuel rod molding process on manual length control. In addition, the microcomputer system is compatible with planned efforts to extend control to fuel rod fissile and fertile material contents

  1. 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)

  2. 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.

  3. 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

  4. Control rod housing alignment and repair method

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  5. 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

  6. French LMFBR's control rods experience and development

    International Nuclear Information System (INIS)

    Arnaud, G.; Guigon, A.; Verset, L.

    1983-06-01

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

  7. 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

  8. Control rod position detector for nuclear reactor

    International Nuclear Information System (INIS)

    Kudo, Mitsuru; Fujiwara, Hiroshi.

    1981-01-01

    Purpose: To improve the reliability of a control rod position detector by detecting a reactive code with a combination of control rod position change signals produced from vertical and horizontal axis decoders, generation an error signal and thus simultaneously detecting the operation of more than two lead switches. Constitution: Horizontal and vertical axis position signals responsive to changes in the control rod position are applied from lead switches connected in a predetermined matrix connection corresponding to the notches of the positions of respective position detecting probes, the reactive output from the decoder is detected by a reactive code detecting circuit, which in turn generates a fault signal, and the control rod position code converted in a notch number generating circuit is converted to a predetermined value indicating invalidity. Accordingly, a fault caused by the simultaneous operation of a plurality of failed lead switches can be effectively detected. (Yoshino, Y.)

  9. Control rod for a nuclear reactor

    Science.gov (United States)

    Roman, Walter G.; Sutton, Jr., Harry G.

    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.

  10. 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

  11. Detection device for control rod scram

    International Nuclear Information System (INIS)

    Ishiyama, Satoshi.

    1989-01-01

    The device of the present invention comprises a control rod dropping separately from a control rod driving mechanism main body, a following tube falling separately accompanying therewith and a guide tube for guiding the dropping of the control rod and the following tube. Further, rare earth permanent magnets are embedded with the pole being axially oriented in the following tube and bobbins each mounted with an inner flange made of high magnetic permeability material are disposed to the guide tube. Coils are wound in the bobbin. In this control rod scram detection device, since magnetic fluxes can effectively be supplied to the coils, it is possible to obtain stable and highly reliable scram detection signals. Further, since the coils and the bobbins can be manufactured separately from the guide tube, their assemblies can be tested independently from the guide tube. (K.M.)

  12. 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.

  13. 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.)

  14. Control rod supporting device in reactor

    International Nuclear Information System (INIS)

    Seki, Osamu; Itooka, Satoshi; Harada, Kiyoshi; Jodoi, Takashi.

    1990-01-01

    Since coolants flowing from a reactor core hit against a control rod and a control rod connection pipe, a considerable amount of bending moment for separating an attracting surface between an electromagnet and an armature is formed. Then, a plurality of grooves are formed on a heat sensitive material to dispose a heat collecting fin, and each of upper and lower contact portions of a control rod supporting portion in which the flanged portion of T-like cross section does not slip out is made into a partial spheric surface and a portion between the electromagnet and the attracted member are engaged by the unevenness. With such a constitution, even if a bending moment is applied, the control rod only swings and the bending moment is not transmitted to the attracted member. Further, since the temperature of the heat sensitive material can be rapidly made closer to the peripheral temperature by using the heat collecting fin, the timing for separation is made accurate. Further, since the engaging portion is brought into contact at the spheric surface, the load distribution on the control rod is made uniform, and the positional relationship is made accurate, to support the control rod reliably and the separation depends only on the temperature of the coolants. (N.H.)

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

    International Nuclear Information System (INIS)

    Sudirdjo, Hari; Setyono; Prasetya, Hendra

    2001-01-01

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

  16. Stabilizing device for control rod tip

    International Nuclear Information System (INIS)

    Verdone, G.F.

    1982-01-01

    A control rod has a spring device on its lower end for eliminating oscillatory contact of the rod against its adjacent guide tube wall. The base of the device is connected to the lower tip of the rod. A plurality of elongated extensions are cantilevered downward from the base. Each extension has a shoulder for contacting the guide tube, and the plurality of shoulders as a group has a transverse dimension that is preset to be larger than the inner diameter of the guide tube such that an interference fit is obtained when the control rod is inserted in the tube. The elongated extensions form an open-ended, substantially hollow member through which most of the liquid coolant flows, and the spaces between adjacent extensions allow the flow to bypass the shoulders without experiencing a significant pressure drop

  17. 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.)

  18. Control rod drives for HTGR type reactor

    International Nuclear Information System (INIS)

    Nishiguchi, Isoharu; Katagiri, Shigeo.

    1991-01-01

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

  19. Control rod drives for FBR type reactor

    International Nuclear Information System (INIS)

    Ikakura, Hiroaki.

    1990-01-01

    The control rod drives for an FBR type reactor of the present invention eliminate obstacles deposited on attracting surfaces between an electromagnet and an armature which connect control rods to recover their retaining power. That is, a sealed chamber capable of controlling its inner pressure by an operation from the outside of a reactor is disposed in an extension pipe, and a nozzle connected to the sealed chamber and facing at the lower end thereof to the attracting surface is disposed. Liquid sodium sucked by evacuating the sealed chamber is jetted out from the nozzle by pressurizing the chamber to simultaneously eliminate obstacles deposited to the attracting surfaces of the electromagnet and the control rod. Alternatively, a nozzle protruding from and retracting to the lower surface of the electromagnet is disposed opposing to each of the attracting surfaces of the electromagnet and the control rod. Similar effect can also be obtained if gases are jetted out in this state. As a result, control rod drives of high reliability for a FBR type reactor can be obtained. (I.S.)

  20. Protective guide structure for reactor control rod

    International Nuclear Information System (INIS)

    Ban, Minoru; Umeda, Kenji; Kubo, Noboru; Ito, Tomohiro.

    1996-01-01

    The present invention provides an improved protective guide structure for control rods, which does not cause swirling of coolants and resonance even though a slit is formed on a protective tube which surrounds a control rod element in a PWR type reactor. Namely, a reactor control rod is constituted with elongated control elements collectively bundled in the form of a cluster. The protective guide structure protectively guides the collected constituent at the upper portion of a reactor container. The protective structure comprises a plurality of protective tubes each having a C-shaped cross section disposed in parallel for receiving control rod elements individually in which the corners of the opening of the cross section of the protective tube are chamfered to an appropriate configuration. With such a constitution, even if coolant flows in a circumferential direction along the protective tubes surrounding the control rod elements, no shearing stream is caused to the coolants flow since the corners of the cross sectional opening (slit) of the tube are chamfered. Accordingly, occurrence of swirlings can be suppressed. (I.S.)

  1. 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

  2. Regulatory perspective on incomplete control rod insertions

    International Nuclear Information System (INIS)

    Chatterton, M.

    1997-01-01

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

  3. Method of driving control rod in reactor

    International Nuclear Information System (INIS)

    Osa, Hirotaka.

    1986-01-01

    Purpose: To improve security and safety of the reactor by reducing reactor output automatically and quickly when circulation of cooling water is stopped. Constitution: When the circulating pump is under operation, fluid pressure in the discharge pipe is transferred to the fluid room of fluid pressure cylinder via the control rod drive pipe and lift up the piston, and then the control rod is drawn out of the reactor core. When the circulating pump is lowered in its functions, discharge pipe fluid pressure decreases, fluid pressure in the fluid room decreases, and with less force of piston movement, the control rod gets lowered by its own weight. At this time, the blocked state of the opening by the piston is released, fluid flows into the room. Lowering of pressure and the control rod is promoted by transferring out fluid below the piston in the fluid room to the upper part of the piston via a small gap when the control rod falls by gravity. (Horiuchi, T.)

  4. Radiological characterization of spent control rod assemblies

    International Nuclear Information System (INIS)

    Lepel, E.A.; Robertson, D.E.; Thomas, C.W.; Pratt, S.L.; Haggard, D.L.

    1995-10-01

    This document represents the final report of an ongoing study to provide radiological characterizations, classifications, and assessments in support of the decommissioning of nuclear power stations. This report describes the results of non-destructive and laboratory radionuclide measurements, as well as waste classification assessments, of BWR and PWR spent control rod assemblies. The radionuclide inventories of these spent control rods were determined by three separate methodologies, including (1) direct assay techniques, (2) calculational techniques, and (3) by sampling and laboratory radiochemical analyses. For the BWR control rod blade (CRB) and PWR burnable poison rod assembly (BPRA), 60 Co and 63 Ni, present in the stainless steel cladding, were the most abundant neutron activation products. The most abundant radionuclide in the PWR rod cluster control assembly (RCCA) was 108m Ag (130 yr halflife) produced in the Ag-In-Cd alloy used as the neutron poison. This radionuclide will be the dominant contributor to the gamma dose rate for many hundreds of years. The results of the direct assay methods agree very well (±10%) with the sampling/radiochemical measurements. The results of the calculational methods agreed fairly well with the empirical measurements for the BPRA, but often varied by a factor of 5 to 10 for the CRB and the RCCA assemblies. If concentration averaging and encapsulation, as allowed by 10CFR61.55, is performed, then each of the entire control assemblies would be classified as Class C low-level radioactive waste

  5. Overview of Japanese control rods development program

    International Nuclear Information System (INIS)

    Koyama, M.

    1984-01-01

    The Japanese control rods development program was established based on the fast breeder reactor program. Therefore, PNC's efforts have been made mainly for the development of analysis, design and fabrication technologies for ''JOYO'' and ''MONJU'' control rods. Laboratory studies were performed to obtain the information for absorber materials. The design and fabrication of the sealed and vented type control rod pins were completed, and water loop tests and in-sodium tests were carried out. Irradiation behavior of enriched B 4 C pellets with low and high density in DFR was examined. Japan's experimental fast reactor, JOYO, has been operated at the rated power of 50MWt and 75MWt since April 1977 when the MK-I core (breeder core) attained initial criticality. Post irradiation examinations on control rod, removed from the reactor, were carried out and their performance behavior were evaluated. In the MK-II core, a control rods monitoring program has been in investigation. Absorber Materials Irradiation Rigs (AMIR) are scheduled to be loaded and irradiated in the JOYO MK-II core from 1984. (author)

  6. A nuclear reactor with buffered control rods

    International Nuclear Information System (INIS)

    Bevilacqua, F.

    1974-01-01

    The control rods for, e.g., water-cooled reactors are fastened as units on common crossbars in vertical downward direction. The fastening on the crossbar is achieved by means of cross-shaped parts, e.g., in the shape of a double 'H'. A cylinder connected with a drive rod in normal operation is joined to each of the crossbars. In an emergency shut-down, this connection is interrupted and the control rod unit drops into the core through the action of gravity. Its fall is slowed down by a cushion or shock absorbing unit. For this purpose a piston is provided mounted on the supporting plate below the cylinder and guided within it. In the cylinder, the coolant is contained as damping medium. An upper opening in the cylinder serves as a ventilation hole. The movement of the piston is limited by a stopping part within the cylinder and slowed down by a spiral spring. (DG) [de

  7. 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.)

  8. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-07-01

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

  10. Control rod drives for nuclear reactors

    International Nuclear Information System (INIS)

    Okada, Shige.

    1981-01-01

    Purpose: To enable the detection for the raptures of the bellows in control rod drives in LMFBR type reactor, by recycling seal gases inside the bellows and measuring the radioactivity in the recycling passage. Constitution: In the control drives, outer extension pipe is surrounded by the bellows, which is put between the cylindrical biological shieldings around the upper potion of an upper guide tube and the disk-like seal members provided at the lower flange of the outer extension pipe. Thus, the inside device of control rod is isolated from the coolants and the cover gases. The outer extension pipe is provided with a suction channel and a return channel. These channels are connected to a seal gas recycling pipeway, a pump and a radioactivity detector, where the seal gases in the bellows is recycling. If failures should occur in the bellows, cover gas leaks into the seal gas and recycles, whereby radioactivity is detected and alarmed. (J.P.N.)

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

    International Nuclear Information System (INIS)

    Magnusson, Per

    2008-01-01

    This work has investigated the Control Rod Worth (CRW) and Nuclear End of Life (NEOL) values for BWR control rods. A study of how different parameters affect NEOL was performed with the transport code PHOENIX4. It was found that NEOL, expressed in terms of 10 B depletion, can be generalized beyond the conditions for which the rod is depleted, such as different power densities and void fractions, the corresponding variation in the NEOL will be about 0.2-0.4% 10 B. It was also found that NEOL results for different fuel types and different fuel enrichments have a variation of about 2-3% in 10 B depletion. A comparative study on NHOL and CRW was made between PHOENIX4 and the stochastic Monte Carlo code MCNP. It was found that there is a significant difference, both due to differences in the codes and to limitations in the geometrical modeling in PHOENIX4. Since MCNP is considered more physically correct, a methodology was developed to calculate the nuclear end of life of BWR control rods with MCNP. The advantages of the methodology are that it does not require other codes to perform the depletion of the absorber material, it can describe control rods of any design and it can deplete the control rod absorber material without burning the fuel. The disadvantage of the method is that is it time-consuming

  12. 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.)

  13. Control rod guide tube cleaning device

    International Nuclear Information System (INIS)

    Tsuji, Tadashi; Shiota, Yoshiaki.

    1990-01-01

    Since there was no exclusive device for cleaning control rods, no effective cleaning could not be conducted and there was a possibility that obstacles may not be recovered. Then, there are disposed a first pump for supplying pressurized water, a spray nozzle for forming a swirling flow in a control rod guide tube, a second pump for pressurizing water introduced by a sucking pipeline and a collecting device for recovering obstacles intruding to water from the second pump. The pressurized water supplied from the first pump is introduced to a head passing through a blowing pipe and jetted from the spray nozzle to the control rod guide tube. In this case, a swirling stream occurs and obstacles in the control guide tube are mixed into water. The water containing the obstacles passes from the sucking port through a pipeline, introduced to the second pump and recovered to the collecting device. Since there is no water staying portion upon cleaning operation, the obstacles accumulating over the entire region of the bottom of the guide tube can be recovered reliably and efficiently. (N.H.)

  14. The lifetime of the control rod drives

    International Nuclear Information System (INIS)

    Avet, B.; Cauquelin, C.

    1989-01-01

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

  15. Passive cooling of control rod drive mechanisms

    International Nuclear Information System (INIS)

    Hankinson, M.F.; Schwirian, R.E.

    1992-01-01

    A method and apparatus are provided for passively cooling the control rod drive mechanisms (CRDMs) in the reactor vessel of a nuclear power plant. Passive cooling is achieved by dispersing a plurality of chimneys within the CRDM array in positions where a control rod is not required. The chimneys induce convective air currents which cause ambient air from within the containment to flow over the CRDM coils. The air heated by the coils is guided into inlets in the chimneys by baffles. The chimney is insulated and extends through the seismic support platform and missile shield disposed above the closure head. A collar of adjustable height mates with plate elements formed at the distal end of the CRDM pressure housings by an interlocking arrangement so that the seismic support platform provides lateral restraint for the chimneys. (Author)

  16. 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)

  17. Magnetic switch for reactor control rod

    International Nuclear Information System (INIS)

    Germer, J.H.

    1986-01-01

    This patent describes a control rod system for a nuclear reactor utilizing an electromagnetic grapple mechanism for holding and releasing a control rod, the improvement comprising a magnetic reed switch assembly having a Curie-point magnetic shunt and responsive to reactor coolant temperature for short circuiting the electromagnetic grapple mechanism causing release of the control rod when the coolant temperature reaches the Curie-point of the magnetic shunt. The magnetic reed switch assembly includes a: a permanent magnet, a pair of magnetic pole pieces located at and in contact with opposite ends of the permanent magnet, the Curie-point magnetic shunt being positioned adjacent the permanent magnet and in contact with the pair of magnetic pole pieces, and a reed switch positioned intermediate the pole pieces and provided with a pair of ferromagnetic reeds, a nonmagnetic enclosure around the reeds, a first of the reeds being secured at one end to a first of the pair of pole pieces, a second of the reeds having one end extending into and secured to a hollow member positioned in and extending through a second of the pair of pole pieces, the one end of the second of the reeds secured to a condector adapted to be connected to the electromagnetic grapple mechanism

  18. 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

  19. Verification test of control rod system for HTR-10

    International Nuclear Information System (INIS)

    Zhou Huizhong; Diao Xingzhong; Huang Zhiyong; Cao Li; Yang Nianzu

    2002-01-01

    There are 10 sets of control rods and driving devices in 10 MW High Temperature Gas-cooled Test Reactor (HTR-10). The control rod system is the controlling and shutdown system of HTR-10, which is designed for reactor criticality, operation, and shutdown. In order to guarantee technical feasibility, a series of verification tests were performed, including room temperature test, thermal test, test after control rod system installed in HTR-10, and test of control rod system before HTR-10 first criticality. All the tests data showed that driving devices working well, control rods running smoothly up and down, random position settling well, and exactly position indicating

  20. 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

  1. Installing and detaching apparatus for a control rod drive mechanism

    International Nuclear Information System (INIS)

    Akimoto, Seiichi; Watanabe, Mitsuhiro; Yoshida, Tomiharu; Sugaya, Jun-ichi; Saito, Takashi.

    1976-01-01

    Object: To facilitate maintenance and repair of a control rod drive mechanism. Structure: The apparatus comprises a means moving in a moving direction of a control rod within a reactor vessel, said moving means having a housing mounted thereon, a means mounted on the reactor vessel to release a connection between a control rod drive mechanism connected to the control rod and the control rod, and a means for mounting and removing a fixing means which connects the reactor vessel to the control rod drive means. With this arrangement, cooling water of high radioactivity level may not be leaked outside to thereby notably reduce dangerousness of exposure and materially cut time required for mounting and removing the control rod drive mechanism. (Ohara, T.)

  2. Safety analysis of control rod drive computers

    International Nuclear Information System (INIS)

    Ehrenberger, W.; Rauch, G.; Schmeil, U.; Maertz, J.; Mainka, E.U.; Nordland, O.; Gloee, G.

    1985-01-01

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

  3. 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

  4. 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

  5. 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.)

  6. 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

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

    International Nuclear Information System (INIS)

    Becvar, J.; Saroch, V.

    1977-01-01

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

  8. Experiment on thermohydraulics of simulated control rod

    International Nuclear Information System (INIS)

    Ogawa, Masuro; Ouchi, Mitsuo; Akino, Norio; Fujimura, Kaoru; Shiina, Yasuaki; Kawamura, Hiroshi

    1984-10-01

    A thermohydraulic study of a control rod channel is required for the core design of the Very High Temperature Gas Cooled Reactor (VHTR). A non-heating experiment with air flow was performed prior to heating experiment with helium flow. Experimental results on stability of flow, flow rate distribution and pressure drop of the control rod channel are reported. In a test section of the experimental apparatus, five simulated control subrods were suspended vertically in a circular duct. Their dimension was in coincide with those of the Detailed Disign (I) of the VHTR. Air of atomospheric pressure was used as a coolant gas, which flowed in inner and outer paths of the subrods. Total flow rate ranged from 0.0011 to 0.0062 kg/s. Flow rate distribution and pressure drop were obtained for various flow rates. Velocity fluctuation in the channel was also observed using a hot wire anemometer. From these experiments, it was found that the flow rate distribution was nearly the same as a disigned value and that turbulent and laminar flows were simultaneously realized in outer and inner paths respectively. These observations supported a feasibility of the present design. (author)

  9. 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.)

  10. 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

  11. Control rod guide tube of nuclear reactor

    International Nuclear Information System (INIS)

    Suda, Yoshitaka; Ito, Kenji; Matsumoto, Kunio.

    1994-01-01

    Zr having a residual tensile stress of 3 to 10kg/mm 2 in a circumferential direction is used for the main ingredient of a control guide tube of a nuclear reactor. For this purpose, an appropriate correction method such as a roll-correction, tension-correction and press-correction method is applied to an existent Zr-base alloy tube with no substantial residual stress. If the residual tensile stress in the circumferential direction is smaller than 3kg/mm 2 , an effect sufficient to suppress irradiation growth is not obtainable, if it exceeds 10kg/mm 2 , dimensional changes, cracks or the like occurs locally since the wall thickness of the control rod guide tube is small and, accordingly, this often results in failed products as the control guide tube. (N.H.)

  12. 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)

  13. 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

  14. Development and testing of control rod drives for ship reactors

    International Nuclear Information System (INIS)

    Bruelheide, K.; Mundt, D.; Peters, C.-H.; Manthey, H.-J.

    1978-01-01

    The following paper deals with the development and testings of a new control rod drive design for marine reactors. Starting from the good operating experience with the advanced pressurized water reactor (FDR) of the NS OTTO HAHN a control rod drive system with an hermetically sealed drive principle was developed. A prototype control rod drive system was put through extensive tests and developed ready for standard production at the 'Gesellschaft fuer Kernenergieverwertung in Schiffbau und Schiffahrt'

  15. The analytic method for calculating the control rod worth

    International Nuclear Information System (INIS)

    Kim, Han Gon; Lee, Byeong Ho; Chang, Soon Heung

    1989-01-01

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

  16. Analysis of control rod worth in experimental fast reactor JOYO

    International Nuclear Information System (INIS)

    Arii, Y.; Aoyama, T.; Okimoto, Y.; Yoshida, A.; Mizoo, N.

    1988-01-01

    In JOYO, the measurement of control rod worths have been carried out in the beginning of the each cycle, using both period method and neutron source multiplication method. In this paper, the calculational method of control rod worths in the design stage and the comparison with the design values and measured ones are shown. The reasons that the control rod worths change slightly in each cycle, are also investigated. (author). 13 figs, 12 tabs

  17. Seismic analysis of hydraulic control rod driving system

    International Nuclear Information System (INIS)

    Zheng, Yanhua; Bo, Hanliang; Dong, Duo

    2002-01-01

    A simplified mathematical model was developed for the Hydraulic Control Rod Driving System (HCRDS) of a 200 MW nuclear heating reactor, which incorporated the design of its chamfer-hole step cylinder, to analyze its seismic response characteristics. The control rod motion was analyzed for different sine-wave vibration loadings on platform vibrator. The vibration frequency domain and the minimum acceleration amplitude of the control rod needed to cause the control rod to step to its next setting were compared with the design acceleration amplitude spectrum. The system design was found to be safety within the calculated limits. The safety margin increased with increasing frequency. (author)

  18. 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)

  19. 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

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

    International Nuclear Information System (INIS)

    Minarcin, Miroslav; Elko, Marek

    1998-01-01

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

  2. 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.)

  3. 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

  4. Hydraulic system for driving control rods

    International Nuclear Information System (INIS)

    Okuzumi, Naoaki.

    1982-01-01

    Purpose: To enable safety reactor shut down upon occurrence of an abnormal excess pressure in a hydraulic control unit. Constitution: The actuation pressure for a pressure switch that generates a scram signal is set lower than the release pressure set to a pressure release valve. Thus, if the pressure of nitrogen gas in a nitrogen container increases such as upon exposure of the hydraulic control unit to a high temperature, the pressure switch is actuated at first to generate the scram signal and a scram valve is opened to supply water at high pressure to control rod drives under the driving force of the nitrogen gas at high pressure to rapidly insert the control element into the reactor and shut down it. If the pressure of the nitrogen gas still increases after the scram, the pressure release valve is opened to release the nitrogen gas at high temperature to the atmosphere. Since the scram is attained before the actuation of the pressure release valve, safety reactor shut down can be attained and the hydraulic control unit can be protected. (Sekiya, K.)

  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. 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.

  7. 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)

  8. Material operating behaviour of ABB BWR control rods

    International Nuclear Information System (INIS)

    Rebensdorff, B.; Bart, G.

    2000-01-01

    The BWR control rods made by ABB use boron carbide (B 4 C and hafnium as absorber material within a cladding of stainless steel. The general behaviour under operation has proven to be very good. ABB and many of their control rod customers have performed extensive inspection programs of control rod behaviour. However, due to changes in the material properties under fast and thermal neutron irradiation defects may occur in the control rods at high neutron fluences. Examinations of irradiated control rod materials have been performed in hot cell laboratories. The examinations have revealed the defect mechanism Irradiation Assisted Stress Corrosion Cracking (IASCC) to appear in the stainless steel cladding. For IASCC to occur three factors have to act simultaneously. Stress, material sensitization and an oxidising environment. Stress may be obtained from boron carbide swelling due to irradiation. Stainless steel may be sensitized to intergranular stress corrosion cracking under irradiation. Normally the reactor environment in a BWR is oxidising. The presentation focuses on findings from hot cell laboratory work on irradiated ABB BWR control rods and studies of irradiated control rod materials in the hot cells at PSI. Apart from physical, mechanical and microstructural examinations, isotope analyses were performed to describe the local isotopic burnup of boron. Consequences (such as possible B 4 C washout) of a under operation in a ABB BWR, after the occurrence of a crack is discussed based on neutron radiographic examinations of control rods operated with cracks. (author)

  9. 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

  10. 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

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

    International Nuclear Information System (INIS)

    Zhou Xiaojia; Mao Fei; Min Peng; Lin Shaoxuan

    2013-01-01

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

  12. Expert system for control rod programming of boiling water reactors

    International Nuclear Information System (INIS)

    Fukuzaki, T.; Yoshida, K.; Kobayashi, Y.; Matsuura, H.; Hoshi, K.

    1986-01-01

    Control rod programming, one of the main tasks in reactor core management of boiling water reactors (BWRs), can be successfully accomplished by well-experienced engineers. By use of core performance evaluation codes, their knowledge plays the main role in searching through optimal control rod patterns and exposure points for adjusting notch positions and exchanging rod patterns. An expert system has been developed, based on a method of knowledge engineering, to lighten the engineer's load in control rod programming. This system utilizes an inference engine suited for planning/designing problems, and stores the knowledge of well-experienced engineers in its knowledge base. In this report, the inference engine, developed considering the characteristics of the control rod programming, is introduced. Then the constitution and function of the expert system are discussed

  13. 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

  14. 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.)

  15. 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

  16. 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

  17. 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.)

  18. Nuclear reactor internals and control rod handling device

    International Nuclear Information System (INIS)

    Betancourt, G.N.; Etzel, W.W.

    1981-01-01

    A method and apparatus for removing, in an essentially continuous operation, the control rods and the upper guide structure from a nuclear reactor vessel during refueling. The apparatus includes a rigid frame which is secured to the upper guide structure after the vessel head is removed. A platform is vertically reciprocable within the frame and is adapted to engage and lift simultaneously all control rod drive shafts to a maximum elevation within the frame. A mechanical interface between the platform and the frame is provided so that continuation of the lifting force on the platform transfers the lift force to the frame whereby the upper guide structure is lifted out of the vessel. Automatically operated stop means are provided to lock the platform and rods in the maximum elevation within the frame in order to prevent accidental dropping of the rods during transfer of the upper guide structure and control rods to a temporary storage area

  19. Large test rigs verify Clinch River control rod reliability

    International Nuclear Information System (INIS)

    Michael, H.D.; Smith, G.G.

    1983-01-01

    The purpose of the Clinch River control test programme was to use multiple full-scale prototypic control rod systems for verifying the system's ability to perform reliably during simulated reactor power control and emergency shutdown operations. Two major facilities, the Shutdown Control Rod and Maintenance (Scram) facility and the Dynamic and Seismic Test (Dast) facility, were constructed. The test programme of each facility is described. (UK)

  20. Review of control rod calibration methods for irradiated AGRs

    Energy Technology Data Exchange (ETDEWEB)

    Telford, A. R.R.

    1975-10-15

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

  1. 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

  2. Cleaning and excavating tool for control rod canopy seals

    International Nuclear Information System (INIS)

    Kucera, R.A.

    1991-01-01

    This patent describes a device for servicing a weld site about the periphery of a control rod drive mechanism. It comprises a housing adapted to be rotated about the periphery of the control rod drive mechanism; a carriage reciprocably received within the housing; first movement means for reciprocating the carriage in a first direction; a tool attachment reciprocably received within the carriage; and second movement means for reciprocating the tool in a second direction; wherein the tool attachment is positioned relative to the control rod drive mechanism by the first and second movement means

  3. Hydraulic system for the drive of control rod

    International Nuclear Information System (INIS)

    Niwano, Masao.

    1978-01-01

    Purpose: To remove thermal stress and improve safety by utilizing water discharged a driving device as a part of cooling water for the device upon driving of control rods. Constitution: A water drain valve is wholly closed and a flow stabilization valve is supplied with an amount of water necessary for driving control rods. Upon driving one control rod, an amount of water required for the driving is caused to flow to the relivant hydraulic control unit and the flow rate in the stabilization valve is reduced by an amount required for the driving to keep the flow rate constant in the flow control valve. Since Excess water conventionally returned to the pressure vessel is utilized as cooling water for the driving device of control rods, the pressure vessel nozzle can be saved. Accordingly, the thermal stress in the nozzle portion can be removed to significantly improve the safety. (Seki, T.)

  4. Water pressure control device for control rod drive

    International Nuclear Information System (INIS)

    Sato, Hideyuki.

    1981-01-01

    Purpose: To minimize the fluctuations in the reactor water level upon occurrence of abnormality by inputting the level signal of the reactor to an arithmetic unit for controlling the pressure of control rod drive water to thereby enable effective reactor level control. Constitution: Signal from a flow rate transmitter is inputted into an arithmetic unit to perform constant flow rate control upon normal operation. While on the other hand, if abnormality occurs such as feedwater pump trips, the arithmetic unit is switched from the constant flow rate control to the reactor water level control. Reactor water level signal is inputted into the arithmetic unit and the control valve is most suitably controlled, whereby water is fed from CST to the reactor by way of control rod drive water system to secure the reactor water level if feedwater to the reactor is interrupted by loss of coolants on the feedwater system. Since this enables to minimize the fluctuations in the reactor water level upon abnormality, the reactor water level can be controlled most suitably by the reactor water level signal. (Moriyama, K.)

  5. 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)

  6. 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.)

  7. 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

  8. 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

  9. 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.

  10. 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

  11. 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)

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

    International Nuclear Information System (INIS)

    Nakata, H.

    1978-01-01

    A study about the method of calculation for the reactivity of control rods in graphite-moderated critical assemblies, is presented. The result of theoretical calculation, developed by super celles and Nordheim-Scalettar methods are compared with experimental results for the critical Assembly of General Atomic. The two methods are then applicable to reactivity calculation of the control rods of graphite moderated critical assemblies [pt

  13. 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

  14. 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.)

  15. 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.)

  16. Method of operating control rods for BWR type reactors

    International Nuclear Information System (INIS)

    Shirakawa, Toshihisa.

    1979-01-01

    Purpose: To eliminate the danger such as fuel element failures due to rapid power increase and form a control rod pattern for obtaining a required power level in a relatively short time. Method: Control rods are disposed so as to vertically enter into and retract from the central region of each four fuel assemblies adjacent to each other respectively. Upon operation of the control rods, every other control rods in the lateral and longitudinal directions among the entire control rods that are inserted completely are extracted completely at the lower flow limit of coolants. Then, the control rods completely inserted are divided into groups inserted deeply and groups inserted less deeply. The less deeply inserted groups are extracted just before the excess of thermal limit value successively in the lower flow limit of the coolants and then the deeply inserted groups are extracted successively till a predetermined power level in the same manner. Therefore, the coolant flow to the reactor core is increased and the power level is raised. (Furukawa, Y.)

  17. Inlet for fuel assembly having finger control rods

    International Nuclear Information System (INIS)

    Berglund, A.; Suvanto, A.; Tornblom, L.

    1975-01-01

    A nuclear reactor with vertically arranged fuel assemblies positioned on supporting members and with control rods displaceably arranged in guide tubes between the fuel rods inside the fuel assemblies is described. The supporting plate is provided with a transverse end piece with throttling means for the liquid flow which passes from below up through the supporting member and past the fuel rods in the fuel assembly. The inlets for the guide tubes for the control rods are located below the end piece and the throttling means. In this way a higher pressure prevails at the inlet to the guide tubes than above the end piece, so that a stronger flow of coolant is produced through guide tubes than through the fuel assembly. (U.S.)

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

    International Nuclear Information System (INIS)

    Zbinden, M.

    1997-01-01

    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)

  19. 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.)

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

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  1. Automatic positioning control device for automatic control rod exchanger

    International Nuclear Information System (INIS)

    Nasu, Seiji; Sasaki, Masayoshi.

    1982-01-01

    Purpose: To attain accurate positioning for a control rod exchanger. Constitution: The present position for an automatic control rod exchanger is detected by a synchro generator. An aimed stopping position for the exchanger, a stop instruction range depending on the distantial operation delay in the control system and the inertia-running distance of the mechanical system, and a coincidence confirmation range depending on the required positioning accuracy are previously set. If there is a difference between the present position and the aimed stopping position, the automatic exchanger is caused to run toward the aimed stopping position. A stop instruction is generated upon arrival at the position within said stop instruction range, and a coincidence confirmation signal is generated upon arrival at the position within the coincidence confirmation range. Since uncertain factors such as operation delay in the control system and the inertia-running distance of the mechanical system that influence the positioning accuracy are made definite by the method of actual measurement or the like and the stop instruction range and the coincidence confirmation range are set based on the measured data, the accuracy for the positioning can be improved. (Ikeda, J.)

  2. 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

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

    International Nuclear Information System (INIS)

    Itagaki, Masafumi; Miyoshi, Yoshinori; Gakuhari, Kazuhiko; Okada, Noboru; Sakai, Tomohiro

    1993-01-01

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

  4. 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

  5. Knowledge based system for control rod programming of BWRs

    International Nuclear Information System (INIS)

    Fukuzaki, Takaharu; Yoshida, Ken-ichi; Kobayashi, Yasuhiro

    1988-01-01

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

  6. 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)

  7. Method for installing a control rod driving device in a reactor

    International Nuclear Information System (INIS)

    Sato, Haruo; Watanabe, Masatoshi.

    1975-01-01

    Object: To install a device using a wire rope, including individually moving up and down a control rod and a control rod driving device thereby enabling to install them within a low house and to reduce time required for installing operation. Structure: The control rod is temporarily attached to a support structure for the control rod driving device, the control rod driving device is suspended on a crane positioned upwardly of the support structure, a rope connected to the control rod driving device is connected to the control rod, a sagged portion of the rope is then wound about a rotary cylinder, the control rod is disconnected from its temporary attachment, and the wound rope is wound back while the rotary cylinder is rotated to move down the control rod. After the rope has been released from the rotary cylinder, the control rod driving device is moved down by the crane. (Kamimura, M.)

  8. Method of controlling a control rod drive exchange apparatus

    International Nuclear Information System (INIS)

    Kase, Keiichi; Yamazaki, Kanji; Hirano, Shigeo; Takeda, Hiroyuki; Oowada, Masataka.

    1981-01-01

    Purpose: To move the mountings means for control rod drives to an aimed position easily in a short time by alternately rotating a rotational moving means and radially moving a lateral transfer means. Constitution: Positions for a rotational moving vehicle and a lateral moving vehicle are inspected respectively by synchro generators A and B. The positional signals detected by the synchro generator A is transformed into an angle by a transducer C and the positional signals detected by the synchro generator B is transformed into a radial distance by a transducer D, whereafter each of the data is transmitted to a computer. The computer controls motors to operate the rotational moving means and the lateral moving means alternately. (Seki, T.)

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

    International Nuclear Information System (INIS)

    Imasaki, Yoshio.

    1997-01-01

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

  10. Seismic analysis of control and safety rod drive mechanism

    International Nuclear Information System (INIS)

    Meher Prasad, A.; Jaya, K.P.; Chellapandi, P.; Rajan Babu, V.; Selvaraj, T.

    2003-01-01

    Control rod and its driving mechanism for a Fast Breeder Reactor is to facilitate safe shutdown of the reactor in case of emergency. A theoretical study on the seismic qualification of control and safety rod driving mechanism is carried out. Earthquake excitations under Operational Basis (ORE) and Safe Shutdown condition (SSE) are considered. The time required for the control rod to reach the bottom position in order to shut down the reaction under excited condition is traced out. The maximum displaced positions and extreme stresses in various parts of the system under excitations are evaluated. The system modeled using beam elements. The connections between different parts are modeled through rigid elements. The interaction between various parts are modeled using GAP elements. (author)

  11. PWR control rods wear by vibrations induced by coolant fluid

    International Nuclear Information System (INIS)

    Reynier, R.

    1997-01-01

    Flow induced vibrations in pressurised water reactors generate the wear of control rods against their guidance systems. Alternate sliding (at 320 deg. C in water) and impact-sliding tests (at room temperature in air) were carried out on 304 L austenitic stainless steel control rods' claddings. Microstructural analysis were made on the wear scars of the tube specimen using Scanning ELectron Microscopy, microhardness measurements and X-ray diffractometry. The alternate sliding leads to an important mass loss, a strong plastic deformation due to the strain hardening of the surface layers and generates strong compressive residual stresses. These results are specific to a severe wear case. Therefore, the impact-sliding mode induces martensitic phase, a cracked oxide layer and a compressive residual stresses weaker than those created in the alternate sliding case. This type of motion leads to a milder wear of the control rods

  12. Analysis of Rod Withdrawal at Power (RWAP) Accident using ATHLET Mod 2.2 Cycle A and RELAP5/mod 3.3 Codes

    International Nuclear Information System (INIS)

    Bencik, V.; Cavlina, N.; Grgic, D.

    2012-01-01

    The system code ATHLET is being developed at Gesellschaft fuer Anlagen-und Reaktorsicherheit (GRS) in Germany. In 1996, the NPP Krsko (NEK) input deck for ATHLET Mod 1.1 Cycle C has been developed at Faculty of Electrical Engineering (FER), University of Zagreb. The input deck was tested by analyzing the realistic plant event 'Main Steam Isolation Valve Closure' and the results were assessed against the measured data. The input deck was established before plant modernization that took place in 2000 and included the power uprate and SG replacement. The released ATHLET version (Mod 2.2 Cycle A) is now being available at FER Zagreb. Accordingly, the NEK input deck for ATHLET Mod 2.2 Cycle A has been developed. A completely new input deck has been created taking into account the large number of changes due to power uprate and SG replacement as well as taking advantage of developmental work on NEK data base performed at FER. The new NEK input deck for ATHLET code has been tested by analyzing the Rod Withdrawal Power (RWAP) accident and the results were assessed against the analysis performed by RELAP5/mod 3.3 code. The RWAP accident can be either Departure from Nucleate Boiling (DNB) ratio or overpower limiting accident depending on initial power and reactivity insertion rate. Since the automatic rod control system is assumed unavailable, the only negative reactivity is due to Doppler and moderator feedback. Consequently, the nuclear power and the transferred heat in the steam generators (SGs) increase. Since the steam flow to the turbine and the extracted power from the SGs remain constant, the SG secondary pressure and the temperatures on the primary side increase. Unless terminated by manual or automatic action, the power mismatch between primary and secondary side and the resultant coolant temperature rise could eventually result in DNB ratio and/or fuel centreline melt. In order to avoid core damage, the reactor protection system is designed to automatically

  13. 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.

  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. Evaluation of control rod motion simulator research reactors

    International Nuclear Information System (INIS)

    Sanda

    2010-01-01

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

  16. BWR ATWS mitigation by Fine Motion Control Rod

    International Nuclear Information System (INIS)

    Rohatgi, U.S.; Cheng, H.S.; Khan, H.; Mallen, A.; Diamond, D.

    1994-01-01

    Two main methods of ATWS mitigation in a SBWR are: fine Motion control Rods (FMCRD) and Boron injection via the Standby Liquid control System (SLCS). This study has demonstrated that the use of FMCRD along with feedwater runback mitigated the conditions due to reactivity insertion and possible ATWS in a BWR which is similar to SBWR

  17. Estimation and control in HTGR fuel rod fabrication

    International Nuclear Information System (INIS)

    Downing, D.J.; Bailey, J.M.

    1980-01-01

    A control algorithm has been derived for an HTGR Fuel Rod Fabrication Process utilizing the method of G.E.P. Box and G.M. Jenkins. The estimator is a Kalman filter and is compared with a Least Square estimator and a standard control chart. The effects of system delays are presented. 1 ref

  18. Characterisation of reactor control rod drives. Specification 1-6

    International Nuclear Information System (INIS)

    1975-03-01

    The committee 'Kernreaktorregelung' of VDI/VDE-Gesellschaft Mess- und Regelungstechnik has developed 6 specifications (Typenblaetter) of reactor control rod drives. The specifications are aimed at giving engineers in reactor control systems an outline concerning the function as well as some construction characteristics. (orig./LN) [de

  19. Estimation and control in HTGR fuel rod fabrication

    International Nuclear Information System (INIS)

    Downing, D.J.; Bailey, M.J.

    1980-01-01

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

  20. A digital position-indication system for control rods

    International Nuclear Information System (INIS)

    Nishizawa, Yukio; Hayakawa, Toshifumi

    1979-01-01

    Systems that detect and indicate the position of the control rods that regulate the thermal output of a nuclear reactor play a particularly important role in monitoring its operational status. Conventionally, control rod position indication in pressurized water reactors has been of the analog type, utilizing the principle of the differential transformer. The present digital system was developed with the objective of achieving greater stability, greater accuracy, and higher reliability. The article gives a general description of the system and describes its advantages. (author)

  1. A mathematical method for boiling water reactor control rod programming

    International Nuclear Information System (INIS)

    Tokumasu, S.; Hiranuma, H.; Ozawa, M.; Yokomi, M.

    1985-01-01

    A new mathematical programming method has been developed and utilized in OPROD, an existing computer code for automatic generation of control rod programs as an alternative inner-loop routine for the method of approximate programming. The new routine is constructed of a dual feasible direction algorithm, and consists essentially of two stages of iterative optimization procedures Optimization Procedures I and II. Both follow almost the same algorithm; Optimization Procedure I searches for feasible solutions and Optimization Procedure II optimizes the objective function. Optimization theory and computer simulations have demonstrated that the new routine could find optimum solutions, even if deteriorated initial control rod patterns were given

  2. Energy analysis of control rod drive mechanism in HTR-10

    International Nuclear Information System (INIS)

    Bo Hanliang; Wu Yuanqiang

    2000-01-01

    This paper presents a theoretical model for the control rod drive mechanism for the 10 MW High Temperature Gas Cooled Reactor (HTR-10) and analyzes accidents which may occur in the drive mechanism, for example, chain break, coupling damage and other damage scenarios. The results show that the matching problem between buffer capability and coupling strength is the main reason for coupling damage; increased temperatures would reduce eddy damping and cause a mismatch between buffer capability and coupling strength; and the displacement of the buffer spring will affect the coupling force. The results provide a theoretical basis for the design of the control rod drive mechanism for HTR-10

  3. 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

  4. 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.

  5. Calculation of control rods in rectangular reactor, and applications (1960)

    International Nuclear Information System (INIS)

    Goshen, S.; Pazy, A.

    1960-01-01

    The aim of this report is to find a method for estimating the anti-reactivity of control rods perpendicular to the axis in a cylindrical pile. The paper is divided into two parts. In the first is given a method of calculating control rods in a rectangular pile, similar to the Nordheim-Scalettar method for cylindrical piles. As an example the formulas are given for the theories of one and two neutron groups, the generalisation for several groups being evident. In the second part we find by a variation method a formula for estimating the Laplacian of a pile, which may be divided into parallelepipeds for which the Laplacian are given. Finally, this formula is used to calculate the anti-reactivity of rods perpendicular to the axis in a cylindrical pile. (author) [fr

  6. 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.)

  7. 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

  8. Control rod for the operation of nuclear reactor

    International Nuclear Information System (INIS)

    Ishida, Hiromi

    1987-01-01

    Purpose: To conduct spectrum shift operation without complicating the reactor core structures, reducing the probability of failures. Constitution: An operation control rod which is driven while passed vertically in the reactor core comprises a strong absorption portion, moderation portion and weak moderation portion defined orderly from above to below and the length for each of the portions is greater than the effective reactor core height. If the operation control rod is lifted to the maximum limit in the upward direction of the reactor core, the weak moderation portion is corresponded over the effective length of the reactor core. Since the weak moderation portion is filled with zirconium and moderators are not present in the operation control rod, water draining gap is formed, neutron spectral shift is formed, excess reactivity is suppressed, absorption of neutrons to fuel fertile material is increased and the formation of nuclear fission material is increased. From the middle to the final stage of the cycle, the control rod is lowered, by which the moderator/fuel effective volume ratio is increased to increase the reactivity. (Kamimura, M.)

  9. 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

  10. Fuel and control rod failure behavior during degraded core accidents

    International Nuclear Information System (INIS)

    Chung, K.S.

    1984-01-01

    As a part of the pretest and posttest analyses of Light Water Reactor Source Term Experiments (STEP) which are conducted in the Transient Reactor Test (TREAT) facility, this paper investigates the thermodynamic and material behaviors of nuclear fuel pins and control rods during severe core degradation accidents. A series of four STEP tests are being performed to simulate the characteristics of the power reactor accidents and investigate the behavior of fission product release during these accidents. To determine the release rate of the fission products from the fuel pins and the control rod materials, information concerning the timing of the clad failure and the thermodynamic conditions of the fuel pins and control rods are needed to be evaluated. Because the phase change involves a large latent heat and volume expansion, and the phase change is a direct cause of the clad failure, the understanding of the phase change phenomena, particularly information regarding how much of the fuel pin and control rod materials are melted are very important. A simple energy balance model is developed to calculate the temperature profile and melt front in various heat transfer media considering the effects of natural convection phenomena on the melting and freezing front behavior

  11. Self-actuation type electromagnet for control rod retaining mechanism

    International Nuclear Information System (INIS)

    Saito, Makoto; Gunji, Minoru.

    1989-01-01

    The present invention concerns a self-actuation type electromagnet for automatically inserting a control rod into a reactor core for the reactor scram upon occurrence of abnormality in FBR type reactors, etc. That is, a mechanism for preventing scorching is disposed to an attracting portion of a split type core thereby enabling reliable detachment of a control rod. For this purpose, less scorching material is embedded to the attracting portion between each of the core portions, with the surface being slightly protruded. In such an attracting portion, a fine gap is formed between each of the core portions by the contact of less scorching materials. Further, the scorching material is embedded into a metal ring, which is screw-coupled to one of the core portions such that the position is adjustable in the direction of the control rod. Further, the less scorching material is made of alumina. As a result, the attracting portion is neither scorched or fused even when it is used for a long period of time in liquid sodium at high temperature. Therefore, when the electromagnet loses the attracting force, the control rod drops surely. (I.S.)

  12. 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

  13. Mechanical components design for PWR - control rod drive mechanism

    International Nuclear Information System (INIS)

    Leme, Francisco Louzano; Mattar Neto, Miguel

    2002-01-01

    The Control Rod Drive Mechanism (CRDM) is usually - a high precision - equipment incorporating mechanical and electrical components designed to move the control rods. The 'control rods' refer to all rods or assemblies that are moved to assess the performance of the reactor. The CRDM here presented is the Nut and Lead Screw type. This type is basically a power screw type magnetically coupled to a slow speed reluctance electric motor that provides a means of axially positioning the movable fuel assemblies in the reactor core for purpose of controlling core reactivity. A helically threaded lead screw assembly, comprising one element of power screw, is attached to a movable fuel assemblies. The CRDM usually has closer and more consistent contact with environment peculiar to the reactor than has only other machinery component. This environment includes not only the radiation field of the reactor, but also the temperature, pressure and chemical properties associated with the material used as the coolant for reactor fuel. Specific and special materials are needed because of the above mentioned application. Due to the importance of the above described CRDM functions, this paper will also consider the nuclear functions and their safety classes as well as the CRDM nuclear design criteria. (author)

  14. 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

  15. Device for rearranging control rods of experimental reactors

    International Nuclear Information System (INIS)

    Louda, J.

    1975-01-01

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

  16. 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

  17. 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.

  18. 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.)

  19. Urea controlled hydrothermal synthesis of ammonium aluminum carbonate hydroxide rods

    Science.gov (United States)

    Wang, Fang; Zhu, Jianfeng; Liu, Hui

    2018-03-01

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

  20. 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

  1. 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

  2. Diagnostic device for failures in control rod drives

    International Nuclear Information System (INIS)

    Okutani, Tetsuro.

    1982-01-01

    Purpose: To enable to concretely point out a failure position when a failure might occur by diagnosing the failure without affecting the variation to the state of a reactor core. Constitution: A frequency switching circuit is provided in an inverter for controlling the rotating speed of a motor for discharging and charging a control rod. Then, a voltage detector is provided at asemiconductor switch provided between the inverter and the motor. When a high frequency control signal is input to the inverter in diagnosing a failure, the switching speed of the switch is accelerated, a current hardly flows through the motor, and even if the inverter is operated, the motor will not rotate. Thus, the failure of a control rod drive can be diagnosed without affecting any influence to the state of a reactor core. (Kamimura, M.)

  3. 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

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

    International Nuclear Information System (INIS)

    Hursin, M.; Kochunas, B.; Downar, T. J.

    2008-01-01

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

  5. 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

  6. 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)

  7. 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

  8. Releasing method of connection of control rod and its drive mechanism in a reactor

    International Nuclear Information System (INIS)

    Ishida, Kazuo; Futatsugi, Masao.

    1976-01-01

    Object: To disengage a control rod from a control rod drive device in a boiling water reactor with a minimal failure of the device, when connection there between cannot be released in a normal manner. Structure: First, a part of a piston tube in the control rod drive device is withdrawn externally of a control rod housing and cut. Next, a discharge tool, which is designed to be connected with the cut piston tube, is connected to the remainder of the piston tube within the housing and the aforesaid piston tube is pushed into the index tube. The index tube is then cut by the discharge tool. Thus, the control rod drive device and the control rod may be separated. Thereafter, the control rod may be removed from the top of the reactor container whereas the control rod drive device removed from the bottom thereof. (Ikeda, J.)

  9. 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

  10. Device for discharging drain in a control rod driving apparatus

    International Nuclear Information System (INIS)

    Ikeda, Tadasu; Ikuta, Takuzo; Yoshida, Tomiji; Tsukahara, Katsumi.

    1975-01-01

    Object: To efficiently and safely collect and discharge drain by a simple construction in which a drain cover and a drain tank in a control rod driving apparatus are integrally formed, and an overhauling wrench of said apparatus and a drain hose are mounted on the drain tank. Structure: When a mounting bolt is untightened by a torque wrench so as to be removed from a flange surface of the control rod driving apparatus in a nuclear reactor, axial movement of said apparatus is absorbed by a spring so that drain containing a radioactive material is discharged into a drain tank through the flange surface of said apparatus and is then guided into a collecting tank through a drain hose. (Kamimura, M.)

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

    International Nuclear Information System (INIS)

    Blazquez, J.; Vallejo, I.

    1998-01-01

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

  12. Hot rolled composite billet for nuclear control rods

    International Nuclear Information System (INIS)

    Miller, G.E.

    1976-01-01

    This invention relates to a composite plate shaped billet, useful in the fabrication of nuclear control rods, which comprises a core of stainless steel containing about 2 percent boron 10, a thin coating of zirconia on the surfaces of said core, and said zirconia coating being completely encased in a jacket of mild steel, said composite having been hot rolled between about 1075 0 and about 1165 0 C. 1 claim, 8 figures

  13. Conceptual design of control rod regulating system for plate type fuels of Triga-2000 reactor

    International Nuclear Information System (INIS)

    Eko Priyono; Saminto

    2016-01-01

    Conceptual design of the control rod regulating system for plate type fuel of TRIGA-2000 reactor has been made. Conceptual design of the control rod regulating system for plate type fuel of TRIGA-2000 reactor was made with refer to study result of instrument and control system which is used in BATAN'S reactor. Conceptual design of the control rod regulating system for plate type fuel of TRIGA-2000 reactor consist of 4 segments that is control panel, translator, driver and display. Control panel is used for regulating, safety and display control rod, translator is used for signal processing from control panel, driver is used for driving control rod and display is used for display control rod level position. The translator was designed in 2 modes operation i.e operation by using PLC modules and IC TTL modules. These conceptual design can be used as one of reference of control rod regulating system detail design. (author)

  14. 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.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-06-01

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

  16. Reconstitutable control assembly having removable control rods with detachable split upper end plugs

    International Nuclear Information System (INIS)

    Gjertsen, R.K.; Knott, R.P.; Sparrow, J.A.

    1989-01-01

    This patent describes, in a reconstitutable control assembly for use with a nuclear fuel assembly, the control assembly including a spider structure and at least one control rod, an attachment joint for detachable fastening the control rod to the spider structure. The attachment joint comprising: a hollow connecting finger on the spider structure; and an elongated detachable split upper end plug on the control rod having a pair of separate upper and lower plug portions, the upper plug portion having integrally-connected tandemly- arranged upper, middle and lower sections. The lower plug portion having integrally-connected tandemly-arranged upper, middle and lower segments

  17. 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.

  18. 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)

  19. 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)

  20. 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)

  1. 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)

  2. 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

  3. Corrugated thimble tube for controlling control rod descent in nuclear reactor

    International Nuclear Information System (INIS)

    Luetzow, H.J.

    1981-01-01

    A thimble tube construction is described which will provide a controlled descent for a control rod while minimizing the reaction forces which must be absorbed by the thimble tube and reducing the possibility that a foreign particle could interfere with the free descent of a control rod. A thimble tube is formed with helically-corrugate internal walls which cooperate with a control rod contained in the tube in an emergency situation to provide a progressively-increasing hydraulic restraining force as each adjacent corrugation is encountered

  4. Determination of equivalent cross sections for representation of control rod regions in diffusion calculations

    International Nuclear Information System (INIS)

    Scherer, W.; Neef, H.J.

    1976-07-01

    The representation of control rod regions in reactor calculations requires a combination of transport and diffusion theory calculations. A method is described which produces equivalent cross sections for a rodded region. These cross sections used in a diffusion theory calcualtion yield the same rod efficiency and reaction rate distribution as the transport theory calculation for the explicit heterogeneous control rod. The description of the method is complemented by sample problems. (orig.) [de

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

    International Nuclear Information System (INIS)

    Flessner, H.; Paeserack, U.

    1974-01-01

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

  6. Force analysis of the advanced neutron source control rod drive latch mechanism

    International Nuclear Information System (INIS)

    Damiano, B.

    1989-01-01

    The Advanced Neutron Source reactor (ANS), a proposed Department of Energy research reactor currently undergoing conceptual design at the Oak Ridge National Laboratory (ORNL), will generate a thermal neutron flux approximating 10 30 M -2 emdash S -1 . The compact core necessary to produce this flux provides little space for the shim safety control rods, which are located in the central annulus of the core. Without proper control rod drive design, the control rod drive magnets (which hold the control rod latch in a ready-to-scram position) may be unable to support the required load due to their restricted size. This paper describes the force analysis performed on the control rod latch mechanism to determine the fraction of control rod weight transferred to the drive magnet. This information will be useful during latch, control rod drive and magnet design. 5 refs., 12 figs

  7. Theoretical calibration of grey and black control rods of gas-graphite power reactors

    International Nuclear Information System (INIS)

    Joksimovic, V.

    1964-01-01

    Full text: Calculation of calibration curve for particular control rod batches is of significant importance for safety and operation reasons. The procedure presented in this paper is based on the two following criteria: Constants of the lattice region with control rods are determined by supercell method. Effective multiplication constant of the core dependent on the insertion of control rods was determined by dividing the core onto two axial and radial zones. Calculation of the black control rods takes into account epithermal absorption. Thermal extrapolated length of the control rods was calculated by using Kushneriuk-McKey relation. The extrapolated length of the grey rods and the epithermal extrapolated length of the black rods were calculated by diffusion theory. Correlation procedure was used for calculation of epithermal extrapolated length. The complete mathematical procedure was programmed for calculations on the digital ZUSE-Z-23 and ELLIOTT-803-B computers

  8. Top closure for control rod drive for nuclear reactor

    International Nuclear Information System (INIS)

    Raas, J.H.; Schwartz, J.I.

    1978-01-01

    A removable top closure and venting assembly for the tubular housing of a control rod drive includes a mounting ring threadably inserted in the upper end of the housing, a fluid-sealing closure member beneath the mounting ring and which is mounted in and coupled to the mounting ring by means of a ball and socket joint, a gas vent defined by interconnecting passages extending through the closure and through the ball and socket joint, and a vent valve accessible from the top of the closure assembly. 3 claims, 2 figures

  9. Fabrication of control rod system of RSG-GAS

    International Nuclear Information System (INIS)

    Hari-Sudirdjo; Setyono; Hendra-Prasetya

    2003-01-01

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

  10. 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

  11. 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.

  12. Method for depleting BWRs using optimal control rod patterns

    International Nuclear Information System (INIS)

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

    1991-01-01

    Control rod (CR) programming is an essential core management activity for boiling water reactors (BWRs). After establishing a core reload design for a BWR, CR programming is performed to develop a sequence of exposure-dependent CR patterns that assure the safe and effective depletion of the core through a reactor cycle. A time-variant target power distribution approach has been assumed in this study. The authors have developed OCTOPUS to implement a new two-step method for designing semioptimal CR programs for BWRs. The optimization procedure of OCTOPUS is based on the method of approximation programming and uses the SIMULATE-E code for nucleonics calculations

  13. Stepping movement analysis of control rod drive mechanism

    International Nuclear Information System (INIS)

    Xu Yantao; Zu Hongbiao

    2013-01-01

    Background: Control rod drive mechanism (CRDM) is one of the important safety-related equipment for nuclear power plants. Purpose: The operating parameters of stepping movement, including lifting loads, step distance and step velocity, are all critical design targets. Methods: FEA and numerical simulation are used to analyze stepping movement separately. Results: The motion equations of the movable magnet in stepping movement are established by load analysis. Gravitation, magnetic force, fluid resistance and spring force are all in consideration in the load analysis. The operating parameters of stepping movement are given. Conclusions: The results, including time history curves of force, speed and etc, can positively used in the design of CRDM. (authors)

  14. 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

  15. 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)

  16. 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

  17. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-05-15

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

  19. The estimation of the control rods absorber burn-up during the VVER-1000 operation

    Energy Technology Data Exchange (ETDEWEB)

    Bolshagin, Sergey N.; Gorodkov, Sergey S.; Sukhino-Khomenko, Evgeniya A. [National Research Centre ' Kurchatov Institute' , Moscow (Russian Federation)

    2013-09-15

    The isotopic composition of the control rods absorber changes under the neutron flux influence, so the control rods efficiency can decrease. In the VVER-1000 control rods boron carbide and dysprosium titanate are used as absorbing materials. In boric part the efficiency decreases due to the {sup 10}B isotope burn-up. Dysprosium isotopes turn into other absorbing isotopes, so the absorbing properties of dysprosium part decrease to a lesser degree. Also the control rod's shells may be deformed as a consequence of boron carbide radiation swelling. This fact should be considered in substantiation of control rods durability. For the estimation of the control rods absorber burn-up two models are developed: VVER-1000 3-D fuel assembly with control rods partially immersed (imitation of the control rods operation in the working group) and VVER-1000 3-D fuel assembly with control rods, located at the upper limit switch (imitation of the control rods operation in groups of the emergency shutdown system). (orig.)

  20. 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

  1. The experimental development and performance test of the pneumatic control-rod drive for the THTR

    International Nuclear Information System (INIS)

    Lange, G.; Boehlo, D.; Heim, H.; Kleine-Tebbe, A.

    1976-01-01

    Reactor control and shutdown of the THTR is accomplished by two independent systems, the first consisting of 36 absorber rods penetrating the graphite reflector region surrounding the core, the second consisting of 42 absorber rods that insert directly into the pebble bed core. This paper describes the design development and testing of the pneumatic rod drives used for movement of the 42 core control rods. The core control rods have two functions: the first, for reactor safety purposes, provides for adequate safe shutdown of the reactor under cold conditions; the second, for operational purposes, provides for compensation of slow changes in reactivity. The safety and operational functions for each absorber rod are respectively carried out by a long-stroke-piston pneumatic drive and by a stepping-piston pneumatic drive, both of these independent, helium-driven drives being incorporated in the rod drive unit for each control rod. To study the performance of the rod drive, a complete prototype control rod and rod drive unit was built and tested under simulated reactor operational conditions. Operational experience under helium temperatures and pressures was gained and the drives were tested under stress and simulated accident conditions. The reliability of this system has been demonstrated to licensing authorities and to the customer. The programme will be completed with the commissioning tests of drives for the THTR-300 reactor. (author)

  2. B4C control rod behavior during severe accident sequences

    International Nuclear Information System (INIS)

    Steinbrueck, M.

    2003-01-01

    The oxidation kinetics of various types of boron carbides (pellets, powder) as well as the degradation of B 4 C control rod segments were investigated in the temperature range between 800 and 1600 deg C. Mass spectrometric gas analysis was used to determine oxidation rates in transient and isothermal tests. The oxidation kinetics of boron carbide are determined by the formation of a liquid boron oxide layer and its loss due to the reaction with surplus steam to form volatile boric acids and at temperatures above 1500 deg C by direct evaporation. Under these test conditions linear oxidation kinetics are established soon after oxidation has initiated. The oxidation kinetics are strongly influenced by the thermal-hydraulic boundary conditions, in particular by the steam flow rate. Only very low amounts of methane were ever produced in these tests. Enhanced degradation of B 4 C control rods starts with the rapid formation of eutectic melts in the systems B 4 C-stainless steel (SS) and SS-Zircaloy at temperatures above 1250 deg C. Initially, this melt is kept within a ZrO 2 scale externally formed at the Zircaloy guide tube. The absorber melt is rapidly oxidized after failure of the oxide shell and aggressively attacks adjacent fuel claddings. (author)

  3. 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.

  4. 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.)

  5. Adequacy of the analysis of mock-up control rod experiment with FCA

    International Nuclear Information System (INIS)

    Mizoo, Nobutatsu; Nakano, Masafumi

    1977-07-01

    A method of numerical analysis has been investigated for the mock-up control rod experiment of FCA VII-1 assembly constructed as the engineering mock-up of prototype fast breeder reactor MONJU. The results of criticality and B 4 C mock-up control rod worths analysis for the assembly are described in comparison with the experimental ones. The tendency of the C/E value with 10 B enrichment and the interaction effect of the multiple rods array was also examined. Reactivities and the mock-up rods worths were obtained with the X-Y geometry six groups diffusion theory. Twelve kinds of the mock-up rods with different 10 B contents and/or enrichments were used in the experiment; effective cross-sections are provided for each rod by calculation using the collision probability method. Criticality of VII-1 90Z assembly is underestimated for 3 reference critical configurations, ranging from -0.65%Δk/k to -0.77%Δk/k. The C/E values at core center for 12 kinds of B 4 C mock-up rods range from 1.03 to 1.09. The overestimate of the rod worth increases with macroscopic absorption cross-section of the rod region. The C/E values for 24 different arrays of the mock-up rods ranging from single rod to five rods lie between 1.04 and 1.08. The C/E value tends to decrease with increase in the number of rods inserted, the values for five rods arrays being about 4% lower than those for single rod arrays. The calculated interaction effects of the multiple rods arrays are slightly more negative than the experimental ones. (auth.)

  6. Automatic exchange unit for control rod drive device

    International Nuclear Information System (INIS)

    Nasu, Seiji; Sasaki, Masayoshi.

    1982-01-01

    Purpose: To enable automatic reoperation and continuation without external power interruption remedy device at the time of recovering the interrupted power soruce during automatic positioning operation. Constitution: In case of an automatic exchange unit for a control rod drive device of the control type for setting the deviation between the positioning target position and the present position of the device to zero, the position data of the drive device of the positioning target value of the device is automatically read, and an interlock of operation inhibit is applied to a control system until the data reading is completed and automatic operation start or restart conditions are sequentially confirmed. After the confirmation, the interlock is released to start the automatic operation or reoperation. Accordingly, the automatic operation can be safely restarted and continued. (Yoshihara, H.)

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

    Energy Technology Data Exchange (ETDEWEB)

    Gallozzi-Ulmann, A.; Couturier, P.; Amgarou, K.; Rothan, D.; Menaa, N. [CANBERRA France,1 rue des Herons, 78182 ST Quentin Yvelines Cedex (France); Chard, P. [CANBERRA UK, Lower Dunbeath House, Forss Business Park, Thurso, Caithness KW14 7UZ (United Kingdom)

    2015-07-01

    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 {sup 60}Co, {sup 110m}Ag and {sup 108m}Ag activity profiles (along the length of the control rods) and neutron results including Total Measurement Uncertainty evaluations. Full self-consistency checks were

  8. 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)

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

    International Nuclear Information System (INIS)

    Nakano, Masaaki; Fujimoto, Nozomu; Yamashita, Kiyonobu

    1999-01-01

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

  10. Control rod interaction models for use in 2D and 3D reactor geometries

    International Nuclear Information System (INIS)

    Bannerman, R.C.

    1985-11-01

    Control rod interaction models are developed for use in two-dimensional and three-dimensional reactor geometries. These models allow the total worth of any combination of control rods inserted to be determined from the individual worths in conjunction with an algorithm representing interaction effects between them. The validity of the assumptions is demonstrated for fast and thermal systems showing modelling errors of 1#percent# or less in inserted control rod worths. The models are ideally suited for most reactor systems. (UK)

  11. HCPWR type fuel elements design with respect to its control rods

    International Nuclear Information System (INIS)

    Abbate, M.J.; Sbaffoni, M.M.; Patino, N.E.; Torasso, O.

    1992-01-01

    The high conversion reactors (HCPWR) can improve the nuclear fuel utilization. One of its present problems is the optimization of the control rods' worth and its relationship with the void coefficient. This investigation means, starting from one reference's design, to analyze several possibilities on the number and distribution of the control rods. As main result, one design including 24 control rods in an optimized distribution, is recommended. (author)

  12. Improved Monte Carlo-perturbation method for estimation of control rod worths in a research reactor

    International Nuclear Information System (INIS)

    Kalcheva, Silva; Koonen, Edgar

    2009-01-01

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

  13. 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.

  14. 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)

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

    International Nuclear Information System (INIS)

    Zhang Zhihua; Qian Dazhi; Xu Xianqi; Huang Hongwen; Zhang Zhengming; Wu Xinxin; Hu Xiao

    2012-01-01

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

  16. Dynamic insertion analysis of control rods of BWR under seismic excitation

    International Nuclear Information System (INIS)

    Nakagawa, Masaki; Koide, Yuichi; Fukushi, Naoki; Ishigaki, Hirokuni; Okumura, Kazue

    2007-01-01

    The dynamic insertion characteristics of the control rods for the boiling water reactors under the seismic excitation are investigated using non-linear analytical models. The control rod insertion capability is one of the most important items for the safety of nuclear power plants under the seismic events. Predicting the control rod insertion behavior during the earthquake is important in the course of the control rod seismic design. We developed the analytical models using the finite element method (FEM). The effect of the interaction force between the control rod and the fuel assemblies is considered in the non-linear analysis. This interaction force courses the resistance force to the control rod during its insertion behavior. The validity of analytical methods was confirmed by comparing the analytical results with the experimental ones. Using the analytical models, the effects of input seismic motion and structural parameters of the control rods and the fuel assemblies, such as the thickness of the channel box, on the insertion time are investigated. These analytical methods can predict insertion time of the control rod, and are useful for the seismic design of the control rod assemblies. (author)

  17. The safety feature of hydraulic driving system of control rod for 200 MW nuclear heating reactor

    International Nuclear Information System (INIS)

    Chi Zongbo; Wu Yuanqiang

    1997-01-01

    The hydraulic driving system of control rod is used as control rod drive mechanism in 200 MW nuclear heating reactor. Design of this system is based on passive system, integrating drive and guide of control rod. The author analyzes the inherent safety and the design safety of this system, with mechanism of control rod not ejecting when the pressure of pressure vessel is lost, and calculating result of core not exposing when the amount of coolant is drained by broken pipe. The results indicate that this system has good safety feature, and assures reactor safety under any accident conditions, providing important technology support for 200 MW nuclear heating reactor with inherent safety feature

  18. Preliminary performance test of control rod position indicator for ballscrew type CEDM

    International Nuclear Information System (INIS)

    Yoo, J. Y.; Kim, J. H.; Hu, H.; Lee, J. S.; Kim, J. I.

    2003-01-01

    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. A survey 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. The prototype of control rod position indicator having the high performance for the ballscrew type CEDM was developed on the basis of RSPT technology identified through the survey. The characteristics of control rod position indicator was defined and documented through design procedure and preliminary performance test

  19. Acupuncture for alcohol withdrawal: a randomized controlled trial.

    Science.gov (United States)

    Trümpler, François; Oez, Suzan; Stähli, Peter; Brenner, Hans Dieter; Jüni, Peter

    2003-01-01

    Previous trials on acupuncture in alcohol addiction were in outpatients and focused on relapse prevention. Rates of dropout were high and interpretation of results difficult. We compared auricular laser and needle acupuncture with sham laser stimulation in reducing the duration of alcohol withdrawal. Inpatients undergoing alcohol withdrawal were randomly allocated to laser acupuncture (n = 17), needle acupuncture (n = 15) or sham laser stimulation (n = 16). Attempts were made to blind patients, therapists and outcome assessors, but this was not feasible for needle acupuncture. The duration of withdrawal symptoms (as assessed using a nurse-rated scale) was the primary outcome; the duration of sedative prescription was the secondary outcome. Patients randomized to laser and sham laser had identical withdrawal symptom durations (median 4 days). Patients randomized to needle stimulation had a shorter duration of withdrawal symptoms (median 3 days; P = 0.019 versus sham intervention), and tended to have a shorter duration of sedative use, but these differences diminished after adjustment for baseline differences. The data from this pilot trial do not suggest a relevant benefit of auricular laser acupuncture for alcohol withdrawal. A larger trial including adequate sham interventions is needed, however, to reliably determine the effectiveness of any type of auricular acupuncture in this condition.

  20. Establishment of analysis procedure for control rod reactivity worth

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-03-01

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

  1. Establishment of analysis procedure for control rod reactivity worth

    International Nuclear Information System (INIS)

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

    2001-03-01

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

  2. The Hitrex Programme: unperturbed HTR lattice and control rod measurements

    Energy Technology Data Exchange (ETDEWEB)

    Beynon, A J; Nunn, D L

    1972-06-15

    Reactivity, power distributions, plutonium production and fast neutron graphite damage are being studied at Berkeley Nuclear Laboratories (BNL) on the HTR 'Hitrex' reactor under cold clean conditions. Rod interactions, important in assessing local criticality hazards, are receiving special attention in the measurements. The proposals for the first two series of measurements on Hitrex are discussed in this note, Hitrex 1a being the unperturbed reactor, and Hitrex 1b the same fuel array but with a number of different control absorber loadings in it. Common to both series will be cross pin, cross block and cross core measurements of power rating, thermal spectrum and damage dose distributions, so that these will be known as functions of the fuel, reflector and absorber environment.

  3. 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)

  4. 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.)

  5. Radial brake assembly for a control rod drive

    International Nuclear Information System (INIS)

    Hekmati, A.; Gibo, E.Y.

    1992-01-01

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

  6. 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)

  7. 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

  8. Control-rod, pressure and flow-induced accident and transient analysis of a direct-cycle, supercritical-pressure, light-water-cooled fast breeder reactor

    International Nuclear Information System (INIS)

    Kitoh, Kazuaki; Koshizuka, Seiichi; Oka, Yoshiaki

    1996-01-01

    The features of the direct-cycle, supercritical-pressure, light-water-cooled fast breeder reactor (SCFBR) are high thermal efficiency and simple reactor system. The safety principle is basically the same as that of an LWR since it is a water-cooled reactor. Maintaining the core flow is the basic safety requirement of the reactor, since its coolant system is the one through type. The transient behaviors at control rod, pressure and flow-induced abnormalities are analyzed and presented in this paper. The results of flow-induced transients of SCFBR were reported at ICONE-3, though pressure change was neglected. The change of fuel temperature distribution is also considered for the analysis of the rapid reactivity-induced transients such as control rod withdrawal. Total loss of flow and pump seizure are analyzed as the accidents. Loss of load, control rod withdrawal from the normal operation, loss of feedwater heating, inadvertent start of an auxiliary feedwater pump, partial loss of coolant flow and loss of external power are analyzed as the transients. The behavior of the flow-induced transients is not so much different from the analyses assuming constant pressure. Fly wheels should be equipped with the feedwater pumps to prolong the coast-down time more than 10s and to cope with the total loss of flow accident. The coolant density coefficient of the SCFBR is less than one tenth of a BWR in which the recirculation flow is used for the power control. The over pressurization transients at the loss of load is not so severe as that of a BWR. The power reaches 120%. The minimum deterioration heat flux ratio (MDHFR) and the maximum pressure are sufficiently lower than the criteria; MDHFR above 1.0 and pressure ratio below 1.10 of 27.5 MPa, maximum pressure for operation. Among the reactivity abnormalities, the control rod withdrawal transient from the normal operation is analyzed

  9. 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

  10. 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.

  11. On-line fuel and control rod integrity management in BWRs

    International Nuclear Information System (INIS)

    Larsson, Irina; Sihver, Lembit

    2011-01-01

    Surveillance of fuel and control rod integrity in a BWR core is essential to maintain a safe and reliable operation of a nuclear power plant. An accurate and prompt way to monitor fuel integrity in a reactor core during reactor operation is by using on-line measurements of the gamma emitting noble gas activities in the off-gas system. The integrity of control rods can be efficiently followed by on-line measurements of the helium (He) concentration in the off-gases. This method also gives information about fuel rod failures since He is used as a fill gas in the fuel rods. To survey fuel and control rod integrity during reactor operation, a system consisting of combined gamma and He on-line measurements in the off-gases should be used. Such a system can detect and follow the behavior of fuel and control rod failures. In addition, it can separate fuel failures from control rod failures since fuel rods contain both He and gamma emitting noble gases, while control rods only contain He. Moreover, the system is able to distinguish primary fuel failures from degradation of already existing ones. In this paper we present a combined system for on-line measurements of He and gamma emitting noble gases in the reactor off-gas system and measuring experiences from different BWRs. (author)

  12. Calculation of control rod worth with mutual interaction

    International Nuclear Information System (INIS)

    Balthar, M.C.V.; Oliveira Vellozo, S. de; Carvalho Vital, H. de

    1989-01-01

    This work presents a two-dimensional model for determining the total worth of a set of N absorbing rods. The model simplifies the evaluation of the interaction coefficient among rods by analysing them in pairs and attributes to it a purely geometrical character. Comparisons with conventional calculational methods indicate that the results are in error by less than 6%. (author) [pt

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

    International Nuclear Information System (INIS)

    Kameswari, K.; Palanisami, K.; Thirugnana Murthy, D.; Murali, N.; Satyamurty, S.A.V.

    2013-01-01

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

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

    International Nuclear Information System (INIS)

    Itagaki, Masafumi; Miyoshi, Yoshinori; Gakuhari, Kazuhiko; Okada, Noboru; Sakai, Tomohiro.

    1993-01-01

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

  16. 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

  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. A connection-disconnection device for the nuclear reactor control rods

    International Nuclear Information System (INIS)

    Drean, H.; Breant, C.

    1992-01-01

    An automatic and relatively light system is developed to connect and disconnect the control rods in a nuclear reactor; it is designed to apply, in a controlled manner, constraints on the rods and to detect a possible misalignment of the connection-disconnection means in the corresponding butt

  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. 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)

  2. Reconstitutable control assembly having removable control rods with detachable split upper end plugs

    International Nuclear Information System (INIS)

    Gjertsen, R.K.; Knott, R.P.; Sparrow, J.A.

    1991-01-01

    This patent describes, for use in facilitating replacement of a neutron absorber control rod on a control assembly spider structure, an end plug. It comprises a pair of separate upper and lower plug portions; the upper section of the upper plug portion being configured for rigid attachment; the middle section of the upper plug portion having angularly displaced flat surfaces formed on the exterior

  3. Optimization of boiling water reactor control rod patterns using linear search

    International Nuclear Information System (INIS)

    Kiguchi, T.; Doi, K.; Fikuzaki, T.; Frogner, B.; Lin, C.; Long, A.B.

    1984-01-01

    A computer program for searching the optimal control rod pattern has been developed. The program is able to find a control rod pattern where the resulting power distribution is optimal in the sense that it is the closest to the desired power distribution, and it satisfies all operational constraints. The search procedure consists of iterative uses of two steps: sensitivity analyses of local power and thermal margins using a three-dimensional reactor simulator for a simplified prediction model; linear search for the optimal control rod pattern with the simplified model. The optimal control rod pattern is found along the direction where the performance index gradient is the steepest. This program has been verified to find the optimal control rod pattern through simulations using operational data from the Oyster Creek Reactor

  4. A method of taking control rod history into account in core simulation calculations for BWR'S

    International Nuclear Information System (INIS)

    Hojerup, C.F.; Nonbol, E.

    1990-01-01

    The problem of taking control rod history into account in core simulator codes using precalculated cross sections has been examined, and two methods have been devised and tested. The very demanding first method, using the accumulated control rod in burn-up as a parameter, turned out to be even more inaccurate than the much less demanding second method, which only requires two full burn-up histories, one with the control rod in all the time, and another with the control rod out all the time. From the analysis it can be seen that the proper treatment of the control rod history is quite important, both for the cross sections, as several per cent on the reactivity are at stake, as for the pin powers, which for some pins are very much affected

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

    International Nuclear Information System (INIS)

    Garis, N.S.; Temesvari, E.; Pazsit, I.

    1996-08-01

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

  6. Experimental study of the pressure discharge process for the hydraulic control rod drive system stepped cylinder

    International Nuclear Information System (INIS)

    Wang, Jinhua; Bo, Hanliang; Zheng, Wenxiang

    2002-01-01

    The pressure discharge process from the stepped cylinder of the Hydraulic Control Rod Drive System (HCRDS) was studied experimentally in the HCRDS experimental loop for the 200 MW Nuclear Heating Reactor (NHR-200). The results showed that the differential pressure between the outside and the inside of the stepped cylinder increased rapidly to the desired value so that the force induced by the differential pressure which pushes the out tube of stepped cylinder was large enough. Therefore, if the hydraulic control rod were jammed, the pressure could push the hydraulic control rod to overcome the frictional resistance to insert the control rod into the reactor core. The experimental results verified that this design would solve the problem of hydraulic control rod jamming during an accident. (author)

  7. Prevention and preservation aid system for control rod drives

    International Nuclear Information System (INIS)

    Ishisato, Shin-ichi; Yamamoto, Yoko.

    1992-01-01

    The system of the present invention can select control rod drives (CRD) as an object of inspection, and can manage maintenance hysteresis even by unskilled persons upon maintenance operation for the CRD. That is, the system of the present invention comprises a data base concerning prevention and preservation for the CRD and hydraulic pressure control unit (HCU), a data base management device for retrieving and managing the intelligence of the data base and a maintenance data base for storing data measured based on the data base on every periodical inspections. Further, it also comprises a function for displaying, on a map, the CRD to be disassembled and inspected upon periodical inspection on every inspection recommendation priority groups, based on these data base. Further, it also comprises a function for evaluating exchange hysteresis maintenance data for incore structures which require periodical exchange. As a result, high reliability of the CRD can be maintained and reliability of a nuclear power plant can further be improved. (I.S.)

  8. Bottom-mounted control rod drive mechanism for KJRR

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jin Haeng; Kim, Sanghaun; Yoo, Yeon-Sik, E-mail: yooys@kaeri.re.kr; Cho, Yeong-Garp; Huh, Hyung; Lee, Hyokwang; Sun, Jong-Oh; Ryu, Jeong-Soo

    2016-04-15

    Highlights: • The basic design features and characteristics of the KJRR BMCRDM are described. • The similarities and differences of some research reactor CRDMs are compared. • The current status of the design and development of the CRDM is described. • The future plan of the qualification tests of the CRDM is summarized. - Abstract: The KIJANG research reactor (KJRR), which is currently being designed by Korea Atomic Energy Research Institute, is a pool type research reactor with 15 MW of thermal power. Contrary to the top-mounted control rod drive mechanism (CRDM), the main drive mechanism of the KJRR CRDM is located in a reactivity control mechanism room under the reactor pool bottom. Recently, we accomplished the design and development of a prototype CRDM. In this paper, we introduce the basic design concept of the bottom-mounted CRDM for KJRR, and compare the similarities and differences of some research reactor CRDMs. The current status of the prototype CRDM development based on a finite element analysis and experimental verification, and the future plan of the CRDM qualification tests, are both described.

  9. 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

  10. TRACY transient experiment databook. 2) ramp withdrawal experiment

    International Nuclear Information System (INIS)

    Nakajima, Ken; Yamane, Yuichi; Ogawa, Kazuhiko; Aizawa, Eiju; Yanagisawa, Hiroshi; Miyoshi, Yoshinori

    2002-03-01

    This is a databook of TRACY ''ramp withdrawal'' experiments. TRACY is a reactor to perform supercritical experiments using low-enriched uranyl nitrate aqueous solution. The excess reactivity of TRACY is 3$ at maximum, and it is inserted by feeding the solution to a core tank or by withdrawing a control rod, which is called as the transient rod, from the core. In the ramp withdrawal experiment, the supercritical experiment is initiated by withdrawing the transient rod from the core in a constant speed using a motor drive system. The data in the present databook consist of datasheets and graphs. Experimental conditions and typical values of measured parameters are tabulated in the datasheet. In the graph, power and temperature profiles are plotted. Those data are useful for the investigation of criticality accidents with fissile solutions, and for validation of criticality accident analysis codes. (author)

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

    International Nuclear Information System (INIS)

    Humbert, G.; Ruelle, B.; Daguzan, G.; Stanculescu, A.; Kappler, F.; Scholtyssek, W.; Bouscavet, D.; Martini, M.; Broccoli, U.

    1982-01-01

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

  12. 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

  13. 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)

  14. 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)

  15. 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

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

    International Nuclear Information System (INIS)

    Hernandez, J.L.; Alonso, G.; Perusquia, R.; Montes, J.L.; Hernandez, H.

    2003-01-01

    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 ∼ 1 pcm in hot condition and of ∼ 20 pcm in cold condition. When burning three cycles those rods more burnt reached the 13,900 MWD/TM, equivalent to 36% of B-10 reduction, near value to 34% proposed by aging in the one lost study of B-10. It was observed that Cm-Presto it doesn't burn the superior node of the control rods when these are completely extracted. A one big lost of B-10, of the order of 50%, it represents only a decrease of 11% of the reactivity value of the rod. One can affirm that even when it is strongly decreased the content of B-10, the rod is continue considering as a black absorber, that is to say, thermal neutron that enters in the neutron rod that is absorbed. (Author)

  17. Calculation of control rod oscillations in a hexagonal flow channel by means of the non-stationary pressure distribution around the rods

    International Nuclear Information System (INIS)

    Grunwald, G.; Mueller, E.

    1983-08-01

    For the computation of control rod oscillations in a flow channel we set up the differential equations for the non-stationary pressure distribution around the control elements which are coupled with the motion equations of the rods. The equation system is solved by means of a finite difference method. An example shows the efficiency of the numerical calculation procedure. (author)

  18. Experience feedback of operational events of the control rod assembly and its drive mechanism in nuclear power plants

    International Nuclear Information System (INIS)

    Zhou Hong; Xiao Zhi; Tao Shusheng; Zheng Lixin; Chen Zhaolin

    2013-01-01

    Seventeen operational events of the control rod assembly and its drive mechanism are collected from 1992 to 2012 important nuclear operational events and feedback in referred nuclear power plants. After investigated and classified, several important issues, such as the impact of control rod swell and fuel assembly distortion, control rod drive mechanism leakage, and the control system reliability of control rod, are emphatically analyzed. Some suggestions of experience feedback are proposed. (authors)

  19. 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

  20. Aging assessment of BWR control rod drive systems

    International Nuclear Information System (INIS)

    Greene, R.H.

    1992-01-01

    This Phase 1 Nuclear Plant Aging Research (NPAR) study examines the aging phenomena associated with boiling water reactor (BWR) control rod drive mechanisms (CRDMs) and assesses the merits of various methods of managing this aging. Information for this study was acquired from (1) the results of a special CRDM aging questionnaire distributed to each US BWR utility, (2) a first-of-its-kind workshop held to discuss CRDM aging and maintenance concerns, (3) an analysis of NPRDS failure cases attributed to the CRD system, and (4) personal information exchange. As part of this study, nearly 3,500 NPRDS failure reports have been analyzed to examine the prevailing failure trends for CRD system components. An investigation has been conducted that summarizes 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 regarding specific actions that utilities can implement to mitigate CRDM aging. An evaluation has also been made of certain practices and tooling which have enabled some utilities to reduce ALARA exposures received from routine CRDM replacement and rebuilding activities

  1. 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

  2. Aging assessment of BWR control rod drive systems

    International Nuclear Information System (INIS)

    Greene, R.H.

    1991-01-01

    This Phase 1 Nuclear Plant Aging Research (NPAR) study examines the aging phenomena associated with boiling water reactor (BWR) control rod drive mechanisms (CRDMs) and assesses the merits of various methods of managing this aging. Information for this study was acquired from (1) the results of a special CRDM aging questionnaire distributed to each US BWR utility, (2) a first-of-its-kind workshop held to discuss CRDM aging and maintenance concerns, (3) an analysis of NPRDS failure cases attributed to the CRD system, and (4) personal information exchange. As part of this study, nearly 3,500 NPRDS failure reports have been analyzed to examine the prevailing failure trends for CRD system components. An investigation has been conducted that summarizes 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 regarding specific actions that utilities can implement to mitigate CRDM aging. An evaluation has also been made of certain practices and tooling which have enabled some utilities to reduce ALARA exposures received from routine CRDM replacement and rebuilding activities

  3. 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

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

    International Nuclear Information System (INIS)

    Bahuguna, Sushil; Dhage, Sangeeta; Nawaj, S.; Salek, C.; Lahiri, S.K.; Marathe, P.P.; Mukhopadhyay, S.; Taly, Y.K.

    2015-01-01

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

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

    International Nuclear Information System (INIS)

    Souza dos Santos, Rubens; Rubens Maiorino, Jose

    1999-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Blotcky, A J; Arsenault, L J [General Medical Research, Veterans Administration Hospital, Omaha (United States)

    1974-07-01

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

  7. 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.

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

    International Nuclear Information System (INIS)

    Blotcky, A.J.; Arsenault, L.J.

    1974-01-01

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

  9. 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)

  10. 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)

  11. Polynomial curve fitting for control rod worth using least square numerical analysis

    International Nuclear Information System (INIS)

    Muhammad Husamuddin Abdul Khalil; Mark Dennis Usang; Julia Abdul Karim; Mohd Amin Sharifuldin Salleh

    2012-01-01

    RTP must have sufficient excess reactivity to compensate the negative reactivity feedback effects such as those caused by the fuel temperature and power defects of reactivity, fuel burn-up and to allow full power operation for predetermined period of time. To compensate this excess reactivity, it is necessary to introduce an amount of negative reactivity by adjusting or controlling the control rods at will. Control rod worth depends largely upon the value of the neutron flux at the location of the rod and reflected by a polynomial curve. Purpose of this paper is to rule out the polynomial curve fitting using least square numerical techniques via MATLAB compatible language. (author)

  12. 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

  13. Calculation of optimum control rod operation programme for boiling water reactor

    International Nuclear Information System (INIS)

    Fehr, L.

    1978-01-01

    Control rod operation programmes are calculated based on a three dimensional Boiling Water Reactor situation model. The position of the control rods at variosu burn-ups is chosen by an optimisation so that the sum of the square deviations of the load density distribution from an optimum distribution ('Haling' distribution) are minimised. Other conditions are remaining critical and observing the thermal limits for central fuel element melting and critical heat surface loading. As an example, an optimum control rod operation programme for the first cycle in Lengen nuclear power station is calculated and is compared with the programme actually used. (orig.) 891 HP [de

  14. Comparison of cell homogenization methods considering interaction effect between fuel cells and control rod cells

    International Nuclear Information System (INIS)

    Takeda, T.; Uto, N.

    1988-01-01

    Several methods to determine cell-averaged group cross sections and anisotropic diffusion coefficients which consider the interaction effect between core fuel cells and control rods or control rod followers have been compared to discuss the physical meaning included in cell homogenization. As the cell homogenization methods considered are the commonly used flux-weighting method, the reaction rate preservation method and the reactivity preservation method. These homogenization methods have been applied to control rod worth calculations in 1-D slab cores to investigate their applicability. (author). 6 refs, 2 figs, 9 tabs

  15. 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.

  16. 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)

  17. 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

  18. 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)

  19. 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.)

  20. Physical resuspension and revaporisation phenomena in control rod aerosols

    International Nuclear Information System (INIS)

    Benson, C.G.; Browsher, B.R.

    1988-12-01

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

  1. 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

  2. NDE Assessment of PWSCC in Control Rod Drive Mechanism Housings

    International Nuclear Information System (INIS)

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

    2006-01-01

    Studies being conducted at the Pacific Northwest National Laboratory (PNNL) in Richland, Washington are focused on assessing the effectiveness of Nondestructive examination (NDE) techniques for inspecting control rod drive mechanism (CRDM) nozzles and J-groove weldments. The primary objective of this work is to provide information to the United States Nuclear Regulatory Commission (US NRC) on the effectiveness of NDE methods as related to the in-service inspection of CRDM nozzles and J-groove weldments, and to enhance the knowledge base of primary water stress corrosion cracking (PWSCC) through destructive characterization of the CRDM assemblies. In describing two CRDM assemblies removed from service, decontaminated, and then used in a series of NDE measurements, this paper will address the following questions: (1) What did each technique detect?, (2) What did each technique miss?, (3) How accurately did each technique characterize the detected flaws? Two CRDM assemblies including the CRDM nozzle, the J-groove weld, buttering, and a portion of the ferritic head material were selected for this study. One contained suspected PWSCC, based on in-service inspection data and through-wall leakage; the other contained evidence suggesting through-wall leakage, but this was unconfirmed. The selected 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 inspecting the J-groove weld and buttering. This paper will also describe the NDE methods used and discuss the NDE results. Future work will involve using the results from these NDE studies to guide the development of a destructive characterization plan to reveal the crack morphology and a comparison of the degradation found by the destructive evaluation with the recorded NDE responses.

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

    International Nuclear Information System (INIS)

    Kalcheva, Silva; Koonen, Edgar

    2008-01-01

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

  4. Design of Simulink module for dynamic reactivity simulation of marine reactor automatic control rod

    International Nuclear Information System (INIS)

    Chen Zhiyun; Luo Lei; Chen Wenzhen; Gui Xuewen

    2010-01-01

    The power of marine reactor varies frequently and acutely, which induces the frequent and acute adjustment of the automatic control rod. According to the characteristics of marine reactor and the problem of improper control rod reactivity insertion in previous literatures, the Simulink module for dynamic reactivity simulation of automatic control rod was designed and adopted as a sub-module of Simulink program for the fast calculation of the physical and thermal parameters of marine reactor. A typical dynamic process of the marine reactor was used as the benchmark, which indicates that the designed Simulink module is capable of the dynamic simulation of automatic control rod position and reactivity, and is adequate to the fast calculation of physic and thermal parameters. The Simulink module is of significant meaning to the simulation of the dynamic process of marine reactor and the fast calculation of the operating parameters. (authors)

  5. Arrangement of permanent magnet and reed switches for control rod position indicator of SMART CEDM

    International Nuclear Information System (INIS)

    Yoo, J. Y.; Kim, J. I.; Kim, J. H.; Hur, H.; Jang, M. H.

    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. A survey 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. The control rod position indicator having the high performance for SMART was developed on the basis of RSPT technology identified through the survey. The arrangement of permanent magnet and reed switches is the most important procedure in the design of control rod position indication. In this study, the characteristics of permanent magnet and reed switches are introduced and the calculation method for arrangement of permanent magnet and reed switch is presented

  6. 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)

  7. Design of magnetic flux concentrator of permancent magnet for control rod position indicator of SMART CEDM

    International Nuclear Information System (INIS)

    Yoo, J. Y.; Kim, J. H.; Hur, H.; Kim, J. I.

    2002-01-01

    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. A survey 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. The control rod position indicator having the high performance for SMART was developed on the basis of RSPT technology identified through the survey. The arrangement of permanent magnet and reed switches is the most important procedure in the design of control rod position indication. In this study, the magnetic flux concentrator of permanent magnet is introduced and the calculation method for effective flux area for reed switch is presented

  8. 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

  9. 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)

  10. 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

  11. 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.

  12. Simulation of the control rod drop under seismic excitations. Experimental program

    International Nuclear Information System (INIS)

    Chaudat, Th.

    2001-01-01

    This paper describes the experimental program that will be performed at the end of 1998 at the CEA Saclay on a specially constructed analytical mock-up of a control rod. The purpose of these tests is to partially validate the current methodology of the drop time numerical calculations of a PWR (pressurized water reactor) control rod under seismic excitations. The French nuclear partners (EDF and FRAMATOME) are involved in this program. (author)

  13. 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

  14. 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.)

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

    International Nuclear Information System (INIS)

    Tokuhiro, Akira T.; Vadakattu, Shreekanth

    2005-01-01

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

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

    International Nuclear Information System (INIS)

    Takahashi, Ken.

    1995-01-01

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

  17. Method of fabricating a poision tube for reactor control rods

    Energy Technology Data Exchange (ETDEWEB)

    Mori, Yasuhiko; Yoshida, Toshimi; Masaoka, Isao; Naruse, Akisuke

    1983-04-28

    A method to unify the neutron absorbing performance, enhance the workability in the insertion of neutron absorber tube and further decrease the stresses acting on the neutron absorber coating tube is described. The neutron absorber coated rod comprising neutron absorbing substance and a metal pipe is fabricated by compressing a metal pipe filled with the neutron absorber. Specifically, neutron absorbing substance such as boron carbide powder or the like is filled in a metal pipe such as made of stainless steel tube by way of vibration packing or the like. Then, after heating the metal pipe, it is applied with compression working such as swaging into a fine tube to increase the packing density of the absorbing substance filled in the pipe to greater than 60% of the theoretical density and completely contacted closely to the inner wall of the pipe. The neutron absorber coated rod thus fabricated can be inserted to an external coating tube with ease at a predetermined gap.

  18. Device for driving control rods in a reactor

    International Nuclear Information System (INIS)

    Mizumura, Yasuhiro.

    1975-01-01

    Object: To lock and release scram rods by means of a notch and latch system and effect upward movement thereof by means of a screw shaft, the scramming operation being effected at a high speed, the adjusting shim being in inching mode. Structure: When a scram bar is moved toward outside by an actuator through a pin, the scram pin is disengaged from a scram guide and the guide moves down to disengage a latch from a notch and as a result, the scram rod is accelerated by a spring to be moved down, after which making of contact between a bellview washer and a shock stopper and making of contact between a snapper and a scram stopper cause a buffer condition to effect the scram operation. When the screw is rotated by a motor, the slider moves down to allow the reset latch to contact with the reset contact pin so that the latch comes into engagement with the notch to slowly move the scram rod upwardly. (Kamimura, M.)

  19. 3-Dimensional Methodology for the Control Rod Ejection Accident Analysis Using UNICORN{sup TM}

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Chan-su; Um, Kil-sup; Ahn, Dawk-hwan [Korea Nuclear Fuel Company, Taejon (Korea, Republic of); Kim, Yo-han; Sung, Chang-kyung [KEPRI, Taejon (Korea, Republic of); Song, Jae-seung [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2006-07-01

    The control rod ejection accident has been analyzed with STRIKIN-II code using the point kinetics model coupled with conservative factors to address the three dimensional aspects. This may result in a severe transient with very high fuel enthalpy deposition. KNFC, under the support of KEPRI and KAERI, is developing 3-dimensional methodology for the rod ejection accident analysis using UNICORNTM (Unified Code of RETRAN, TORC and MASTER). For this purpose, 3-dimensional MASTER-TORC codes, which have been combined with the dynamic-link library by KAERI, are used in the transient analysis of the core and RETRAN code is used to estimate the enthalpy deposition in the hot rod.

  20. 3-Dimensional Methodology for the Control Rod Ejection Accident Analysis Using UNICORNTM

    International Nuclear Information System (INIS)

    Jang, Chan-su; Um, Kil-sup; Ahn, Dawk-hwan; Kim, Yo-han; Sung, Chang-kyung; Song, Jae-seung

    2006-01-01

    The control rod ejection accident has been analyzed with STRIKIN-II code using the point kinetics model coupled with conservative factors to address the three dimensional aspects. This may result in a severe transient with very high fuel enthalpy deposition. KNFC, under the support of KEPRI and KAERI, is developing 3-dimensional methodology for the rod ejection accident analysis using UNICORNTM (Unified Code of RETRAN, TORC and MASTER). For this purpose, 3-dimensional MASTER-TORC codes, which have been combined with the dynamic-link library by KAERI, are used in the transient analysis of the core and RETRAN code is used to estimate the enthalpy deposition in the hot rod

  1. 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....

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

    Energy Technology Data Exchange (ETDEWEB)

    Bjoereus, K

    1969-07-01

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

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

    International Nuclear Information System (INIS)

    Taswanda Taryo

    2004-01-01

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

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

    International Nuclear Information System (INIS)

    Garcia M, T.; Cruz G, H. S.; Ruiz C, M. A.; Angeles C, A.

    2014-10-01

    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)

  5. 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

  6. Tri-code inductance control rod position indicator with several multi-coding-bars

    International Nuclear Information System (INIS)

    Shi Jibin; Jiang Yueyuan; Wang Wenran

    2004-01-01

    A control rod position indicator named as tri-code inductance control rod position indicator with multi-coding-bars, which possesses simple structure, reliable operation and high precision, is developed. The detector of the indicator is composed of K coils, a compensatory coil and K coding bars. Each coding bar consists of several sections of strong magnetic cores, several sections of weak magnetic cores and several sections of non-magnetic portions. As the control rod is withdrawn, the coding bars move in the center of the coils respectively, while the constant alternating current passes the coils and makes them to create inductance alternating voltage signals. The outputs of the coils are picked and processed, and the tri-codes indicating rod position can be gotten. Moreover, the coding principle of the detector and its related structure are introduced. The analysis shows that the indicator owns more advantage over the coils-coding rod position indicator, so it can meet the demands of the rod position indicating in nuclear heating reactor (NHR). (authors)

  7. Calculation of Excore Detector Responses upon Control Rods Movement in PGSFR

    Energy Technology Data Exchange (ETDEWEB)

    Ha, Pham Nhu Viet; Lee, Min Jae; Kang, Chang Moo; Kim, Sang Ji [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    The Prototype Generation-IV Sodium-cooled Fast Reactor (PGSFR) safety design concept, which aims at achieving IAEA's safety objectives and GIF's safety goals for Generation-IV reactor systems, is mainly focused on the defense in depth for accident detection, prevention, control, mitigation and termination. In practice, excore neutron detectors are widely used to determine the spatial power distribution and power level in a nuclear reactor core. Based on the excore detector signals, the reactor control and protection systems infer the corresponding core power and then provide appropriate actions for safe and reliable reactor operation. To this end, robust reactor power monitoring, control and core protection systems are indispensable to prevent accidents and reduce its detrimental effect should one occur. To design such power monitoring and control systems, numerical investigation of excore neutron detector responses upon various changes in the core power level/distribution and reactor conditions is required in advance. In this study, numerical analysis of excore neutron detector responses (DRs) upon control rods (CRs) movement in PGSFR was carried out. The objective is to examine the sensitivity of excore neutron detectors to the core power change induced by moving CRs and thereby recommend appropriate locations to locate excore neutron detectors for the designing process of the PGSFR power monitoring systems. Section 2 describes the PGSFR core model and calculation method as well as the numerical results for the excore detector spatial weighting functions, core power changes and detector responses upon various scenarios of moving CRs in PGSFR. The top detector is conservatively safe because it overestimated the core power level. However, the lower and bottom detectors still functioned well in this case because they exhibited a minor underestimation of core power of less than ∼0.5%. As a secondary CR was dropped into the core, the lower detector was

  8. 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.

  9. Study on the transient behaviours of MNSR reactor for control rod withdrawal

    International Nuclear Information System (INIS)

    Yang Shunhai

    1995-10-01

    The transient behaviours of Miniature Neutron Source Reactor MNSR are analyzed and calculated with the reactor thermohydraulics RETRAN-02 program and the reactor physics MARIA program. The obtained event sequence and consequence from the calculation are compared with the experiments. The effective resonance integral for study on Doppler effect is taken into account. The reactivity temperature coefficient weighting factors are computed. The transient parameters related to reactor power peaking, coolant inlet temperatures, outlet temperatures and coolant mass flow, etc. are computed and compared with the experimental results. (6 refs., 2 figs., 5 tabs.)

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

    International Nuclear Information System (INIS)

    Kao, P.-W.; Lin, L.-S.; Yang, J.-T.

    2004-01-01

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

  11. Transient flow analysis of the single cylinder for the control rod hydraulic driving system

    International Nuclear Information System (INIS)

    Sun, Xinming; Qin, Benke; Bo, Hanliang

    2017-01-01

    Highlights: • The control rod hydraulic driving system(CRHDS) is a new type of built-in control rod drive technology. The hydraulic cylinder is the main component of the CRHDS. • Transient flow phenomenon in the CRHDS is studied by experiments under different working conditions. • The working mechanism of the hydraulic cylinder step motion and the key characteristic parameters are analyzed based on the experimental results. - Abstract: The control rod hydraulic driving system (CRHDS) is a new type of built-in control rod drive technology. In the CRHDS the pulse flow from the pump into the hydraulic cylinder of the control rod hydraulic drive mechanism (CRHDM) is regulated by the integrated valve to perform the step motion of the reactor control rod. Transient flow occurs in the CRHDS during control rod step motion process which is studied by experiments. The time-history curves of flow rate, pressure and inner cylinder displacement were analyzed, and the results show that the water hammer pressure peak during the step-up motion is high, while there are no obvious pressure fluctuations in the corresponding step-down motion. In the step-up process, the pressure fluctuation amplitude increases with the increase of CRHDS driving pressure. The step-up time and the pressure increasing time before step-up decreases with the driving pressure. The step-up pressure increases with the driving pressure. In the step-down process, the step-down time, the step-down pressure and the pressure decreasing time before step-down do not change with the increase of the driving pressure. The experimental results lay the base for the working principle and vibration reduction analysis of the CRHDS and it’s also helpful for improvement of the working performance of the key facilities and instruments of the CRHDS loop.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

  13. Transient Analysis of a Gas-cooled Fast Reactor for Single Control Assembly Withdrawal

    International Nuclear Information System (INIS)

    Choi, Hangbok

    2014-01-01

    The Energy Multiplier Module (EMZ) system response has been evaluated for control assembly withdrawal transients. Currently the EM2 core is equipped with six cylindrical drum-type control assemblies in the reflector zone for excess reactivity control and power maneuvering during the operating core life. This study investigates the system response to the control assembly withdrawal accident with various rotational speeds and reactivity worth to determine feasible control assembly design requirements from the physics viewpoint. The simulations have been conducted for single control assembly withdrawal transients without scram by a gas-cooled reactor plant simulator, which is based on a simplified plant nodal model, including the point reactor kinetics, single channel core thermal-fluid model, and a turbo-machinery performance model. Simulations were conducted for the middle-of- cycle core, when the excess reactivity of the core is the highest. Control assembly withdrawal times were varied from 1 (runaway) to 180 sec and reactivity worth was varied from 100 to 400 pcm. For a single control assembly withdrawal, the simulation has shown that the peak fuel temperature is expected to be ~1820°C when the assembly worth is 200 pcm and the runaway time is 1 sec per 180 degree rotation. The peak temperature could be reduced to ~1780°C if the assembly is rotated out in a moderate speed such as 1 degree/sec. These peak temperatures give a thermal margin of 22 to 24% to the melting point of uranium carbide fuel. The results also indicate that the current design with a single control assembly worth of 314 pcm may need adjustments in the future design. (author)

  14. 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.

  15. 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.)

  16. 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

  17. Phenobarbital for acute alcohol withdrawal: a prospective randomized double-blind placebo-controlled study.

    Science.gov (United States)

    Rosenson, Jonathan; Clements, Carter; Simon, Barry; Vieaux, Jules; Graffman, Sarah; Vahidnia, Farnaz; Cisse, Bitou; Lam, Joseph; Alter, Harrison

    2013-03-01

    Acute alcohol withdrawal syndrome (AAWS) is encountered in patients presenting acutely to the Emergency Department (ED) and often requires pharmacologic management. We investigated whether a single dose of intravenous (i.v.) phenobarbital combined with a standardized lorazepam-based alcohol withdrawal protocol decreases intensive care unit (ICU) admission in ED patients with acute alcohol withdrawal. This was a prospective, randomized, double-blind, placebo-controlled study. Patients were randomized to receive either a single dose of i.v. phenobarbital (10 mg/kg in 100 mL normal saline) or placebo (100 mL normal saline). All patients were placed on the institutional symptom-guided lorazepam-based alcohol withdrawal protocol. The primary outcome was initial level of hospital admission (ICU vs. telemetry vs. floor ward). There were 198 patients enrolled in the study, and 102 met inclusion criteria for analysis. Fifty-one patients received phenobarbital and 51 received placebo. Baseline characteristics and severity were similar in both groups. Patients that received phenobarbital had fewer ICU admissions (8% vs. 25%, 95% confidence interval 4-32). There were no differences in adverse events. A single dose of i.v. phenobarbital combined with a symptom-guided lorazepam-based alcohol withdrawal protocol resulted in decreased ICU admission and did not cause increased adverse outcomes. Copyright © 2013 Elsevier Inc. All rights reserved.

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

    International Nuclear Information System (INIS)

    Grove, E.; Gunther, W.

    1991-01-01

    Though mechanically different, the control rod drive (CRD) systems used at both Combustion Engineering (CE) and Babcock and Wilcox (B and W) plants position the control rod assemblies (CRA) in the core in response to automatic or manual reactivity control signals. Both systems are also designed to provide a rapid insertion of the CRAs upon a loss of AC power. The CRD system consists of the actual drive mechanisms, power and control, rod position indication, and cooling system components. This aging evaluation included the individual absorber rods, and the fuel assembly and upper internal guide tubes, since failure of these components could preclude the insertion of the control assemblies. Aging and environmental degradation have resulted in system and component failures. Many of these failures caused dropped or slipped rods which adversely affected plant operations by resulting in power reductions, scrams, and safety system actuation. No CRD system failure has ever resulted in the inability to shut down a reactor. However, unplanned, automatic trips challenge the operation of the plants safety systems. Consequently, their occurrence represents a potentially significant increase in plant risk. System and component failures have resulted in four Information Notices during the past decade

  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; 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

  20. Development of control rod position indicator using seismic-resistance reed switches for integral reactor

    International Nuclear Information System (INIS)

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

    2008-01-01

    The Reed Switch Position Transmitter (RSPT) is used as a position indicator for the control rod in commercial nuclear power plants made by ABB-CE. But this position indicator has some problems when directly adopting it to the integral reactor. 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. Therefore the resolution of the position indicator for the integral reactor should be achieved to sense the position of the control rod more precisely than that of the RSPT of the ABB-CE. This paper adopts seismic resistance reed switches to the position indicator in order to reduce the damages or impacts during the handling of the position indicator and earthquake

  1. 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

  2. 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.)

  3. Investigation of a frequency dependent transfer function and its application to control rod localization

    International Nuclear Information System (INIS)

    Garis, N.S.; Pazsit, I.

    1997-02-01

    Control rod vibrations can be detected via the fluctuations they generate in the neutron flux, i.e. the neutron noise. In a previous paper, a neural network-based algorithm for locating a vibrating control rod from the measured neutron noise was developed. The transfer function used for the core model was based on the so called power-reactor approximation resulting in a simple, real-valued solution which means that the phase delay of the signal propagation is neglected. In the present work a more realistic transfer function is used, without the approximations of the previous model. The transfer function is calculated from the Fourier transformed diffusion equation with a complex, frequency dependent buckling leading to a complex solution. In physical terms, this means that the phase delay of the signal propagation is accounted for. Using such a complex core model, the present paper investigates the effectiveness of applying neural networks for control rod localisation. 7 refs, 4 figs

  4. 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)

  5. Optimization of in-core fuel management and control rod strategy in equilibrium fuel cycle

    International Nuclear Information System (INIS)

    Sekimizu, Koichi

    1975-01-01

    An in-core fuel management problem is formulated for the equilibrium fuel cycle in an N-region nuclear reactor model. The formulation shows that the infinite multiplication factor k infinity requisite for newly charged fuel can be separated into two terms - one corresponding to the average k infinity at the end of the cycle and the other representing the direct contribution of the shuffling scheme and control rod programming. This formulation is applied to a three-region cylindrical reactor to obtain simultaneous optimization of shuffling and control rod programming. It is demonstrated that this formulation aids greatly in gaining a better understanding of the effects of changes in the shuffling scheme and control rod programming on equilibrium fuel cycle performance. (auth.)

  6. 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)

  7. Simulated experimental research on flow field near control rod guide tubes

    International Nuclear Information System (INIS)

    Yu Ping'an; Shen Xiuzhong; Yang Guanyue; He Fangzheng; Gao Weiguo; Zhang Zhiyi; Tian Ji'an

    1997-01-01

    The paper presents the velocity measurement in the 1/4 scale transparent model of PWR pressure vessel upper plenum of 300 MW nuclear power plant by employing dynamic resistance strain foil velocity measurement technology and laser Doppler velocity measurement technology which have no effect on the flow field. In the experiment water is chosen as the fluid. As a result of the measurement the hydraulic load on the control rods is clarified and the experimental basis is provided for the analysis of whether the control rods are moving upward and downward freely and drop rapidly in emergency case by order. Meantime it also provides the experimental basis for the optical design of the control rod guide tubes and bundles

  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. Mechanical reliability of falling systems. Case of control rods of nuclear reactors

    International Nuclear Information System (INIS)

    Rezkallah, B.

    1991-05-01

    After having briefly recalled the main principles adopted for nuclear reactor safety (barrier based method, emergency stop system, emergency cooling system) and how control rod operation anomalies are described, this research thesis first gives an overview of knowledge in the field of reliability (principles, qualitative and quantitative methods, failure trees, performance function, shock modelling and mode analysis). Then, the author presents the modelling of the drop and rotation of a control rod. The computation technique is the modal synthesis. The author highlights the influence of physical (friction, initial conditions) and numerical (modal truncation) parameters. He reports a probabilistic analysis of the problem and proposes a stochastic model for the different shock forces (notably friction). He compares the results with those obtained with the DEMT software. The last part reports the investigation of a rare case: the jamming or untimely stop of the control rod device

  10. Effect of CPAP Withdrawal on BP in OSA: Data from Three Randomized Controlled Trials.

    Science.gov (United States)

    Schwarz, Esther I; Schlatzer, Christian; Rossi, Valentina A; Stradling, John R; Kohler, Malcolm

    2016-12-01

    Based on meta-analyses, the BP-lowering effect of CPAP therapy in patients with OSA is reported to be approximately 2 to 3 mm Hg. This figure is derived from heterogeneous trials, which are often limited by poor CPAP adherence, and thus the treatment effect may possibly be underestimated. We analyzed morning BP data from three randomized controlled CPAP withdrawal trials, which included only patients with optimal CPAP compliance. Within the three trials, 149 patients with OSA who were receiving CPAP were randomized to continue therapeutic CPAP (n = 65) or to withdraw CPAP (n = 84) for 2 weeks. Morning BP was measured at home before and after sleep studies in the hospital. CPAP withdrawal was associated with a return of OSA (apnea-hypopnea index [AHI] at a baseline of 2.8/h and at follow-up of 33.2/h). Office systolic BP (SBP) increased in the CPAP withdrawal group compared with the CPAP continuation group by +5.4 mm Hg (95% CI, 1.8-8.9 mm Hg; P = .003) and in the home SBP group by +9.0 mm Hg (95% CI, 5.7-12.3 mm Hg; P CPAP withdrawal results in a clinically relevant increase in BP, which is considerably higher than in conventional CPAP trials; it is also underestimated when office BP is used. Greater OSA severity is associated with a higher BP rise in response to CPAP withdrawal. ClinicalTrials.gov; No.: NCT01332175 and NCT01797653) URL: www.clinicaltrials.gov and ISRCTN registry (ISRCTN 93153804) URL: http://www.isrctn.com/. Copyright © 2016 American College of Chest Physicians. Published by Elsevier Inc. All rights reserved.

  11. Characterisation of reactor control rod drives. Specification 1-6. Reaktorstellstabantriebe. Typenblaetter 1-6

    Energy Technology Data Exchange (ETDEWEB)

    1975-03-01

    The committee 'Kernreaktorregelung' of VDI/VDE-Gesellschaft Mess- und Regelungstechnik has developed 6 specifications (Typenblaetter) of reactor control rod drives. The specifications are aimed at giving engineers in reactor control systems an outline concerning the function as well as some construction characteristics. (orig./LN).

  12. Nuclear reactor control device by vertical displacement of neutron absorber scram rods

    International Nuclear Information System (INIS)

    Defaucheux, Jacques; Pasqualini, Gilbert; Wiart, Albert; Martin, Jean.

    1981-01-01

    Nuclear reactor control system by vertical displacement of an assembly absorbing the neutrons inside a reactor core and drop of the absorbing assembly in maximum insertion position under the effect of its own weight for emergency shutdown. The absorbing assembly is secured to the bottom end of a vertical control rod, the displacement of which is actuated by an electro-magnetic device [fr

  13. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-11-01

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

  15. 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)

  16. 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

  17. 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)

  18. Power peak in vicinity of WWER-440 control rod at end of fuel cycle

    International Nuclear Information System (INIS)

    Mikus, J.

    2003-01-01

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

  19. 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)

  20. 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.

  1. 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

  2. Status of fast reactor control rod development in the United Kingdom

    International Nuclear Information System (INIS)

    Kelly, B.T.

    1984-01-01

    The two large fast reactors constructed in the United Kingdom, that is the Dounreay Fast Reactor (DFR) and the Prototype Fast Reactor (PFR) differed substantially in their control systems. DFR was controlled by variation of the neutron leakage from the core while PFR uses conventional control rods containing neutron absorbing materials. This paper describes the development of the PFR control systems, the progressive design of the control systems for the prototype Civil Fast Reactor (CFR) and the supporting research and development programmes. (author)

  3. Enhanced thermal expansion control rod drive lines for improving passive safety of fast reactors

    International Nuclear Information System (INIS)

    Edelmann, M.; Baumann, W.; Kuechle, M.; Kussmaul, G.; Vaeth, W.; Bertram, A.

    1992-01-01

    The paper presents a device for increasing the thermal expansion effect of control rod drive lines on negative reactivity feedback in fast reactors. The enhanced thermal expansion of this device can be utilized for both passive rod drop and forced insertion of absorbers in unprotected transients, e.g. ULOF. In this way the reactor is automatically brought into a permanently subcritical state and temperatures are kept well below the boiling point of the coolant. A prototype of such a device called ATHENa (German: Shut-down by THermal Expansion of Na) is presently under construction and will be tested. The paper presents the principle, design features and thermal properties of ATHENs as well as results of reactor dynamics calculations of ULOF's for EFR with enhanced thermal expansion control rod drive lines. (author)

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

    International Nuclear Information System (INIS)

    Hozer, Z.; Nagy, I.; Windberg, P.; Balasko, M.; Matus, L.; Prokopiev, O.; Pinter, A.; Horvath, M.; Gyenes, Gy.; Czitrovszky, A.; Nagy, A.; Jani, P.

    2003-11-01

    The CODEX-B4C 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 high temperature degradation of a VVER-1000 type bundle with B 4 C control rod was investigated with electrically heated fuel rods. The experiment was carried out according to a scenario selected in favour of methane formation. Degradation of control rod and fuel bundle took place at temperatures ∼2000 deg C, cooling down of the bundle was performed in steam atmosphere. The gas composition measurement indicated no methane production during the experiment. High release of aerosols was detected in the high temperature oxidation phase. The on-line measured data are collected into a database and are available for code validation and development. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Hozer, Z; Nagy, I; Windberg, P; Balasko, M; Matus, L; Prokopiev, O; Pinter, A; Horvath, M; Gyenes, Gy [KFKI Atomic Energy Research Institute, Budapest (Hungary); Czitrovszky, A; Nagy, A; Jani, P [Research Institute for Solid State Physics and Optics, Budapest (Hungary)

    2003-11-01

    The CODEX-B4C bundle test has been successfully performed on 25{sup th} May 2001 in the framework of the COLOSS project of the EU 5{sup th} FWP. The high temperature degradation of a VVER-1000 type bundle with B{sub 4}C control rod was investigated with electrically heated fuel rods. The experiment was carried out according to a scenario selected in favour of methane formation. Degradation of control rod and fuel bundle took place at temperatures {approx}2000 deg C, cooling down of the bundle was performed in steam atmosphere. The gas composition measurement indicated no methane production during the experiment. High release of aerosols was detected in the high temperature oxidation phase. The on-line measured data are collected into a database and are available for code validation and development. (author)

  6. Degradation in steam of 60 cm-long B{sub 4}C control rods

    Energy Technology Data Exchange (ETDEWEB)

    Dominguez, C., E-mail: christina.dominguez@irsn.fr; Drouan, D.

    2014-08-01

    In the framework of nuclear reactor core meltdown accident studies, the degradation of boron carbide control rod segments exposed to argon/steam atmospheres was investigated up to about 2000 °C in IRSN laboratories. The sequence of the phenomena involved in the degradation has been found to take place as expected. Nevertheless, the ZrO{sub 2} oxide layer formed on the outer surface of the guide tube was very protective, significantly delaying and limiting the guide tube failure and therefore the boron carbide pellet oxidation. Contrary to what was expected, the presence of the control rod decreases the hydrogen release instead of increasing it by additional oxidation of boron compounds. Boron contents up to 20 wt.% were measured in metallic mixtures formed during degradation. It was observed that these metallic melts are able to attack the surrounding fuel rods, which could have consequences on fuel degradation and fission product release kinetics during severe accidents.

  7. 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)

  8. Experimental Study of Hydraulic Control Rod Drive Mechanism for Passive IN-core Cooling System of Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-15

    CAREM 25 (27 MWe safety systems using hydraulic control rod drives (CRD) studied critical issues that were rod drops with interrupted flow [3]. Hydraulic control rod drive suggested fast shutdown condition using a large gap between piston and cylinder in order to fast drop of neutron absorbing rods. A Passive IN-core Cooling system (PINCs) was suggested for safety enhancement of pressurized water reactors (PWR), small modular reactor (SMR), sodium fast reactor (SFR) in UNIST. PINCs consist of hydraulic control rod drive mechanism (Hydraulic CRDM) and hybrid control rod assembly with heat pipe combined with control rod. The schematic diagram of the hydraulic CRDM for PINCs is shown in Fig. 1. The experimental results show the steady state and transient behavior of the upper cylinder at a low pressure and low temperature. The influence of the working fluid temperature and cylinder mass are investigated. Finally, the heat removal between evaporator section and condenser section is compared with or without the hybrid control rod. Heat removal test of the hybrid heat pipe with hydraulic CRDM system showed the heat transfer coefficient of the bundle hybrid control rod and its effect on evaporator pool. The preliminary test both hydraulic CRDM and heat removal system was conducted, which showed the possibility of the in-core hydraulic drive system for application of PINCs.

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

    International Nuclear Information System (INIS)

    Rodenas, J.; Abarca, A.; Gallardo, S.

    2011-01-01

    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.

  10. 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)

  11. 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)

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

    International Nuclear Information System (INIS)

    Sepold, L.; Lind, T.; Csordas, A. Pinter; Stegmaier, U.; Steinbrueck, M.; Stuckert, J.

    2009-01-01

    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 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 of H 2

  13. Monte Carlo verification of control-rod worth for the Savannah River K reactor

    International Nuclear Information System (INIS)

    Mosteller, R.D.

    1992-01-01

    The Savannah River K Reactor is a heavy-water reactor that relies on control-rod movement to control its reactivity and power distribution during normal operations. It is necessary, therefore, to have an accurate estimate of the reactivity worth of its control rods in order to predict the behavior of the reactor. Westinghouse Savannah River Company (WSRC) uses the GLASS lattice-physics code to calculate few-group cross sections for fuel and control-rod assemblies in the K reactor. This paper compares the control-rod worth calculated by GLASS to that calculated by the MCNP Monte Carlo program. The GLASS calculations utilize its standard 37-group cross-section library, while the MCNP calculations employ continuous-energy isotopic cross-section libraries derived from ENDF/B-V. The MCNP calculations therefore combine the most rigorous analytical model and the most accurate cross sections currently available for thermal-reactor analysis. Consequently, the MCNP results comprise a computational benchmark against which the accuracy of the GLASS code can be evaluated

  14. A new physics design of control safety rods for prototype fast breeder reactor

    International Nuclear Information System (INIS)

    Devan, K.; Riyas, A.; Alagan, M.; Mohanakrishnan, P.

    2008-01-01

    The absorber rods of 500 MWe prototype fast breeder reactor (PFBR), which is under construction at Kalpakkam, have been designed to provide sufficient shutdown margin during normal and accidental conditions for ensuring the safe shut down. There are nine control and safety rods (CSR) and 3 diverse safety rods (DSR). Absorber material used is initially 65% enriched B 4 C. Based on the reported experiments in PHENIX reactor and design of absorber rods in SUPERPHENIX, the design of CSR is modified by introducing 20 cm length natural B 4 C at the top and bottom of absorber column and maintaining the remaining portion with 65% enriched B 4 C. This design ensures sufficient shutdown margin (SDM) during normal operation and also during the one stuck rod condition. For comparison of the above two designs, a CSR of 57% of enrichment was considered which gives the same worth as the revised CSR design with natural B 4 C sections in top and bottom. There is significant savings in the initial inventory of enriched B 4 C for CSR. The annual requirement of enriched boron also reduces. This new CSR can last for about 5 cycles, based on its clad life. But, it is planned to be replaced after every 3 cycles (1 cycle equals 180 efpd) of operation due to radiation damage effects in hexcan D9 steel. Use of ferritic steel for hexcan can extend the life of CSR to 5 cycles

  15. Analysis of a control rod ejection transient in a mox-fuelled PWR

    International Nuclear Information System (INIS)

    Lenain, R.; Mathonniere, G.; Perrutel, J.P.; Schaeffer, H.; Stelletta, S.; Lam Hime, M.

    1988-09-01

    The decision to use mixed-oxide (MOX) fuel in PWR's involved re-investigation of a certain number of accidents and notably control rod ejection transients. It has thus been shown that this accident would be no more severe than in the case of all-uranium cores, since the positive effects on the ejected rod worth would counterbalance the negative effects on the delayed neutron fraction. A new approach to the kinetics aspect of the calculation method for this accident is also presented, involving a 3-D kinetic calculation with only a few axial meshes

  16. On the neutron noise diagnostics of pressurized water reactor control rod vibrations II. Stochastic vibrations

    International Nuclear Information System (INIS)

    Pazsit, I.; Glockler, O.

    1984-01-01

    In an earlier publication, using the theory of neutron fluctuations induced by a vibrating control rod, a complete formal solution of rod vibration diagnostics based on neutron noise measurements was given in terms of Fourier-transformed neutron detector time signals. The suggested procedure was checked in numerical simulation tests where only periodic vibrations could be considered. The procedure and its numerical testing are elaborated for stochastic two-dimensional vibrations. A simple stochastic theory of two-dimensional flow-induced vibrations is given; then the diagnostic method is formulated in the stochastic case, that is, in terms of neutron detector auto- and crosspower spectra. A previously suggested approximate rod localization technique is also formulated in the stochastic case. Applicability of the methods is then investigated in numerical simulation tests, using the proposed model of stochastic two-dimensional vibrations when generating neutron detector spectra that simulate measured data

  17. Cutting of Control Rod Guide Tubes and Jet Pumps at Wuergassen NPP/ Germany

    International Nuclear Information System (INIS)

    Ralf Loock

    2006-01-01

    KWW Nuclear Power Plant has been erected and put into operation in the time period 1968 to end of 1971. In 23 years of operation about 73 billions kWh electric power have been generated. The electric rated power was 670 MW. In May 1995 the decommission was started for economical reasons, since April 1997 the demolition is in progress. In autumn 2003 FANP has been called upon to cut the Control Rod Guide Tubes and Jet Pumps. After a 10 months preparatory phase, cutting of the Control Rod Guide Tubes and Jet Pumps was started in September 2004. For radiation protection reasons the Control Rod Guide Tubes and Jet Pumps were cut and packed remote controlled under water. The cutting process had been selected particularly under consideration of radiation protection aspects. A combination of the technologies band sawing and nibbling was assigned for cutting of the Control Rod Guide Tubes and Jet Pumps. The band saw consists of a substructure and of a feed system installed on it. The feed system can be equipped with one or two band saws optionally. A turntable is integrated in the substructure. It is used to support the CRGT respectively the residual piece of the CRGT and to move it out of the cut position after finishing the cutting. After getting the agreement to the specification, the equipment has been designed and manufactured according to the specification preferences. After finishing the internal testing, the factory acceptance test as well as acceptance and functional tests of every facility were carried out. After cold testing being performed successfully, the cutting equipment was installed in the NPP and checked for operational safety additionally. Cutting of the Control Rod Guide Tubes and Jet Pumps has been carried out on two working places inside the fuel element storage pool. During the total project duration on site the processing has been supervised by the radiation protection staff permanently. As a result 110 Control Rod Guide Tubes and 18 Jet Pumps were

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

    International Nuclear Information System (INIS)

    Beran, Tâm; Seltborg, Per; Lindahl, Sten-Örjan; Bieli, Roger; Ledergerber, Guido

    2014-01-01

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

  19. BWR feedwater nozzle and control-rod-drive return line nozzle cracking

    International Nuclear Information System (INIS)

    Anon.

    1981-01-01

    In its 1978 Annual Report to Congress, the Nuclear Regulatory Commission identified as an unresolved safety issue the appearance of cracks in feedwater nozzles at boiling-water reactors (BWRs). Later similar cracking, detected in return water lines for control-rod-drive systems at BWRs, was designated Part II of the issue. This article outlines the resolution of these cracking problems

  20. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Weissglas, P [The Swedish State Power Board, Stockholm (Sweden)

    1960-11-15

    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.

  2. Enhancement of control rod drive mechanism seating position detector for JRR-3

    International Nuclear Information System (INIS)

    Ohuchi, Satoshi; Kurumada, Osamu; Kamiishi, Eigo; Sato, Masayuki; Ikekame, Yoshinori; Wada, Shigeru

    2016-06-01

    The purpose of the control rod drive mechanism seating position detector for JRR-3 is one of methods for confirming the shutdown condition of the reactor and sending out the seat position signal to other systems. The detector has been utilizing more than 25 years with maintenance regularly. However, some troubles occurred recently. Moreover, the detector has already been discontinued, and it is confirmed that the successor detector is unsuitable for the control rod drive mechanism of JRR-3. Therefore, it was necessary to select the adequate detector to the control rod drive mechanism of JRR-3. Accordingly, we built a test device with the aim of verifying several detectors for integrity and function. At the time of the test for performance confirmation, it was occurred unexpected problems. Nevertheless, we devise improvement of the problems and took measures. Thus we were able to make adequate detector for JRR-3 and replace to enhanced detector. This paper reports the Enhanced of Control rod drive mechanism seating position detector. (author)

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

    International Nuclear Information System (INIS)

    Castillo, Alejandro; Ortiz, Juan Jose; Montes, Jose Luis; Perusquia, Raul

    2007-01-01

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

  4. Accidental situations analysis in BN-800 reactor bounded with an untimely ascension of the control rods

    International Nuclear Information System (INIS)

    Gorbunov, V.S.; Zaets, N.P.

    1987-12-01

    In this document the conditions and the results of one or more control rods untimely ascension out of the BN-800 core are examined. The mathematical model which describes the reactor kinetic, the temperature core and the feedback is presented [fr

  5. Procedure and code for calculating black control rods taking into account epithermal absorption, code CAS-1

    International Nuclear Information System (INIS)

    Martinc, R.; Trivunac, N.; Zivkovic, Z.

    1964-12-01

    This report describes the computer code CAS-1, calculation method and procedure applied for calculating the black control rods taking into account the epithermal neutron absorption. Results obtained for supercell method applied for regular lattice reflected in the multiplication medium is part of this report in addition to the computer code manual

  6. 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)

  7. Analysis of water hammer in control rod drive systems of boiling water reactor nuclear power plants

    International Nuclear Information System (INIS)

    Safwat, H.H.; Arastu, A.H.; Lau, S.

    1983-01-01

    The method of characteristics is applied to analyze water hammer in BWR (Boiling Water Reactor) Control Rod Drive (CRD) Systems following fast opening of scram valves. The modelling of the CRD mechanism is presented. Numerical predictions are compared to experimental data. (author)

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

    International Nuclear Information System (INIS)

    Schulz, W.

    2000-01-01

    Operating experience has proved so far that BWR control rods cannot be used for the total reactor life time as originally presumed, but instead has to be considered as a consumable article. After only few operating cycles, the mechanism of absorber failure has been shown to be neutron induced boron carbide swelling and stress cracking of the absorber tubes, followed by erosion of the absorber material. In the case that operation of such a control rod is continued in control cells, this can lead to an increase of the local power density distribution in the core and, under certain conditions, can even cause fuel rod damage. A non destructive testing method has been developed called 'UNDERWATER NEUTRON RADIOGRAPHY' applicable for any BWR control rod. 'Lead-control rods' being radiographed are used to evaluate their actual nuclear worth by the help of a special analytical procedure developed and verified by the author. Nuclear worth data plotted against bum up history data will allow to create an 'EMPIRIC MODEL'. This model includes the basic idea of operating control rods of a certain design first in a control position up to a target fluence limited to an amount just below the appearance of control rod washout. Afterwards they have to be moved in a shut down position to work therefor the total remaining holding period. The initial model is applicable to any CR-design as long as sufficient measuring-data and thus data about the nuclear worth are available. The results of these experiences are extrapolated to the whole reactor holding period. After modelling no further measurements of this particular control rod type are necessary in any reactor. The second focal point is to provide an APPROXIMATION EQUATION. By knowing the absorber radius, B 4 C density and absorber enclosure data an engineer will calculate reliably the working life of any control rod design on control position. indicated as maximum allowable neutron fluence margin until absorber wash-out starts. This

  9. Numerical Analysis on the Free Fall Motion of the Control Rod Assembly for the Sodium Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Se-Hong; Choi, Choengryul; Son, Sung-Man [ELSOLTEC, Yongin (Korea, Republic of); Kim, Jae-Yong; Yoon, Kyung-Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    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. However, because of its complex shape, it is difficult to estimate the drop time by theoretical method. In this study, numerical analysis has been carried out in order to estimate drop time and falling velocity of the control rod assembly to provide the underlying data for the design optimization. Numerical analysis has been carried out to estimate the drop time and falling velocity of the control rod assembly for sodium-cooled fast reactor. Before performing the numerical analysis for the control rod assembly, sphere dropping experiment has been carried out for verification of the CFD methodology. The result of the numerical analysis for the method verification is almost same as the result of the experiment. Falling velocity and drag force increase rapidly in the beginning. And then it goes to the stable state. When the piston head of the control rod assembly is inserted into the damper, the drag force increases instantaneously and the falling velocity decreases quickly. The falling velocity is reduced about 14 % by damper. The total drop time of the control rod assembly is about 1.47s. In the next study, the experiment for the control rod assembly will be carried out, and its result is going to be compared with the CFD analysis result.

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

    International Nuclear Information System (INIS)

    Hendrich, W.R.

    1995-08-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-09-15

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

  12. 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)

  13. 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)

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

    International Nuclear Information System (INIS)

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

    1988-12-01

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

  15. Design of a model predictive load-following controller by discrete optimization of control rod speed for PWRs

    International Nuclear Information System (INIS)

    Kim, Jae Hwan; Park, Soon Ho; Na, Man Gyun

    2014-01-01

    Highlights: • A model predictive controller for load-following operation was developed. • Genetic algorithm optimizes the five nonlinear discrete control rod speeds. • The boron concentration is adjusted with automatic adjustment logic. • The proposed controller reflects the realistic control rod drive mechanism movement. • The performance was confirmed to be satisfactory by simulation from BOC to EOC. - Abstract: Currently, most existing nuclear power plants alter the reactor power by adjusting the boron concentration in the coolant because it has a smaller effect on the reactor power distribution. Frequent control rod movements for load-following operation induce xenon-oscillation. Therefore, a controller that can subdue this phenomenon effectively is needed. At an APR1400 nuclear power plant which is a pressurized water reactor (PWR), the reactor power is controlled automatically using a Reactor Regulating System (RRS) but the power distribution is controlled manually by operators. Therefore, for APR+ nuclear power plants which is an improved version of APR1400 nuclear reactor, a new concept of a reactor controller is needed to control both the reactor power and power distribution automatically. The model predictive control (MPC) method is applicable to multiple-input multiple-output control, and can be applied for complex and nonlinear systems, such as the nuclear power plants. In this study, an MPC controller was developed by applying a genetic algorithm to optimize the discrete control rod speeds and by reflecting the realistic movement of the control rod drive mechanism that moves at only five discrete speeds. The performance of the proposed controller was confirmed to be satisfactory by simulating the load-following operation of an APR+ nuclear power plant through interface with KISPAC-1D code

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1985-07-01

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

  17. 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

  18. 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.

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

    International Nuclear Information System (INIS)

    He Keyu; Han Weishi

    2010-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-11-01

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

  1. Effect of local automatic control rods on three-dimensional calculations of the power distribution in an RBMK

    International Nuclear Information System (INIS)

    Pogosbekyan, L.R.; Lysov, D.A.; Bronitskii, L.L.

    1993-01-01

    Numerical simulators and information systems that support nuclear reactor operators must have fast models to estimate how fuel reloads and control rod displacement affect neutron and power distributions in the core. The consequences of reloads and control rod displacement cannot be evaluated correctly without considering local automatic control-rod operations in maintaining the radial power distribution. Fast three-dimensional models to estimate the effects of reloads and displacement of the control and safety rods have already been examined. I.V. Zonov et al. used the following assumptions in their calculational model: (1) the full-scale problem could be reduced a three-dimensional fragment of a locally perturbed core, and (2) the boundary conditions of the fragment and its total power were constant. The last assumption considers approximately how local automatic control rods stabilize the radial power distribution, but three dimensional calculations with these rods are not considered. These assumptions were introduced to obtain high computational speed. I.L. Bronitskii et al. considered in more detail how moving the local automatic control rods affect the power dimensional in the three-dimensional fragment, because, with on-line monitoring of the reload process, information on control rod positions is periodically renewed, and the calculations are done in real time. This model to predict the three-dimensional power distribution to (1) do a preliminary reload analysis, and (2) prepare the core for reloading did not consider the effect of perturbations from the local automatic control rods. Here we examine a model of a stationary neutron distribution. On one hand it gives results in an acceptable computation time; on the other it is a full-scale three-dimensional model and considers how local automatic control rods affect both the radial and axial power distribution

  2. 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

  3. 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.

  4. 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

  5. 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

  6. 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

  7. Integrity Evaluation of Control Rod Assembly for Sodium-Cooled Fast Reactor due to Drop Impact

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Hyun Seung; Yoon, Kyung Ho; Kim, Hyung Kyu; Cheon, Jin Sik; Lee, Chan Bock [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2017-03-15

    The CA (Control Assembly) of an SFR has a CRA(Control Rod Assembly) with an inner duct and control rod. During an emergency situation, the CRA falls into the duct of the CA for a rapid shut-down. The drop time and impact velocity of the CRA are important parameters with respect to the reactivity insertion time and the structural integrity of the CRA. The objective of this study was to investigate the dynamic behavior and integrity of the CRA owing to a drop impact. The impact analysis of the CRA under normal/abnormal drop conditions was carried out using the commercial FEM code LS-DYNA. Results of the drop impact analysis demonstrated that the CRA maintained structural integrity, and could be safely inserted into the flow hole of the damper under abnormal conditions.

  8. Simulation and operation of the EBR-II automatic control rod drive system

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  9. Neutronic analysis of absorbing materials for the control rod system in reactor ALLEGRO

    Energy Technology Data Exchange (ETDEWEB)

    Cajko, Frantisek; Secansky, Michal; Chrebet, Tomas; Zajac, Radoslav; Darilek, Petr [VUJE, a.s., Trnava (Slovakia)

    2016-09-15

    Experimental reactor ALLEGRO is a gas cooled fast reactor in the design stage. The current design of its reactivity control system is based on control rods filled with boron carbide as the absorber. Because of disadvantages connected to high boron enrichment a possibility of using other absorbent materials was explored to lower the boron enrichment and increase the worth of the control rods. The results of neutronic Monte-Carlo analyses in a computational supercell are presented in this paper. Three absorbent materials most suitable for a use in reactor ALLEGRO (B{sub 4}C, EuB{sub 6} and ReB{sub 2}) have been analysed also in a full core model. A possible benefit of a neutron trap concept is explored as well but materials with satisfactory neutronic properties proved to be not suitable for expected high temperatures in the reactor.

  10. Simulation and operation of the EBR-II automatic control rod drive system

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  11. Integrity Evaluation of Control Rod Assembly for Sodium-Cooled Fast Reactor due to Drop Impact

    International Nuclear Information System (INIS)

    Lee, Hyun Seung; Yoon, Kyung Ho; Kim, Hyung Kyu; Cheon, Jin Sik; Lee, Chan Bock

    2017-01-01

    The CA (Control Assembly) of an SFR has a CRA(Control Rod Assembly) with an inner duct and control rod. During an emergency situation, the CRA falls into the duct of the CA for a rapid shut-down. The drop time and impact velocity of the CRA are important parameters with respect to the reactivity insertion time and the structural integrity of the CRA. The objective of this study was to investigate the dynamic behavior and integrity of the CRA owing to a drop impact. The impact analysis of the CRA under normal/abnormal drop conditions was carried out using the commercial FEM code LS-DYNA. Results of the drop impact analysis demonstrated that the CRA maintained structural integrity, and could be safely inserted into the flow hole of the damper under abnormal conditions.

  12. 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

  13. 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)

  14. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-10-15

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

  16. 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

  17. 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.

  18. Design, construction and simulation of a multipurpose system for precision movement of control rods in nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Shirazi, S.A. Mousavi, E-mail: a_moosavi@azad.ac.i [Department of Physics, Islamic Azad University, South Tehran Branch, Tehran (Iran, Islamic Republic of); Aghanajafi, C. [Department of Mechanical Engineering, K.N.T. University of Technology, Tehran (Iran, Islamic Republic of); Sadoughi, S. [Department of Electrical Engineering, Sharif University of Technology, Tehran (Iran, Islamic Republic of); Sharifloo, N. [Department of Physics, Imam Hossein University, Tehran (Iran, Islamic Republic of)

    2010-12-15

    This article presents the design and implementation of a microcontroller-based system for the automatic movement of control rods in nuclear reactors of either power or research types. This system is controlled automatically, is linked to a personal computer system, and has manual controlling ability as well. The important features of this system are: automatic scram of the control rods, activation of alarm in emergency situations, and the ability to tune the control rod movement course both upwards and downwards. In this system, a small tank has been improvised as a coolant reservoir for pool type reactors such as Tehran Research Reactor and its water level is continuously adjusted by special sensors. Also, this system can be applied for controlling various types of control rods such as the regulating rods, safety rods and shim rods; can be connected to all reactor measurement tools and systems such as the period meter, power meter and flux meter; and can receive feedback signals from them. The devised system can be calibrated with these measurement tools by two special potentiometers in the related electronic board. The processes of this system have been simulated by the SIMULINK tool kit of MATLAB software and all responses of the system, including oscillation and transient responses, have been analyzed.

  19. The variation of the reactivity with the number, diameter and length of the control rods in a heavy water natural uranium reactor

    Energy Technology Data Exchange (ETDEWEB)

    McCriric, H

    1958-05-15

    Starting with the known reactor constants for a heavy water moderated reactor with reflector and a given number of control rods of a certain size, the reactivity equivalence of the control rods is calculated. The calculation is given in detail. The number, length and diameter of the control rods is then varied and the effect of these parameters on the reactivity is shown graphically. Flux plots are also given for the reactor with and without control rods.

  20. Multiple fuel rod gripper

    International Nuclear Information System (INIS)

    Shields, E.P.

    1987-01-01

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