WorldWideScience

Sample records for controlled reactors systems

  1. Reactor control system. PWR

    International Nuclear Information System (INIS)

    2009-01-01

    At present, 23 units of PWR type reactors have been operated in Japan since the start of Mihama Unit 1 operation in 1970 and various improvements have been made to upgrade operability of power stations as well as reliability and safety of power plants. As the share of nuclear power increases, further improvements of operating performance such as load following capability will be requested for power stations with more reliable and safer operation. This article outlined the reactor control system of PWR type reactors and described the control performance of power plants realized with those systems. The PWR control system is characterized that the turbine power is automatic or manually controlled with request of the electric power system and then the nuclear power is followingly controlled with the change of core reactivity. The system mainly consists of reactor automatic control system (control rod control system), pressurizer pressure control system, pressurizer water level control system, steam generator water level control system and turbine bypass control system. (T. Tanaka)

  2. Reactor power control system

    International Nuclear Information System (INIS)

    Tomisawa, Teruaki.

    1981-01-01

    Purpose: To restore reactor-power condition in a minimum time after a termination of turbine bypass by reducing the throttling of the reactor power at the time of load-failure as low as possible. Constitution: The transient change of the internal pressure of condenser is continuously monitored. When a turbine is bypassed, a speed-control-command signal for a coolant recirculating pump is generated according as the internal pressure of the condenser. When the signal relating to the internal pressure of the condenser indicates insufficient power, a reactor-control-rod-drive signal is generated. (J.P.N.)

  3. Reactor limit control system

    International Nuclear Information System (INIS)

    Rubbel, F.E.

    1982-01-01

    The very extensive use of limitations in the operational field between protection system and closed-loop controls is an important feature of German understanding of operational safety. The design of limitations is based on very large activities in the computational field but mostly on the high level of the plant-wide own commissioning experience of a turnkey contractor. Limitations combine intelligence features of closed-loop controls with the high availability of protection systems. (orig.)

  4. Digital control system of advanced reactor

    International Nuclear Information System (INIS)

    Peng Huaqing; Zhang Rui; Liu Lixin

    2001-01-01

    This article produced the Digital Control System For Advanced Reactor made by NPIC. This system uses Siemens SIMATIC PCS 7 process control system and includes five control system: reactor power control system, pressurizer level control system, pressurizer pressure control system, steam generator water level control system and dump control system. This system uses three automatic station to realize the function of five control system. Because the safety requisition of reactor is very strict, the system is redundant. The system configuration uses CFC and SCL. the human-machine interface is configured by Wincc. Finally the system passed the test of simulation by using RETRAN 02 to simulate the control object. The research solved the key technology of digital control system of reactor and will be very helpful for the nationalization of digital reactor control system

  5. The Optimization of power reactor control system

    International Nuclear Information System (INIS)

    Danupoyo, S.D.

    1997-01-01

    A power reactor is an important part in nuclear powered electrical plant systems. Success in controlling the power reactor will establish safety of the whole power plant systems. Until now, the power reactor has been controlled by a classical control system that was designed based on output feedback method. To meet the safety requirements that are now more restricted, the recently used power reactor control system should be modified. this paper describes a power reactor control system that is designed based on a state feedback method optimized with LQG (Linear-quadrature-gaussian) method and equipped with a state estimator. A pressurized-water type reactor has been used as the model. by using a point kinetics method with one group delayed neutrons. the result of simulation testing shows that the optimized control system can control the power reactor more effective and efficient than the classical control system

  6. 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)

  7. Propose Reactor Control and Monitoring System for RTP

    International Nuclear Information System (INIS)

    Mohd Sabri Minhat; Izhar Abu Hussin; Mohd Idris Taib; Mohd Khairulezwan Abdul Manan; Nurfarhana Ayuni Joha

    2011-01-01

    Reactor control and monitoring system is a one of the important features used in reactor. The control and monitoring must come together to provide safety, excellent performance and reliable in nuclear reactor technology application. Objectives of this technical paper are to design and propose reactor control system and reactor monitoring system in Research Reactor (RTP) for Reactor Upgrading Project. (author)

  8. TREAT Reactor Control and Protection System

    International Nuclear Information System (INIS)

    Lipinski, W.C.; Brookshier, W.K.; Burrows, D.R.; Lenkszus, F.R.; McDowell, W.P.

    1985-01-01

    The main control algorithm of the Transient Reactor Test Facility (TREAT) Automatic Reactor Control System (ARCS) resides in Read Only Memory (ROM) and only experiment specific parameters are input via keyboard entry. Prior to executing an experiment, the software and hardware of the control computer is tested by a closed loop real-time simulation. Two computers with parallel processing are used for the reactor simulation and another computer is used for simulation of the control rod system. A monitor computer, used as a redundant diverse reactor protection channel, uses more conservative setpoints and reduces challenges to the Reactor Trip System (RTS). The RTS consists of triplicated hardwired channels with one out of three logic. The RTS is automatically tested by a digital Dedicated Microprocessor Tester (DMT) prior to the execution of an experiment. 6 refs., 5 figs., 1 tab

  9. Distributed expert systems for nuclear reactor control

    International Nuclear Information System (INIS)

    Otaduy, P.J.

    1992-01-01

    A network of distributed expert systems is the heart of a prototype supervisory control architecture developed at the Oak Ridge National Laboratory (ORNL) for an advanced multimodular reactor. Eight expert systems encode knowledge on signal acquisition, diagnostics, safeguards, and control strategies in a hybrid rule-based, multiprocessing and object-oriented distributed computing environment. An interactive simulation of a power block consisting of three reactors and one turbine provides a realistic, testbed for performance analysis of the integrated control system in real-time. Implementation details and representative reactor transients are discussed

  10. ANALYTICAL SYNTHESIS OF CHEMICAL REACTOR CONTROL SYSTEM

    Directory of Open Access Journals (Sweden)

    Alexander Labutin

    2017-02-01

    Full Text Available The problem of the analytical synthesis of the synergetic control system of chemical reactor for the realization of a complex series-parallel exothermal reaction has been solved. The synthesis of control principles is performed using the analytical design method of aggregated regulators. Synthesized nonlinear control system solves the problem of stabilization of the concentration of target component at the exit of reactor and also enables one to automatically transfer to new production using the equipment.

  11. Level controlling system in BWR type reactors

    International Nuclear Information System (INIS)

    Joge, Toshio; Higashigawa, Yuichi; Oomori, Takashi.

    1981-01-01

    Purpose: To reasonably attain fully automatic water level control in the core of BWR type nuclear power plants. Constitution: A feedwater flow regulation valve for reactor operation and a feedwater flow regulation valve for starting are provided at the outlet of a motor-driven feedwater pump in a feedwater system, and these valves are controlled by a feedwater flow rate controller. While on the other hand, a damp valve for reactor clean up system is controlled either in ''computer'' mode or in ''manual'' mode selected by a master switch, that is, controlled from a computer or the ON-OFF switch of the master switch by way of a valve control analog memory and a turn-over switch. In this way, the water level in the nuclear reactor can be controlled in a fully automatic manner reasonably at the starting up and shutdown of the plant to thereby provide man power saving. (Seki, T.)

  12. Advanced liquid metal reactor plant control system

    International Nuclear Information System (INIS)

    Dayal, Y.; Wagner, W.; Zizzo, D.; Carroll, D.

    1993-01-01

    The modular Advanced Liquid Metal Reactor (ALMR) power plant is controlled by an advanced state-of-the-art control system designed to facilitate plant operation, optimize availability, and protect plant investment. The control system features a high degree of automatic control and extensive amount of on-line diagnostics and operator aids. It can be built with today's control technology, and has the flexibility of adding new features that benefit plant operation and reduce O ampersand M costs as the technology matures

  13. Nuclear reactor plants and control systems therefor

    International Nuclear Information System (INIS)

    de Boer, G.A.; de Hex, M.

    1976-01-01

    A nuclear reactor plant is described comprising at least two hydraulically separated but thermally interconnected heat conveying circuits, of which one is the reactor circuit filled with a non-water medium and the other one is the water-steam-circuit equipped with a steam generator, a feed water conduit controlled by a valve and a steam turbine, and a control system mainly influenced by the pressure drop caused in said feed water conduit and its control valve and having a value of at least 10 bars at full load

  14. Distributed computer control system for reactor optimization

    International Nuclear Information System (INIS)

    Williams, A.H.

    1983-01-01

    At the Oldbury power station a prototype distributed computer control system has been installed. This system is designed to support research and development into improved reactor temperature control methods. This work will lead to the development and demonstration of new optimal control systems for improvement of plant efficiency and increase of generated output. The system can collect plant data from special test instrumentation connected to dedicated scanners and from the station's existing data processing system. The system can also, via distributed microprocessor-based interface units, make adjustments to the desired reactor channel gas exit temperatures. The existing control equipment will then adjust the height of control rods to maintain operation at these temperatures. The design of the distributed system is based on extensive experience with distributed systems for direct digital control, operator display and plant monitoring. The paper describes various aspects of this system, with particular emphasis on: (1) the hierarchal system structure; (2) the modular construction of the system to facilitate installation, commissioning and testing, and to reduce maintenance to module replacement; (3) the integration of the system into the station's existing data processing system; (4) distributed microprocessor-based interfaces to the reactor controls, with extensive security facilities implemented by hardware and software; (5) data transfer using point-to-point and bussed data links; (6) man-machine communication based on VDUs with computer input push-buttons and touch-sensitive screens; and (7) the use of a software system supporting a high-level engineer-orientated programming language, at all levels in the system, together with comprehensive data link management

  15. ADAPTIVE CONTROL SYSTEM OF INDUSTRIAL REACTORS

    Directory of Open Access Journals (Sweden)

    Vyacheslav K. Mayevski

    2014-01-01

    Full Text Available This paper describes a mathematical model of an industrial chemical reactor for production of synthetic rubber. During reactor operation the model parameters vary considerably. To create a control algorithm performed transformation of mathematical model of the reactor in order to obtain a dependency that can be used to determine the model parameters are changing during reactor operation.

  16. Power control system in BWR type reactors

    International Nuclear Information System (INIS)

    Nishizawa, Yasuo.

    1980-01-01

    Purpose: To control the reactor power so that the power distribution can satisfy the limiting conditions, by regulating the reactor core flow rate while monitoring the power distribution in the reactor core of a BWR type reactor. Constitution: A power distribution monitor determines the power distribution for the entire reactor core based on the data for neutron flux, reactor core thermal power, reactor core flow rate and control rod pattern from the reactor and calculates the linear power density distribution. A power up ratio computing device computes the current linear power density increase ratio. An aimed power up ratio is determined by converting the electrical power up ratio transferred from a load demand input device into the reactor core thermal power up ratio. The present reactor core thermal power up ratio is subtracted from the limiting power up ratio and the difference is sent to an operation amount indicator and the reactor core flow rate is changed in a reactor core flow rate regulator, by which the reactor power is controlled. (Moriyama, K.)

  17. RIMACS, Reactor Inspection Main Control System

    International Nuclear Information System (INIS)

    2008-01-01

    1 - Description of program or function: RIMACS prepares for automatic inspection files on each inspection item for the reactor. These automatic inspection files provide the data to move RIROB (Reactor Inspection Robot) with laser by interpreting the coordinates of LASPO (Laser Positioner) and the laser detecting device of RIROB in three dimensional space. In addition, when RIROB arrives at the inspecting location, the files provide all values of the manipulator's motions to acquire the ultrasonic data. RIMACS provides various modules in order to perform these complex functions, and the functions are programmed on graphic user interface for the convenience of the user. RIMACS provides various functions, such as insertion of reactor production data, selection of the reactor for inspection, the creation of automatic inspection file, the selection of the inspection item, inspection simulation, and automatic inspection procedures. It also provides all other functions, which are necessary for the inspection, such as operating program download and manual control of LASPO and RIROB, the inspection simulation and the inspection status display by means of the graphic screen, and SODAS (ultra-Sonic Data Acquisition System) drive verification. 2 - Methods: Moving path and operation procedures for inspection robot are generated automatically with Kinematics algorithm. 3 - Restrictions on the complexity of the problem: A graphics display with MS-Window capability is required

  18. Feedwater control system in BWR type reactor

    International Nuclear Information System (INIS)

    Tanji, Jun-ichi; Oomori, Takashi.

    1980-01-01

    Purpose: To improve the water level control performance in BWR type reactor by regulating the water level set to the reactor depending on the rate of change in the recycling amount of coolant to thereby control the fluctuations in the water level resulted in the reactor within an aimed range even upon significant fluctuations in the recycling flow rate. Constitution: The recycling flow rate of coolant in the reactor is detected and the rate of its change with time is computed to form a rate of change signal. The rate of change signal is inputted to a reactor level setter to amend the actual reactor water level demand signal and regulate the water level set to the reactor water depending on the rate of change in the recycling flow rate. Such a regulation method for the set water level enables to control the water level fluctuation resulted in the reactor within the aimed range even upon the significant fluctuation in the recycling flow rate and improve the water level control performance of the reactor, whereby the operationability for the reactor is improved to enhance the operation rate. (Moriyama, K.)

  19. Heating control system for nuclear reactor

    International Nuclear Information System (INIS)

    Shinohara, Kaoru.

    1981-01-01

    Purpose: To automatically control reactor heating while keeping the condition of temperature rising rate by determining the deviations based on the reactor water temperature, the aimed temperature and the aimed temperature rising rate and operating control rods. Constitution: Actual temperature in the reactor is measured by a temperature detector and compared with a value from a setter to determine the temperature deviation. While on the other hand, the rising rate for the measured temperature is calculated in a differentiator and compared with a value from a setter to determine the deviation, which is passed through an integrator to calculate the deviation for the temperature rising rate. The signals for the temperature deviation and the temperature rising rate deviation are selected in a lower value preference circuit and the operation amount for the control rod is judged in a control rod operation judging section depending on the deviation amount. The control rod to be operated is determined in a sequence control section for the selection of control rod. The control rod selected and the direction of the operation are displayed on a display and the selected control rod is automatically driven by a control rod drives to thereby carry our reactor heating. (Furukawa, Y.)

  20. Autonomous Control of Space Reactor Systems

    International Nuclear Information System (INIS)

    Belle R. Upadhyaya; K. Zhao; S.R.P. Perillo; Xiaojia Xu; M.G. Na

    2007-01-01

    Autonomous and semi-autonomous control is a key element of space reactor design in order to meet the mission requirements of safety, reliability, survivability, and life expectancy. Interrestrial nuclear power plants, human operators are available to perform intelligent control functions that are necessary for both normal and abnormal operational conditions

  1. Autonomous Control of Space Reactor Systems

    Energy Technology Data Exchange (ETDEWEB)

    Belle R. Upadhyaya; K. Zhao; S.R.P. Perillo; Xiaojia Xu; M.G. Na

    2007-11-30

    Autonomous and semi-autonomous control is a key element of space reactor design in order to meet the mission requirements of safety, reliability, survivability, and life expectancy. Interrestrial nuclear power plants, human operators are avilable to perform intelligent control functions that are necessary for both normal and abnormal operational conditions.

  2. Development of Reactor Protection System (RPS) in Reactor Digital Instrumentation and Control System (ReDICS)

    International Nuclear Information System (INIS)

    Mohd Khairulezwan Abdul Manan; Mohd Sabri Minhat; Ridzuan Abdul Mutalib

    2013-01-01

    RTP Research Reactor are in the process upgraded from analogue control console system to a digital control console system . Upgrade process requires a statistical study to improve safety during reactor operation. RPS was developed to meet the needs of operational safety and at the same time comply with the guidelines set by the IAEA. RPS is in analog and hardware with industry standard interfaced with digital DAC (Data Acquisition and Control) and OWS (Operator Work Station). (author)

  3. Method of controlling ECCS system in reactors

    International Nuclear Information System (INIS)

    Oohashi, Hideaki; Ikehara, Morihiko.

    1982-01-01

    Purpose: To eliminate the risk of misoperation and thereby improve the reliability of ECCS system upon accident. Method: ECCS system for nuclear reactor is automatically started by either of signals from a water level detector in a pressure vessel or from a pressure detector in a reactor container. Further, the ECCS system is started or stopped by the manual operation irrespective of the signals, and the signals from the pressure detector are isolated from the ECCS-starting signal by the contacts which actuate interlocked with the stopping operation of the manual operation switch. Then, after stopping the ECCS system by the manual operation, the ECCS system is started by the signals from the water level detector irrespective of the signals from the pressure detector. (Seki, T.)

  4. Micro processor based research reactor instrumentation and control system

    International Nuclear Information System (INIS)

    Hyde, W.K.

    1987-01-01

    The system consists of a Control System Computer (CSC) incorporated into a Reactor Control Console (RCC) and a Data Acquisition and Control Unit (DAC) adjacent to the reactor. The CSC has a high resolution color graphics CRT monitor which provides real-time graphic simulation of the reactor and a number of bar graphs displaying strategic parameters of the reactor system. In addition, abnormal or dangerous conditions are displayed. The CSC is equipped with two printers eliminating manual logging of reactor data. The reactor display and pulse mode display may also be printed. Historical data is saved in the system's large capacity memory and may be replayed and/or printed. Because of the CSC's inherent high speed math capability, raw reactor data will be quickly converted and displayed in real-time. Data can be presented in meaningful engineering units. The DAC provides a high speed data acquisition and control capability adjacent to the reactor. It continuously collects data from the reactor system, concentrates the data into a database and transmits it to the CSC when requested. Data transmission is over one of two data trunks to the CSC. The secondary trunk is used if the primary trunk fails. The data trunks drastically reduce the wiring requirements between the reactor and the Control Console. During steady-state operation of the reactor, operator commands to adjust the rod positions is transmitted from the CSC to the DAC which re-issues the commands to the drive mechanisms. In the automatic mode, the DAC will control the position of the rods via a PID algorithm. The system is independently monitored by two or more safety computers. Their function is to monitor the power level, the rate of change of power and fuel temperature of the reactor and to independently shut the reactor down in the event of a potentially dangerous (scram) condition. (author)

  5. Reactor core flow rate control system

    International Nuclear Information System (INIS)

    Sakuma, Hitoshi; Tanikawa, Naoshi; Takahashi, Toshiyuki; Miyakawa, Tetsuya.

    1996-01-01

    When an internal pump is started by a variable frequency power source device, if magnetic fields of an AC generator are introduced after the rated speed is reached, neutron flux high scram occurs by abrupt increase of a reactor core flow rate. Then, in the present invention, magnetic fields for the AC generator are introduced at a speed previously set at which the fluctuation range of the reactor core flow rate (neutron flux) by the start up of the internal pump is within an allowable value. Since increase of the speed of the internal pump upon its start up is suppressed to determine the change of the reactor core flow rate within an allowable range, increase of neutron fluxes is suppressed to enable stable start up. Then, since transition boiling of fuels caused by abrupt decrease of the reactor core flow rate upon occurrence of abnormality in an external electric power system is prevented, and the magnetic fields for the AC generator are introduced in such a manner to put the speed increase fluctuation range of the internal pump upon start up within an allowable value, neutron flux high scram is not caused to enable stable start-up. (N.H.)

  6. Nuclear reactor power control system based on flexibility model

    International Nuclear Information System (INIS)

    Li Gang; Zhao Fuyu; Li Chong; Tai Yun

    2011-01-01

    Design the nuclear reactor power control system in this paper to cater to a nonlinear nuclear reactor. First, calculate linear power models at five power levels of the reactor as five local models and design controllers of the local models as local controllers. Every local controller consists of an optimal controller contrived by the toolbox of Optimal Controller Designer (OCD) and a proportion-integration-differentiation (PID) controller devised via Genetic Algorithm (GA) to set parameters of the PID controller. According to the local models and controllers, apply the principle of flexibility model developed in the paper to obtain the flexibility model and the flexibility controller at every power level. Second, the flexibility model and the flexibility controller at a level structure the power control system of this level. The set of the whole power control systems corresponding to global power levels is to approximately carry out the power control of the reactor. Finally, the nuclear reactor power control system is simulated. The simulation result shows that the idea of flexibility model is feasible and the nuclear reactor power control system is effective. (author)

  7. Digital, remote control system for a 2-MW research reactor

    International Nuclear Information System (INIS)

    Battle, R.E.; Corbett, G.K.

    1988-01-01

    A fault-tolerant programmable logic controller (PLC) and operator workstations have been programmed to replace the hard-wired relay control system in the 2-MW Bulk Shielding Reactor (BSR) at Oak Ridge National Laboratory. In addition to the PLC and remote and local operator workstations, auxiliary systems for remote operation include a video system, an intercom system, and a fiber optic communication system. The remote control station, located at the High Flux Isotope Reactor 2.5 km from the BSR, has the capability of rector startup and power control. The system was designed with reliability and fail-safe features as important considerations. 4 refs., 3 figs

  8. Simulation of the TREAT-Upgrade Automatic Reactor Control System

    International Nuclear Information System (INIS)

    Lipinski, W.C.; Kirsch, L.W.; Valente, A.D.

    1984-01-01

    This paper describes the design of the Automatic Reactor Control System (ARCS) for the Transient Reactor Test Facility (TREAT) Upgrade. A simulation was used to facilitate the ARCS design and to completely test and verify its operation before installation at the TREAT facility

  9. Research Reactor Power Control System Design by MATLAB/SIMULINK

    International Nuclear Information System (INIS)

    Baang, Dane; Suh, Yong Suk; Kim, Young Ki; Im, Ki Hong

    2013-01-01

    In this study it is presented that MATLAB/SIMULINK can be efficiently used for modeling and power control system design for research reactors. The presented power control system deals with various functions including reactivity control, signals processing, reactivity calculation, alarm request generation, etc., thus it is required to test all the software logic using proper model for reactor, control rods, and field instruments. In MATLAB/SIMULINK tool, point kinetics, thermal model, control absorber rod model, and other instrument models were developed based on reactor parameters and known properties of each component or system. The software for power control system was invented and linked to the model to test each function. From the simulation result it is shown that the power control performance and other functions of the system can be easily tested and analyzed in the proposed simulation structure

  10. A computer control system for a research reactor

    International Nuclear Information System (INIS)

    Crawford, K.C.; Sandquist, G.M.

    1987-01-01

    Most reactor applications until now, have not required computer control of core output. Commercial reactors are generally operated at a constant power output to provide baseline power. However, if commercial reactor cores are to become load following over a wide range, then centralized digital computer control is required to make the entire facility respond as a single unit to continual changes in power demand. Navy and research reactors are much smaller and simpler and are operated at constant power levels as required, without concern for the number of operators required to operate the facility. For navy reactors, centralized digital computer control may provide space savings and reduced personnel requirements. Computer control offers research reactors versatility to efficiently change a system to develop new ideas. The operation of any reactor facility would be enhanced by a controller that does not panic and is continually monitoring all facility parameters. Eventually very sophisticated computer control systems may be developed which will sense operational problems, diagnose the problem, and depending on the severity of the problem, immediately activate safety systems or consult with operators before taking action

  11. MAPLE-X10 reactor digital control system

    International Nuclear Information System (INIS)

    Deverno, M.T.; Hinds, H.W.

    1991-10-01

    The MAPLE-X10 reactor, currently under construction at the Chalk River Laboratories of Atomic Energy of Canada Limited, is a 10 MW t , pool-type, light-water reactor. It will be used for radioisotope production and silicon neutron transmutation doping. The reactor is controlled by a Digital Control System (DCS) and protected against abnormal process events by two independent safety systems. The DCS is an integrated control system used to regulate the reactor power and process systems. The safety philosophy for the control system is to minimize unsafe events arising from system failures and operational errors. this is achieved through redundancy, fail-safe design, automatic fault detection, and the selection of highly reliable components. The DCS provides both computer-controlled reactor regulation from the shutdown state to full power and automated reactor shutdown if safe limits are exceeded or critical sensors malfunction. The use of commercially available hardware with enhanced quality assurance makes the system cost effective while providing a high degree of reliability

  12. Study, design and evaluation of nuclear reactor computer control system

    International Nuclear Information System (INIS)

    Menacer, S.

    1988-01-01

    Nuclear reactor control is a complex process that varies with each reactor and there is no universal agreement as to the best type of control system. After the use of conventional systems for a long time, attention turned towards digital techniques in the reactor control system. This interest emerged because of the difficulties faced in the data manipulation, mainly for post-incident analysis. However, it is not sufficient to insert a computer in a system to solve all the data-handling problems and also the insertion of a computer in a real-time system is not without any effect on the overall system. The scope of this thesis is to show the important parameters that have to be taken into account when choosing and evaluate the performances of the selected system

  13. Synthesis of relay control systems for nuclear reactors

    International Nuclear Information System (INIS)

    Postnikov, N.S.

    1996-01-01

    The problem on stabilizing an oscillatory-unstable reactor by a single-link relay system, the characteristics whereof have a dead zone and hysteresis loop, is considered. The methodology of synthesis of feedback law, providing for stochastic steady-state mode of reactor operation with the minimum frequency of control impact introduction is proposed. This methodology is applicable to general-type relay systems with arbitrary oscillatory-unstable objects. 6 refs., 5 figs

  14. Fault tolerant digital control systems for boiling water reactors

    International Nuclear Information System (INIS)

    Chakraborty, S.; Cash, N.R.

    1986-01-01

    In a Boiling Water Reactor nuclear power plant, the power generation control function is divided into several systems, each system controlling only a part of the total plant. Presently, each system is controlled by conventional analog or digital logic circuits with little interaction for coordinated control. The advent of microprocessors has allowed the development of distributed fault-tolerant digital controls. The objective is to replace these conventional controls with fault-tolerant digital controls connected together with digital communication links to form a fully integrated nuclear power plant control system

  15. Backfitting in Rossendorf research reactor control and instrumentation system

    International Nuclear Information System (INIS)

    Klebau, J.; Seidler, S.

    1985-01-01

    The paper generally describes a decentralized Hierarchical Information System (HIS) which has been developed for backfitting in Rossendorf Research Reactor (RFR) control and instrumentation system. The RFR was put into operation in 1957 and reconstructed from 2 MW up to a thermal power of 10 MW at the end of the sixties. Backfitting is planned by use of an advanced computerized control system for the next years. Main tasks of HIS are: Processmonitoring, online-disturbance analysis, technical diagnosis, direct digital control and use of a special industrial robot for discharging of irradiated materials out of the reactor. Experiences obtained by HIS during a testperiod will be presented. (author)

  16. Safety regulations concerning instrumentation and control systems for research reactors

    International Nuclear Information System (INIS)

    El-Shanshoury, A.I.

    2009-01-01

    A brief study on the safety and reliability issues related to instrumentation and control systems in nuclear reactor plants is performed. In response, technical and strategic issues are used to accomplish instrumentation and control systems safety. For technical issues there are ; systems aspects of digital I and C technology, software quality assurance, common-mode software, failure potential, safety and reliability assessment methods, and human factors and human machine interfaces. The strategic issues are the case-by-case licensing process and the adequacy of the technical infrastructure. The purpose of this work was to review the reliability of the safety systems related to these technical issues for research reactors

  17. Modernization of turbine control system and reactor control system in Almaraz 1 and 2

    International Nuclear Information System (INIS)

    Pulido, C.; Diez, J.; Carrasco, J. A.; Lopez, L.

    2005-01-01

    The replacement of the Turbine Control System and Reactor Control System are part of the Almaraz modernization program for the Instrumentation and Control. For these upgrades Almaraz has selected the Ovation Platform that provides open architecture and easy expansion to other systems, these platforms is highly used in many nuclear and thermal plants around the world. One of the main objective for this project were to minimize the impact on the installation and operation of the plant, for that reason the project is implemented in two phases, Turbine Control upgrade and Reactor Control upgrade. Another important objective was to increase the reliability of the control system making them fully fault tolerant to single failures. The turbine Control System has been installed in Units 1 and 2 while the Reactor Control System will be installed in 2006 and 2007 outages. (Author)

  18. Automatic control system in the reactor peggy

    International Nuclear Information System (INIS)

    Bertrand, J.; Mourchon, R.; Da Costa, D.; Desandre-Navarre, Ch.

    1967-01-01

    The equipment makes it possible for the reactor to attain a given power automatically and for the power to be maintained around this level. The principle of its operation consists in the changing from one power to another, at constant period, by means of a programmer transforming a power-step request into a voltage variation which is linear with time and which represents the logarithm of the required power. The real power is compared continuously with the required power. Stabilization occurs automatically as soon as the difference between the reactor power and the required power diminishes to a few per cent. (authors) [fr

  19. Architectural conceptual definition of the CAREM-25 reactor's control system

    International Nuclear Information System (INIS)

    Perez, J.C.; Santome, D.; Drexler, J.; Escudero, S.

    1990-01-01

    This work presents the conceptual definition of the CAREM 25 reactor's digital and monitoring control system structure. The requirements of the system are analyzed and different implementation alternatives are studied where possible basic architectures of the system and its topology are considered and evaluated. (Author) [es

  20. The digital reactor protection system for the instrumentation and control of reactor TRIGA PUSPATI (RTP)

    International Nuclear Information System (INIS)

    Nurfarhana Ayuni Joha; Izhar Abu Hussin; Mohd Idris Taib; Zareen Khan Abdul Jalil Khan

    2010-01-01

    Reactor Protection System (RPS) is important for Reactor Instrumentation and Control System. The RPS comprises all redundant electrical devices and circuitry involved in the generation of those initiating signals associated to the trip protective function. The instrumentation system for the RPS provides automatic protection signals against unsafe and improper reactor operation. The physical separation is provided for all of the redundant instrumentation systems to preserve redundancy. The safety protection systems using circuits composed of analog instruments and relays with relay contacts is difficult to realize from various reasons. Therefore, an application of digital technology can be said a logical conclusion also in the light of its functional superiority. (author)

  1. Application of fuzzy logic control system for reactor feed-water control

    International Nuclear Information System (INIS)

    Iijima, T.; Nakajima, Y.

    1994-01-01

    The successful actual application of a fuzzy logic control system to the a nuclear Fugen nuclear power reactor is described. Fugen is a heavy-water moderated, light-water cooled reactor. The introduction of fuzzy logic control system has enabled operators to control the steam drum water level more effectively in comparison to a conventional proportional-integral (PI) control system

  2. Reactor trip on turbine trip inhibit control system for nuclear power generating system

    International Nuclear Information System (INIS)

    Torres, J.M.; Musick, C.R.

    1976-01-01

    A reactor trip on turbine trip inhibit control system for a nuclear power generating system which utilizes steam bypass valves is described. The control system inhibits a normally automatic reactor trip on turbine trip when the bypass valves have the capability of bypassing enough steam to prevent reactor trip limits from being reached and/or to prevent opening of the secondary safety pressure valves. The control system generates a bypass valve capability signal which is continuously compared with the reactor power. If the capability is greater than the reactor power, then an inhibit signal is generated which prevents a turbine trip signal from tripping the nuclear reactor. 10 claims, 4 figures

  3. Integrated Management System, Configuration and Document Control for Research Reactors

    International Nuclear Information System (INIS)

    Steynberg, B.J.; Bruyn, J.F. du

    2017-01-01

    An integrated management system is a single management framework establishing all the processes necessary for the organisation to address all its goals and objectives. Very often only quality, environment and health & safety goals are included when referred to an integrated management system. However, within the research reactor environment such system should include goals pertinent to economic, environmental, health, operational, quality, safeguards, safety, security, and social considerations. One of the important objectives of an integrated management is to create the environment for a healthy safety culture. Configuration management is a disciplined process that involves both management and technical direction to establish and document the design requirements and the physical configuration of the research reactor and to ensure that they remain consistent with each other and the documentation. Configuration is the combination of the physical, functional, and operational characteristics of the structures, systems, and components (SSCs) or parts of the research reactor, operation, or activity. The basic objectives and general principles of configuration management are the same for all research reactors. The objectives of configuration management are to: a) Establish consistency among design requirements, physical configuration, and documentation (including analyses, drawings, and procedures) for the research reactor; b) Maintain this consistency throughout the life of the research reactor, particularly as changes are being made; and c) Retain confidence in the safety of the research reactor. The key elements needed to manage the configuration of research reactors are design requirements, work control, change control, document control, and configuration management assessments. The objective of document control is to ensure that only the most recently approved versions of documents are used in the process of operating, maintaining, and modifying the research reactor

  4. NSSS Component Control System Design of Integral Reactor

    International Nuclear Information System (INIS)

    Lee, Joon Koo; Kwon, Ho Je; Jeong, Kwong Il; Park, Heui Youn; Koo, In Soo

    2005-01-01

    MMIS(Man Machine Interface System) of an integral reactor is composed of a Control Room, Plant Protection System, Control System and Monitoring System which are related with the overall plant operation. MMIS is being developed with a new design concept and digital technology to reduce the Human Factor Error and improve the systems' safety, reliability and availability. And CCS(component control system) is also being developed with a new design concept and digital hardware technology A fully digitalized system and design concept are introduced in the NSSS CCS

  5. Reactivity control system of the high temperature engineering test reactor

    International Nuclear Information System (INIS)

    Tachibana, Yukio; Sawahata, Hiroaki; Iyoku, Tatsuo; Nakazawa, Toshio

    2004-01-01

    The reactivity control system of the high temperature engineering test reactor (HTTR) consists of a control rod system and a reserve shutdown system. During normal operation, reactivity is controlled by the control rod system, which consists of 32 control rods (16 pairs) and 16 control rod drive mechanisms except for the case when the center control rods are removed to perform an irradiation test. In an unlikely event that the control rods fail to be inserted, reserve shutdown system is provided to insert pellets of neutron-absorbing material into the core. Alloy 800H is chosen for the metallic parts of the control rods. Because the maximum temperature of the control rods reaches about 900 deg. C at reactor scrams, structural design guideline and design material data on Alloy 800H are needed for the high temperature design. The design guideline for the HTTR control rod is based on ASME Code Case N-47-21. Design material data is also determined and shown in this paper. Observing the guideline, temperature and stress analysis were conducted; it can be confirmed that the target life of the control rods of 5 years can be achieved. Various tests conducted for the control rod system and the reserve shutdown system are also described

  6. Comparative analysis of nuclear reactor control system designs

    International Nuclear Information System (INIS)

    Russcher, G.E.

    1975-01-01

    Control systems are vital to the safe operation of nuclear reactors. Their seismic design requirements are some of the most important criteria governing reactor system design evaluation. Consequently, the seismic analysis for nuclear reactors is directed to include not only the mechanical and structural seismic capabilities of a reactor, but the control system functional requirements as well. In the study described an alternate conceptual design of a safety rod system was compared with a prototypic system design to assess their relative functional reliabilities under design seismic conditions. The comparative methods utilized standard success tree and decision tree techniques to determine the relative figures of merit. The study showed: (1) The methodology utilized can provide both qualitative and quantitative bases for design decisions regarding seismic functional capabilities of two systems under comparison, (2) the process emphasizes the visibility of particular design features that are subject to common mode failure while under seismic loading, and (3) minimal improvement was shown to be available in overall system seismic performance of an independent conceptual design, however, it also showed the system would be subject to a new set of operational uncertainties which would have to be resolved by extensive development programs

  7. Using the digital reactor control systems at NPP

    International Nuclear Information System (INIS)

    Schirl, G.; Hertel, J.

    2006-01-01

    A conception of application of the digital reactor control systems (RCS) at NPP is presented. The digital RCS architecture and safety ensuring are considered. The strategy and algorithm of the operating NPP equipping with the new digital RCS are given too [ru

  8. Steam relief valve control system for a nuclear reactor

    International Nuclear Information System (INIS)

    Torres, J.M.

    1976-01-01

    Described is a turbine follow system and method for Pressurized Water Reactors utilizing load bypass and/or atmospheric dump valves to provide a substitute load upon load rejection by bypassing excess steam to a condenser and/or to the atmosphere. The system generates a variable pressure setpoint as a function of load and applies an error signal to modulate the load bypass valves. The same signal which operates the bypass valves actuates a control rod automatic withdrawal prevent to insure against reactor overpower

  9. 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

  10. Fully integrated analysis of reactor kinetics, thermalhydraulics and the reactor control system in the MAPLE-X10 research reactor

    International Nuclear Information System (INIS)

    Shim, S.Y.; Carlson, P.A.; Baxter, D.K.

    1992-01-01

    A prototype research reactor, designated MAPLE-X10 (Multipurpose Applied Physics Lattice Experimental - X 10MW), is currently being built at AECL's Chalk River Laboratories. The CATHENA (Canadian Algorithm for Thermalhydraulic Network Analysis) two-fluid code was used in the safety analysis of the reactor to determine the adequacy of core cooling during postulated reactivity and loss-of-forced-flow transients. The system responses to a postulated transient are predicted including the feedback between reactor kinetics, thermalhydrauilcs and the reactor control systems. This paper describes the MAPLE-X10 reactor and the modelling methodology used. Sample simulations of postulated loss-of-heat-sink and loss-of-regulation transients are presented. (author)

  11. Enabling autonomous control for space reactor power systems

    International Nuclear Information System (INIS)

    Wood, R. T.

    2006-01-01

    The application of nuclear reactors for space power and/or propulsion presents some unique challenges regarding the operations and control of the power system. Terrestrial nuclear reactors employ varying degrees of human control and decision-making for operations and benefit from periodic human interaction for maintenance. In contrast, the control system of a space reactor power system (SRPS) employed for deep space missions must be able to accommodate unattended operations due to communications delays and periods of planetary occlusion while adapting to evolving or degraded conditions with no opportunity for repair or refurbishment. Thus, a SRPS control system must provide for operational autonomy. Oak Ridge National Laboratory (ORNL) has conducted an investigation of the state of the technology for autonomous control to determine the experience base in the nuclear power application domain, both for space and terrestrial use. It was found that control systems with varying levels of autonomy have been employed in robotic, transportation, spacecraft, and manufacturing applications. However, autonomous control has not been implemented for an operating terrestrial nuclear power plant nor has there been any experience beyond automating simple control loops for space reactors. Current automated control technologies for nuclear power plants are reasonably mature, and basic control for a SRPS is clearly feasible under optimum circumstances. However, autonomous control is primarily intended to account for the non optimum circumstances when degradation, failure, and other off-normal events challenge the performance of the reactor and near-term human intervention is not possible. Thus, the development and demonstration of autonomous control capabilities for the specific domain of space nuclear power operations is needed. This paper will discuss the findings of the ORNL study and provide a description of the concept of autonomy, its key characteristics, and a prospective

  12. Human machine interface for research reactor instrumentation and control system

    International Nuclear Information System (INIS)

    Mohd Sabri Minhat; Mohd Idris Taib; Izhar Abu Hussin; Zareen Khan Abdul Jalil Khan; Nurfarhana Ayuni Joha

    2010-01-01

    Most present design of Human Machine Interface for Research Reactor Instrumentation and Control System is modular-based, comprise of several cabinets such as Reactor Protection System, Control Console, Information Console as well as Communication Console. The safety, engineering and human factor will be concerned for the design. Redundancy and separation of signal and power supply are the main factor for safety consideration. The design of Operator Interface absolutely takes consideration of human and environmental factors. Physical parameters, experiences, trainability and long-established habit patterns are very important for user interface, instead of the Aesthetic and Operator-Interface Geometry. Physical design for New Instrumentation and Control System of RTP are proposed base on the state-of- the-art Human Machine Interface design. (author)

  13. The reactor power control system based on digital control in nuclear power plant

    International Nuclear Information System (INIS)

    Liu Chong; Zhou Jianliang; Tan Ping

    2010-01-01

    The PLC (Programmable Logical Controller), digital communication and redundant techniques are applied in the rod control and position indication system(namely the reactor power control system) to perform the power control in the 300 MW reactor automatically and integrally in Qinshan Phase I project. This paper introduces the features, digital design methods of hardware of the instrumentation and control system (I and C) in the reactor power control. It is more convenient for the information exchange by human-machine interface (HMI), operation and maintenance, and the system reliability has been greatly improved after the project being reconstructed. (authors)

  14. Slow control systems of the Reactor Experiment for Neutrino Oscillation

    International Nuclear Information System (INIS)

    Choi, J.H.; Jang, H.I.; Choi, W.Q.; Choi, Y.; Jang, J.S.; Jeon, E.J.; Joo, K.K.; Kim, B.R.; Kim, H.S.; Kim, J.Y.; Kim, S.B.; Kim, S.Y.; Kim, W.; Kim, Y.D.; Ko, Y.J.; Lee, J.K.; Lim, I.T.; Pac, M.Y.; Park, I.G.; Park, J.S.

    2016-01-01

    The RENO experiment has been in operation since August 2011 to measure reactor antineutrino disappearance using identical near and far detectors. For accurate measurements of neutrino mixing parameters and efficient data taking, it is crucial to monitor and control the detector in real time. Environmental conditions also need to be monitored for stable operation of detectors as well as for safety reasons. In this paper, we report the design, hardware, operation, and performance of the slow control system.

  15. Control device for start-up of reactor depressurization system

    International Nuclear Information System (INIS)

    Suzuki, Hiroshi; Saito, Minoru; Oda, Shingo; Miura, Satoshi; Hashimoto, Koji; Tate, Hitoshi; Fujii, Kazunobu

    1998-01-01

    The present invention concerns are emergency reactor core cooling system (ECCS) of a BWR type reactor and provides a control device for start-up of an automatic depressurization system. Namely, the device has an object of preventing erroneous opening of a main steam escape safety value when testing a start-up signal circuit of an automatic depressurization system for testing the automatic depressurization system. A start-up signal circuit receives both signals of a reactor container pressure high signal and a reactor pressure vessel water level low signal and outputs an automatic start-up signal for compulsorily opening a main steam escape safety valve automatically. A test switch having a self-holding circuit is disposed to a central control chamber. A test signal circuit is disposed for preventing transfer of an erroneous start-up signal to the main steam escape safety valve due to a simulation signal during output test signals by the test switch. (I.S.)

  16. 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

  17. Inteligent control system for a CANDU 600 type reactor process

    International Nuclear Information System (INIS)

    Venescu, B.; Zevedei, D.; Jurian, M.; Venescu, R.

    2013-01-01

    The present paper is set on presenting a highly intelligent configuration, capable of controlling, without the need of the human factor, a complete nuclear power plant type of system, giving it the status of an autonomous system. The urge for such a controlling system is justified by the amount of drawbacks that appear in real life as disadvantages, loses and sometimes even inefficiency in the current controlling and comanding systems of the nuclear reactors. The application stands in the comand sent from the auxiliary feedwater flow control valves to the steam generators. As an environment fit for development I chose Matlab Simulink to simulate the behaviour of the process and the adjusted system. Comparing the results obtained after the fuzzy regulation with those obtained after the classical regulation, we can demonstrate the necessity of implementing artificial intelligence techniques in nuclear power plants and we can agree to the advantages of being able to control everything automatically. (authors)

  18. Autonomous Control Capabilities for Space Reactor Power Systems

    International Nuclear Information System (INIS)

    Wood, Richard T.; Neal, John S.; Brittain, C. Ray; Mullens, James A.

    2004-01-01

    The National Aeronautics and Space Administration's (NASA's) Project Prometheus, the Nuclear Systems Program, is investigating a possible Jupiter Icy Moons Orbiter (JIMO) mission, which would conduct in-depth studies of three of the moons of Jupiter by using a space reactor power system (SRPS) to provide energy for propulsion and spacecraft power for more than a decade. Terrestrial nuclear power plants rely upon varying degrees of direct human control and interaction for operations and maintenance over a forty to sixty year lifetime. In contrast, an SRPS is intended to provide continuous, remote, unattended operation for up to fifteen years with no maintenance. Uncertainties, rare events, degradation, and communications delays with Earth are challenges that SRPS control must accommodate. Autonomous control is needed to address these challenges and optimize the reactor control design. In this paper, we describe an autonomous control concept for generic SRPS designs. The formulation of an autonomous control concept, which includes identification of high-level functional requirements and generation of a research and development plan for enabling technologies, is among the technical activities that are being conducted under the U.S. Department of Energy's Space Reactor Technology Program in support of the NASA's Project Prometheus. The findings from this program are intended to contribute to the successful realization of the JIMO mission

  19. Development of Power Controller System based on Model Reference Adaptive Control for a Nuclear Reactor

    International Nuclear Information System (INIS)

    Mohd Sabri Minhat; Izhar Abu Hussin; Ridzuan Abdul Mutalib

    2014-01-01

    The Reactor TRIGA PUSPATI (RTP)-type TRIGA Mark II was installed in the year 1982. The Power Controller System (PCS) or Automated Power Controller System (APCS) is very important for reactor operation and safety reasons. It is a function of controlled reactivity and reactor power. The existing power controller system is under development and due to slow response, low accuracy and low stability on reactor power control affecting the reactor safety. The nuclear reactor is a nonlinear system in nature, and it is power increases continuously with time. The reactor parameters vary as a function of power, fuel burnup and control rod worth. The output power value given by the power control system is not exactly as real value of reactor power. Therefore, controller system design is very important, an adaptive controller seems to be inevitable. The method chooses is a linear controller by using feedback linearization, for example Model Reference Adaptive Control. The developed APCS for RTP will be design by using Model Reference Adaptive Control (MRAC). The structured of RTP model to produce the dynamic behaviour of RTP on entire operating power range from 0 to 1MWatt. The dynamic behavior of RTP model is produced by coupling of neutronic and thermal-hydraulics. It will be developed by using software MATLAB/Simulink and hardware module card to handle analog input signal. A new algorithm for APCS is developed to control the movement of control rods with uniformity and orderly for RTP. Before APCS test to real plant, simulation results shall be obtained from RTP model on reactor power, reactivity, period, control rod positions, fuel and coolant temperatures. Those data are comparable with the real data for validation. After completing the RTP model, APCS will be tested to real plant on power control system performance by using real signal from RTP including fail-safe operation, system reliable, fast response, stability and accuracy. The new algorithm shall be a satisfied

  20. Computer simulation system of neural PID control on nuclear reactor

    International Nuclear Information System (INIS)

    Chen Yuzhong; Yang Kaijun; Shen Yongping

    2001-01-01

    Neural network proportional integral differential (PID) controller on nuclear reactor is designed, and the control process is simulated by computer. The simulation result show that neutral network PID controller can automatically adjust its parameter to ideal state, and good control result can be gotten in reactor control process

  1. Expert system driven fuzzy control application to power reactors

    International Nuclear Information System (INIS)

    Tsoukalas, L.H.; Berkan, R.C.; Upadhyaya, B.R.; Uhrig, R.E.

    1990-01-01

    For the purpose of nonlinear control and uncertainty/imprecision handling, fuzzy controllers have recently reached acclaim and increasing commercial application. The fuzzy control algorithms often require a ''supervisory'' routine that provides necessary heuristics for interface, adaptation, mode selection and other implementation issues. Performance characteristics of an on-line fuzzy controller depend strictly on the ability of such supervisory routines to manipulate the fuzzy control algorithm and enhance its control capabilities. This paper describes an expert system driven fuzzy control design application to nuclear reactor control, for the automated start-up control of the Experimental Breeder Reactor-II. The methodology is verified through computer simulations using a valid nonlinear model. The necessary heuristic decisions are identified that are vitally important for the implemention of fuzzy control in the actual plant. An expert system structure incorporating the necessary supervisory routines is discussed. The discussion also includes the possibility of synthesizing the fuzzy, exact and combined reasoning to include both inexact concepts, uncertainty and fuzziness, within the same environment

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

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  3. Robust reactor power control system design by genetic algorithm

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yoon Joon; Cho, Kyung Ho; Kim, Sin [Cheju National University, Cheju (Korea, Republic of)

    1997-12-31

    The H{sub {infinity}} robust controller for the reactor power control system is designed by use of the mixed weight sensitivity. The system is configured into the typical two-port model with which the weight functions are augmented. Since the solution depends on the weighting functions and the problem is of nonconvex, the genetic algorithm is used to determine the weighting functions. The cost function applied in the genetic algorithm permits the direct control of the power tracking performances. In addition, the actual operating constraints such as rod velocity and acceleration can be treated as design parameters. Compared with the conventional approach, the controller designed by the genetic algorithm results in the better performances with the realistic constraints. Also, it is found that the genetic algorithm could be used as an effective tool in the robust design. 4 refs., 6 figs. (Author)

  4. Robust reactor power control system design by genetic algorithm

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yoon Joon; Cho, Kyung Ho; Kim, Sin [Cheju National University, Cheju (Korea, Republic of)

    1998-12-31

    The H{sub {infinity}} robust controller for the reactor power control system is designed by use of the mixed weight sensitivity. The system is configured into the typical two-port model with which the weight functions are augmented. Since the solution depends on the weighting functions and the problem is of nonconvex, the genetic algorithm is used to determine the weighting functions. The cost function applied in the genetic algorithm permits the direct control of the power tracking performances. In addition, the actual operating constraints such as rod velocity and acceleration can be treated as design parameters. Compared with the conventional approach, the controller designed by the genetic algorithm results in the better performances with the realistic constraints. Also, it is found that the genetic algorithm could be used as an effective tool in the robust design. 4 refs., 6 figs. (Author)

  5. Maintenance quality surveillance for nuclear reactor instrument and control systems

    International Nuclear Information System (INIS)

    Clement, T.M.

    1976-01-01

    A description is given of a formal program of mandatory testing and inspection for assurance of reliability at the N-reactor. The techniques and procedures are called Equipment Maintenance Standards (EMS) and cover the nuclear plant systems which affect nuclear safety, environmental control, continuity of operation and reactor life. The scope of the program is such that all electrical, chemical, mechanical, and nuclear devices with interconnecting circuitry, linkages, and piping are functionally checked and/or inspected from sensors through actuating media necessary for carrying out the defined functions under startup, operating, and shutdown conditions and also under any credible accident condition. Records are kept of the AS FOUND results of the tests and inspections along with adjustments and repairs required to correct any out-of-limits conditions found. These records are accumulated and are used as a source for quantitative reliability information and for evaluating wear-out or design problems

  6. Supervisory Control System Architecture for Advanced Small Modular Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Cetiner, Sacit M [ORNL; Cole, Daniel L [University of Pittsburgh; Fugate, David L [ORNL; Kisner, Roger A [ORNL; Melin, Alexander M [ORNL; Muhlheim, Michael David [ORNL; Rao, Nageswara S [ORNL; Wood, Richard Thomas [ORNL

    2013-08-01

    This technical report was generated as a product of the Supervisory Control for Multi-Modular SMR Plants project within the Instrumentation, Control and Human-Machine Interface technology area under the Advanced Small Modular Reactor (SMR) Research and Development Program of the U.S. Department of Energy. The report documents the definition of strategies, functional elements, and the structural architecture of a supervisory control system for multi-modular advanced SMR (AdvSMR) plants. This research activity advances the state-of-the art by incorporating decision making into the supervisory control system architectural layers through the introduction of a tiered-plant system approach. The report provides a brief history of hierarchical functional architectures and the current state-of-the-art, describes a reference AdvSMR to show the dependencies between systems, presents a hierarchical structure for supervisory control, indicates the importance of understanding trip setpoints, applies a new theoretic approach for comparing architectures, identifies cyber security controls that should be addressed early in system design, and describes ongoing work to develop system requirements and hardware/software configurations.

  7. CAMAC-controlled calibration system for nuclear reactor instruments

    International Nuclear Information System (INIS)

    McDowell, W.P.; Cornella, R.J.

    1977-01-01

    The hardware and the software which have been developed to implement a nuclear instrument calibration system for the Argonne National Laboratory ZPR-VI and ZPR-IX reactor complex are described. The system is implemented using an SEL-840 computer with its associated CAMAC crates and a hardware interface to generate input parameters and measure the required outputs on the instrument under test. Both linear and logarithmic instruments can be calibrated by the system and output parameters can be measured at various automatically selected values of ac line voltage. A complete report on each instrument is printed as a result of the calibration and out-of-tolerance readings are flagged. Operator interface is provided by a CAMAC-controlled Hazeltine terminal. The terminal display leads the operator through the complete calibration procedure. This computer-controlled system is a significant improvement over previously used methods of calibrating nuclear instruments since it reduces reactor downtime and allows rapid detection of long-term changes in instrument calibration

  8. Nuclear reactor and production systems with digital controls

    International Nuclear Information System (INIS)

    Luger, P.P.

    1976-01-01

    Several digital sensing devices are described for use in automated production systems. The first described is for use in the automatic operation of a reactor. This device employs a binant electrometer using a quartz fiber mounted at one end but free to vibrate at the other in an AC field. The fiber oscillates if a charge is placed upon it. An optical slit replaces the ordinary eyepiece reticule scale. With the quartz fiber adjusted so its image is in focus at the optical slit, photoelectric signals are obtained at null charge on the fiber. The quartz fiber is repeatedly charged and allowed to discharge by collecting ions from a source under measurement. Each photoelectric signal causes a digital time reading to be taken. The time readings are used to evaluate the current due to the collected charge. The photoelectric signals, by feedback, also operate the electrometer for continuous or intermittent-continuous operation. Basically, the system is a current digitizer. Application is made to reactor monitoring and control as well as to other types of production systems. Finally, other types of sensing devices are also described and their use in automated controlled processes is shown. 3 claims, 19 figures

  9. Reactor power control method upon accidents of electrical power system

    International Nuclear Information System (INIS)

    Hirose, Masao.

    1983-01-01

    Purpose: To enable to continue the operation of a BWR type reactor by avoiding the scram while suppressing the reactor power, just after the external disturbance such as earth-trouble in power-transmission network. Method: Steep power drop of an electrical generator is to be detected not only by a current-type power-load-unbalance relay but also with a power-type power-load-unbalance-relay. If steep power-drop was detected by the latter relay, a previously selected control rod is rapidly inserted into the reactor. In this way, in the case where there is a possibility of the reactor scram, the scram can be avoided by suppressing the reactor power, thus the reactor operation can be continued. (Kamimura, M.)

  10. System and method for air temperature control in an oxygen transport membrane based reactor

    Science.gov (United States)

    Kelly, Sean M

    2016-09-27

    A system and method for air temperature control in an oxygen transport membrane based reactor is provided. The system and method involves introducing a specific quantity of cooling air or trim air in between stages in a multistage oxygen transport membrane based reactor or furnace to maintain generally consistent surface temperatures of the oxygen transport membrane elements and associated reactors. The associated reactors may include reforming reactors, boilers or process gas heaters.

  11. Tritium Mitigation/Control for Advanced Reactor System

    Energy Technology Data Exchange (ETDEWEB)

    Sun, Xiaodong; Christensen, Richard; Saving, John P

    2018-03-31

    A tritium removal facility, which is similar to the design used for tritium recovery in fusion reactors, is proposed in this study for fluoride-salt-cooled high-temperature reactors (FHRs) to result in a two-loop FHR design with the elimination of an intermediate loop. Using this approach, an economic benefit can potentially be obtained by removing the intermediate loop, while the safety concern of tritium release can be mitigated. In addition, an intermediate heat exchanger (IHX) that can yield a similar tritium permeation rate to the production rate of 1.9 Ci/day in a 1,000 MWe PWR needs to be designed to prevent the residual tritium that is not captured in the tritium removal system from escaping into the power cycle and ultimately the environment. The main focus of this study is to aid the mitigation of tritium permeation issue from the FHR primary side to significantly reduce the concentration of tritium in the secondary side and the process heat application side (if applicable). The goal of the research is to propose a baseline FHR system without the intermediate loop. The specific objectives to accomplish the goals are: 1. To estimate tritium permeation behavior in FHRs; 2. To design a tritium removal system for FHRs; 3. To meet the same tritium permeation level in FHRs as the tritium production rate of 1.9 Ci/day in 1,000 MWe PWRs; 4. To demonstrate economic benefits of the proposed FHR system via comparing with the three-loop FHR system. The objectives were accomplished by designing tritium removal facilities, developing a tritium analysis code, and conducting an economic analysis. In the fusion reactor community, tritium extraction has been widely investigated and researched. Borrowing the experiences from the fusion reactor community, a tritium control and mitigation system was proposed. Based on mass transport theories, a tritium analysis code was developed, and the tritium behaviors were analyzed using the developed code. Tritium removal facilities

  12. Reactor control device

    International Nuclear Information System (INIS)

    Fukami, Haruo; Morimoto, Yoshinori.

    1981-01-01

    Purpose: To operate a reactor always with safety operation while eliminating the danger of tripping. Constitution: In a reactor control device adapted to detect the process variants of a reactor, control a control rod drive controlling system based on the detected signal to thereby control the driving the control rods, control the reactor power and control the electric power generated from an electric generator by the output from the reactor, detection means is provided for the detection of the electric power from said electric generator, and a compensation device is provided for outputting control rod driving compensation signals to the control rod driving controlling system in accordance with the amount of variation in the detected value. (Seki, T.)

  13. Mathematical modelling and quality indices optimization of automatic control systems of reactor facility

    International Nuclear Information System (INIS)

    Severin, V.P.

    2007-01-01

    The mathematical modeling of automatic control systems of reactor facility WWER-1000 with various regulator types is considered. The linear and nonlinear models of neutron power control systems of nuclear reactor WWER-1000 with various group numbers of delayed neutrons are designed. The results of optimization of direct quality indexes of neutron power control systems of nuclear reactor WWER-1000 are designed. The identification and optimization of level control systems with various regulator types of steam generator are executed

  14. Analysis of man-machine interaction for control and display system in main control room of light water reactor

    International Nuclear Information System (INIS)

    Santosa, Kussigit; Supriatna, Piping; Karlina, Itjeu; Widagdo, Suharyo; Darlis; Sudiono, Bambang

    1998-01-01

    One of potential hazard in Nuclear Power Plant is the failure of its operation. The accident or operation failure in the reactor must be concerned event its probability is low. The important thing should be concerned is 'Analysis of Man-Machine Interaction (MMI) for Control and Display System in Main Control Room (MCR) of Nuclear Power Reactor', especially LWR type. Control and Display System in MCR of Reactor is the main part of MMI link process in Reactor MCR work system. Signal from display system showed performance process in reactor, while this signal will be received by operator. This signal will be described through central nerve for making decision what kind must be done. Then the operator manage the next process of reactor operation through control system. So by knowing Analysis of Man-Machine Interaction for Control and Display System in Main Control Room of Power Reactor, we can understand human error probability of the operator in reactor operation

  15. A basic design of SR4 instrumentation and control system for research reactor

    International Nuclear Information System (INIS)

    Syahrudin Yusuf; M Subhan; Ikhsan Shobari; Sutomo Budihardjo

    2010-01-01

    An SR4 instrumentation and control systems of research reactor is the equipment of nuclear research reactors as power protection devices and control systems. The equipment is to monitor safety parameters and process parameters in the state of reactor shut down, start-up, and in operation at fixed power. In the engineering of Instrumentation and control systems SR4 research reactor, its basic design consists of technical specifications of the reactor protection system devices, technical specifications of the reactor power control system devices, technical specifications information system devices, and systems process termination cabling as a support system. This basic design is used as the basis for the preparation of detailed design and subsequent engineering development of instrumentation systems and control system integrated. (author)

  16. Water quality control device and water quality control method for reactor primary coolant system

    International Nuclear Information System (INIS)

    Wada, Yoichi; Ibe, Eishi; Watanabe, Atsushi.

    1995-01-01

    The present invention is suitable for preventing defects due to corrosion of structural materials in a primary coolant system of a BWR type reactor. Namely, a concentration measuring means measures the concentration of oxidative ingredients contained in a reactor water. A reducing electrode is disposed along a reactor water flow channel in the primary coolant system and reduces the oxidative ingredients. A reducing counter electrode is disposed along the reactor water flow channel in the primary coolant system, and electrically connected to the reducing electrode. The reactor structural materials are used as a reference electrode providing a reference potential to the reducing electrode and the reducing counter electrode. A potential control means controls the potential of the reducing electrode relative to the reference potential based on the signals from the concentration measuring means. A stable reference potential in a region where an effective oxygen concentration is stable can be obtained irrespective of the change of operation conditions by using the reactor structural materials disposed to a boiling region in the reactor core as a reference electrode. As a result, the water quality can be controlled at high accuracy. (I.S.)

  17. Design of Multi Objectives Control Systems to Control Nuclear Reactor Power

    International Nuclear Information System (INIS)

    Abdelaal, M.M.Z.

    2013-01-01

    The Egyptian Testing Research Reactor (ETRR-2) nonlinear twelfth order model is linearized and reduced to lower order model. Model order reduction methodologies such as balanced truncation, Schur reduction method, Hankel approximation and Coprime factorization have been used in the reduction process. The reactor actually controlled by PD controller with fixed tuning parameters. LMI state feedback, LMI-pool assignment, H ∞ and observer based controllers based third order model are proposed to be used in the reactor power control instead of the PD controller. A comparison of LMI, LMI-Pole placement,H ∞ control systems and those of based observer relative to the PD controller has been performed which showed better response and disturbance rejection for the proposed controllers.

  18. Reactor power control device

    International Nuclear Information System (INIS)

    Kobayashi, Akira.

    1980-01-01

    Purpose: To prevent misoperation in a control system for the adjustment of core coolant flow rate, and the increase in the neutron flux density caused from the misoperation in BWR type reactors. Constitution: In a reactor power control system adapted to control the reactor power by the adjustment of core flow rate, average neutron flux signals of a reactor core, entire core flow rate signals and operation state signals for coolant recycling system are inputted to a microcomputer. The outputs from the computer are sent to a recycling MG set speed controller to control the reactor core flow rate. The computer calculates the change ratio with time in the average neutron flux signals, correlation between the average neutron flux signals and the entire core flow rate signals, change ratio with time in the operation state signals for the coolant recycling system and the like and judges the abnormality in the coolant recycling system based on the calculated results. (Ikeda, J.)

  19. Fluid moderator control system reactor internals distribution system

    International Nuclear Information System (INIS)

    Fensterer, H.F.; Klassen, W.E.; Veronesi, L.; Boyle, D.E.; Salton, R.B.

    1987-01-01

    This patent describes a spectral shift pressurized water nuclear reactor employing a low neutron moderating fluid for the spectral shift including a reactor pressure vessel, a core comprising a plurality of fuel assemblies, a core support plate, apparatus comprising means for penetrating the reactor vessel for introducing the moderating fluid into the reactor vessel. Means associated with the core support plate for directly distributing the moderating fluid to and from the fuel assemblies comprises at least one inlet flow channel in the core plate; branch inlet feed lines connect to the inlet flow channel in the core plate; vertical inlet flow lines flow connected to the branch inlet feed lines; each vertical flow line communicates with a fuel assembly; the distribution means further comprise lines serving as return flow lines, each of which is connected to one of the fuel assemblies; branch exit flow lines in the core plate flow connected to the return flow lines of the fuel assembly; and at least one outlet flow channel flow connected to the branch exit flow lines; and a flow port interposed between the penetration means and the distribution means for flow connecting the penetration means with the distribution means

  20. Development of automated controller system for controlling reactivity by using FPGA in research reactor application

    International Nuclear Information System (INIS)

    Mohd Sabri Minhat; Izhar Abu Hussin; Mohd Idris Taib

    2012-01-01

    The scope for this research paper is to produce a detail design for Development of Automated Controller System for Controlling Reactivity by using FPGA in Research Reactor Application for high safety nuclear operation. The development of this project including design, purchasing, fabrication, installation, testing and validation and verification for one prototype automated controller system for controlling reactivity in industry local technology for human capacity and capability development towards the first Nuclear Power Programme (NPP) in Malaysia. The specific objectives of this research paper are to Development of Automated Controller System for Controlling Reactivity (ACSCR) in Research Reactor Application (PUSPATI TRIGA Reactor) by using simultaneous movement method; To design, fabricate and produce the accuracy of Control Rods Drive Mechanism to 0.1 mm resolution using a stepper motor as an actuator; To design, install and produce the system response to be more faster by using Field Programmable Gate Array (FPGA) and High Speed Computer; and to improve the Safety Level of the Research Reactor in high safety nuclear operation condition. (author)

  1. Reactor power control device

    International Nuclear Information System (INIS)

    Imaruoka, Hiromitsu.

    1994-01-01

    A high pressure water injection recycling system comprising injection pipelines of a high pressure water injection system and a flow rate control means in communication with a pool of a pressure control chamber is disposed to a feedwater system of a BWR type reactor. In addition, the flow rate control means is controlled by a power control device comprising a scram impossible transient event judging section, a required injection flow rate calculation section for high pressure water injection system and a control signal calculation section. Feed water flow rate to be supplied to the reactor is controlled upon occurrence of a scram impossible transient event of the reactor. The scram impossible transient event is judged based on reactor output signals and scram operation demand signals and injection flow rate is calculated based on a predetermined reactor water level, and condensate storage tank water or pressure control chamber pool water is injected to the reactor. With such procedures, water level can be ensured and power can be suppressed. Further, condensate storage tank water of low enthalpy is introduced to the pressure suppression chamber pool to directly control elevation of water temperature and ensure integrity of the pressure vessel and the reactor container. (N.H.)

  2. Reactor power control systems in nuclear power plants

    International Nuclear Information System (INIS)

    Nakajima, Kazuo.

    1980-01-01

    Purpose: To enable power control by automatic control rod operation based on the calculated amounts of operation for the control rods determined depending on a power set value from reactor operators or on power variation amounts from other devices. Constitution: When an operator designates an automatic selection by way of a control rod operation panel, automatic signals are applied to a manual-automatic switching circuit and the mode judging circuit of a rod pattern control device. Then, mode signals such as for single operation, load setting, load following and the like produced by the operator are judged in a circuit, wherein a control rod pattern operation circuit calculates the designation for the control rods and the operation amounts for the control rods depending on the designated modes and automatic control is conducted for the control rods by a rod position control circuit, a rod drive control device and the like connected at a rod position monitor device. The reactor power is thus controlled automatically to reduce the operator's labours. The automatic power control can also be conducted in the same manner by the amount of power variations applied to the device from the external device. (Yoshino, Y.)

  3. 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)

  4. Implementation of digital control and protection systems of China advanced research reactor

    International Nuclear Information System (INIS)

    Zeng Hai; Jin Huajin; Xu Qiguo; Zhang Mingkui

    2005-01-01

    China Advanced Research Reactor (CARR), a reactor of the 21st century with high performance is being constructed in China. The requirements of reliability and stability on the control and protection (c and p) system are the main points raised. Especially, with the development of digital technology, the c and p system of CARR is demanded to match the trend of digitization in the field of reactor control. The c and p system, including reactor protection system, reactor monitoring and control system, reactor power regulating system, and the mitigation system for ATWS (Anticipate Transient Without Scram), adopts digital technology, and the digital display screen will replace the analog panels in the main control room. The c and p system of CARR adopts redundant technology with 2 or 3 redundant channels to improve the system reliability. The 10/100 Mbps self-adaptive redundant optic fiber industry Ethernet ring network is used to interlink operator workstations, supervisor workstation, and I/O control stations. Commercial grade equipment with mature experience in industrial application are applied to the c and p system of CARR, which have high reliability, good interchangeability, and is easily purchased, the software-developing tools fully match the international industry standards. The realization of digital c and p system of CARR will promote the progress of digital control technology for reactors in China, and certainly become a technical basic platform for developing informational and intelligent reactors in China. (authors)

  5. Description of a research reactor control system using a programmable controller

    International Nuclear Information System (INIS)

    Battle, R.E.

    1986-01-01

    This paper describes the design features, testing methods, and operational experience of a programmable controller (PC) installed as a neutron flux controller in the Oak Ridge Research Reactor (ORR) at Oak Ridge National Laboratory (ORNL). The PC was designed to control neutron flux from 1 to 105% for three selectable ranges. The PC generates a flux setpoint under operator control, calculates the reactor heat power from flow and temperature signals, calculates a neutron flux calibration factor based on the heat power, and positions a control rod based on the flux-setpoint difference. The programmable controller was tested by controlling an analog computer model of the ORR. The equipment was installed in August 1985, and except for some startup problems, the system has performed well

  6. An automatic regulating control system for a graphite moderated reactor using digital techniques

    International Nuclear Information System (INIS)

    Carvalho Goncalves Junior, J. de.

    1989-01-01

    The work propose an automatic regulating control system for a graphite moderated reactor using digital techniques. The system uses a microcomputer to monitor the power and the period, to run the control algorithm, and to generate electronic signals to excite the motor, which moves vertically the control rod banks. A nuclear reactor simulator was developed to test the control system. The simulator consists of a software based on the point kinetic equations and implanted in an analogical computer. The results show that this control system has a good performance and versatility. In addition, the simulator is capable of reproducing with accuracy the behavior of a nuclear reactor. (author)

  7. 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

  8. French experience in using programmable systems for the control and the protection of nuclear reactors

    International Nuclear Information System (INIS)

    Jover, P.

    1986-01-01

    The paper presents the results obtained in the use of two automated systems important to safety in 1300 MWE French nuclear reactors: the numerical integrated protection system (SPIN) and the logical control system (CONTROBLOC)

  9. Elements on reactor control

    International Nuclear Information System (INIS)

    Bruna, G.B.

    1998-01-01

    In order to achieve the two-fold goal of maximizing the energy obtained from reactor fuel and ensuring the large flexibility of plant operation in respect to safety regulations and keeping the reactor integrity the control of PWRs is generally based on real time monitoring and analysing of independent neutronic parameters: thermal power release, axial power distribution in the core and temperatures of the primary loop. Two control chains more or less coupled according to the control chosen mode are in charge of the control of these parameters. With the brief history of control in French power reactors the advanced X control mode adopted by Framatome for N4 plants is described in detail. A summary of N4 reactor control and protection system is included

  10. Control and instrumentation system of the Zero Power Reactor at IEA, Sao Paulo (Brazil)

    International Nuclear Information System (INIS)

    Peluso, M.A.V.; Matsuda, K.; Hukai, R.

    1974-01-01

    The control and instrumentation system of the Zero Power Reactor at the IEA (Institute of Atomic Energy - Sao Paulo, Brazil) is described. Technical specifications of the main items of equipment are presented in a general way. Information is also given on the connection between the system described and the electrical supply system of the IEA reactor physics laboratory [pt

  11. 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)

  12. KS-150 reactor control

    International Nuclear Information System (INIS)

    Wagner, K.

    1974-01-01

    A thorough description is presented of the control and protection system of the Bohunice A-1 reactor. The system including auxiliary facilities was developed, manufactured and installed at the reactor by the SKODA Works, Plzen. The system parameters are listed and a brief account is also given of the development efforts and of the physical and power start-up of the A-1 nuclear power plant. (L.O.)

  13. Application study of EPICS-based redundant method for reactor control system

    International Nuclear Information System (INIS)

    Zhang Ning; Han Lifeng; Chen Yongzhong; Guo Bing; Yin Congcong

    2013-01-01

    In the reactor control system prototype development of TMSR (Thorium Molten Salt Reactor system, CAS) project, EPICS (Experimental Physics and Industrial Control System) is adopted as Instrument and Control software platform. For the aim of IOC (Input/Output Controller) redundancy and data synchronization of the system, the EPICS-based RMT (Redundancy Monitor Task ) software package and its data-synchronization component CCE (Continuous Control Executive) were introduced. By the development of related IOC driver, redundant switch-over control of server IOC was implemented. The method of redundancy implementation using RMT in server and redundancy performance test for power control system are discussed in this paper. (authors)

  14. Upgrading the Reactor Power Control Concept with a Modern Digital Control System

    International Nuclear Information System (INIS)

    Laengle, M.; Schildheuer, R.

    2011-01-01

    Within the framework of a retrofit project, a reactor power instrumentation and control system (REALL) - consisting of a limiting system and the respective reactor control systems - was retrofitted and modernized in a 1450-MW-nuclear-power-plant in Baden-Wuerttemberg. The REALL process control functions were implemented within a modern and completely digitized control system that has been designed for use in safety I and C applications. Along with the installation of the digital control system, the associated hardware was adapted to today's state of the art. At the same time, the given potential for improvement, as revealed during the plant's operation so far, was taken into account in the programming. In order to provide for transparent and quality-assured project management, the implementation was based on a stage plan consisting of several steps, along with specific milestones. Final commissioning of the modern digital control system took place during the 2008 plant overhaul. Despite the complex commissioning procedure, it was possible to avoid a major prolongation of the plant's downtime and to keep within a rough 4-week timeframe that had originally been defined for the plant overhaul to adequate structuring of the project, goal-oriented implementation of preparatory infrastructural measures and adequate scheduling of the coordinated activities of the installation and commissioning teams entrusted with the commissioning of the digital control system during the overhaul activities. (author)

  15. FFTF reactor assembly system technology

    International Nuclear Information System (INIS)

    Mangelsdorf, T.A.

    1975-01-01

    An overview is presented of the FFTF reactor and plant together with descriptions of core components, core internals, core system, primary and secondary control rod system, reactor instrumentation, reactor vessel and closure head, and supporting test programs

  16. Reactor System Design

    International Nuclear Information System (INIS)

    Chi, S. K.; Kim, G. K.; Yeo, J. W.

    2006-08-01

    SMART NPP(Nuclear Power Plant) has been developed for duel purpose, electricity generation and energy supply for seawater desalination. The objective of this project IS to design the reactor system of SMART pilot plant(SMART-P) which will be built and operated for the integrated technology verification of SMART. SMART-P is an integral reactor in which primary components of reactor coolant system are enclosed in single pressure vessel without connecting pipes. The major components installed within a vessel includes a core, twelve steam generator cassettes, a low-temperature self pressurizer, twelve control rod drives, and two main coolant pumps. SMART-P reactor system design was categorized to the reactor coe design, fluid system design, reactor mechanical design, major component design and MMIS design. Reactor safety -analysis and performance analysis were performed for developed SMART=P reactor system. Also, the preparation of safety analysis report, and the technical support for licensing acquisition are performed

  17. Retrofitting the instrumentation and control system of primary cooling circuit from TRIGA INR 14 MW reactor

    International Nuclear Information System (INIS)

    Preda, M.; Ciocanescu, M.; Ana, E. M.; Cristea, D.

    2008-01-01

    Activities of retrofitting the instrumentation and control system from TRIGA INR primary cooling circuit consists in replacement of actual system for: - parameter measurement; - safety; - reactor external scramming; - protection, command and supply for electrical elements of the system. This retrofitting project is designed to ensure the necessary features of reactor external safety and for technological parameter measurement. The new safety system of main cooling circuit is completely separated from its operating system and is arranged in a panel assembly in reactor control room. The operating system has the following features: - data acquisition; - parameter value and state of command elements displaying; - command elements on hierarchical levels; - operator information through visual and acoustic alarm. (authors)

  18. A method to determine the dampening system of control rod drop mechanism for PWR reactors

    International Nuclear Information System (INIS)

    Trindade, C.E.; Mattos, J.R.L. de; Perrotta, J.A.

    1988-08-01

    A method to determine the Control Assembly damping drop system (dashpot/guide tube) was developed. It's presented a theoretical model, an experimental device and the procedures to determine this system, which is used in PWR reactors. (author) [pt

  19. Control of advanced reactor-coupled heat exchanger system: Incorporation of reactor dynamics in system response to load disturbances

    Energy Technology Data Exchange (ETDEWEB)

    Skavdahi, Isaac; Utgikar, Vivek [Dept. of Chemical and Materials Engineering, University of Idaho, Moscow (United States); Christensen, Richard [Nuclear Engineering Program, University of Idaho, Idaho Falls (United States); Chen, Ming Hui; Sun, Xiao Dong [Nuclear Engineering Program, Department of Mechanical and Aerospace Engineering, The Ohio State University, Columbus (United States); Sabharwall, Piyush [Idaho National Laboratory, Idaho Falls (United States)

    2016-12-15

    Alternative control schemes for an Advanced High Temperature Reactor system consisting of a reactor, an intermediate heat exchanger, and a secondary heat exchanger (SHX) are presented in this paper. One scheme is designed to control the cold outlet temperature of the SHX (T{sub co}) and the hot outlet temperature of the intermediate heat exchanger (Th{sub o2}) by manipulating the hot-side flow rates of the heat exchangers (F{sub h}/F{sub h2}) responding to the flow rate and temperature disturbances. The flow rate disturbances typically require a larger manipulation of the flow rates than temperature disturbances. An alternate strategy examines the control of the cold outlet temperature of the SHX (T{sub co}) only, since this temperature provides the driving force for energy production in the power conversion unit or the process application. The control can be achieved by three options: (1) flow rate manipulation; (2) reactor power manipulation; or (3) a combination of the two. The first option has a quicker response but requires a large flow rate change. The second option is the slowest but does not involve any change in the flow rates of streams. The third option appears preferable as it has an intermediate response time and requires only a minimal flow rate change.

  20. Control of Advanced Reactor-Coupled Heat Exchanger System: Incorporation of Reactor Dynamics in System Response to Load Disturbances

    Directory of Open Access Journals (Sweden)

    Isaac Skavdahl

    2016-12-01

    Full Text Available Alternative control schemes for an Advanced High Temperature Reactor system consisting of a reactor, an intermediate heat exchanger, and a secondary heat exchanger (SHX are presented in this paper. One scheme is designed to control the cold outlet temperature of the SHX (Tco and the hot outlet temperature of the intermediate heat exchanger (Tho2 by manipulating the hot-side flow rates of the heat exchangers (Fh/Fh2 responding to the flow rate and temperature disturbances. The flow rate disturbances typically require a larger manipulation of the flow rates than temperature disturbances. An alternate strategy examines the control of the cold outlet temperature of the SHX (Tco only, since this temperature provides the driving force for energy production in the power conversion unit or the process application. The control can be achieved by three options: (1 flow rate manipulation; (2 reactor power manipulation; or (3 a combination of the two. The first option has a quicker response but requires a large flow rate change. The second option is the slowest but does not involve any change in the flow rates of streams. The third option appears preferable as it has an intermediate response time and requires only a minimal flow rate change.

  1. Control of advanced reactor-coupled heat exchanger system: Incorporation of reactor dynamics in system response to load disturbances

    International Nuclear Information System (INIS)

    Skavdahi, Isaac; Utgikar, Vivek; Christensen, Richard; Chen, Ming Hui; Sun, Xiao Dong; Sabharwall, Piyush

    2016-01-01

    Alternative control schemes for an Advanced High Temperature Reactor system consisting of a reactor, an intermediate heat exchanger, and a secondary heat exchanger (SHX) are presented in this paper. One scheme is designed to control the cold outlet temperature of the SHX (T_c_o) and the hot outlet temperature of the intermediate heat exchanger (Th_o_2) by manipulating the hot-side flow rates of the heat exchangers (F_h/F_h_2) responding to the flow rate and temperature disturbances. The flow rate disturbances typically require a larger manipulation of the flow rates than temperature disturbances. An alternate strategy examines the control of the cold outlet temperature of the SHX (T_c_o) only, since this temperature provides the driving force for energy production in the power conversion unit or the process application. The control can be achieved by three options: (1) flow rate manipulation; (2) reactor power manipulation; or (3) a combination of the two. The first option has a quicker response but requires a large flow rate change. The second option is the slowest but does not involve any change in the flow rates of streams. The third option appears preferable as it has an intermediate response time and requires only a minimal flow rate change

  2. The simulation study on the Nuclear Heating Reactor's power auto-control system

    International Nuclear Information System (INIS)

    Yang Zhijun; Liu Longzhi; Hu Guifen

    2000-01-01

    The power automatic control system on nuclear heating reactor (NHR) is a multi-input and multi-output non-linear system. The power automatic control system on NHR is studied by modern control theory. Through the simulation experiments, it is clear that adopting μ outer-loop and LQR inner-loop feedback, the best control results are obtained

  3. Simulation of the fuzzy-smith control system for the high temperature gas cooled reactor

    International Nuclear Information System (INIS)

    Li Deheng; Xu Xiaolin; Zheng Jie; Guo Renjun; Zhang Guifen

    1997-01-01

    The Fuzzy-Smith pre-estimate controller to solve the control of the big delay system is developed, accompanied with the development of the mathematical model of the 10 MW high temperature gas cooled test reactor (HTR-10) and the design of its control system. The simulation results show the Fuzzy-Smith pre-estimate controller has the advantages of both fuzzy control and Smith pre-estimate controller; it has better compensation to the delay and better adaptability to the parameter change of the control object. So it is applicable to the design of the control system for the high temperature gas cooled reactor

  4. The application and design of distributed control system in reactor shutdown system of Qinshan phase III

    International Nuclear Information System (INIS)

    Su Guoquan; Liu Wangtian; Yu Yijun; Xiong Weihua

    2006-03-01

    The design, commissioning and running of the reactor trip parameter monitoring system used in Qinshan Phase III are introduced. The applying technology of Distributed Control System realized trip parameter monitoring and realized the function of trip parameters quick data acquisitioning, transferring, saving, alarm, query. The applying of trip parameters monitoring system improved the abilities of plant status monitoring and event analyzing, and increased the security and economy of nuclear power plant. (authors)

  5. A method of reactor power decrease by 2DOF control system during BWR power oscillation

    International Nuclear Information System (INIS)

    Ishikawa, Nobuyuki; Suzuki, Katsuo

    1998-09-01

    Occurrence of power oscillation events caused by void feedback effects in BWRs operated at low-flow and high-power condition has been reported. After thoroughly examining these events, BWRs have been equipped with the SRI (Selected Rod Insertion) system to avoid the power oscillation by decreasing the power under such reactor condition. This report presents a power control method for decreasing the reactor power stably by a two degree of freedom (2DOF) control. Performing a numerical simulation by utilizing a simple reactor dynamics model, it is found that the control system designed attains a satisfactory control performance of power decrease from a viewpoint of setting time and oscillation. (author)

  6. Nuclear reactor shutdown system

    International Nuclear Information System (INIS)

    Mangus, J.D.; Cooper, M.H.

    1982-01-01

    An improved nuclear reactor shutdown system is described comprising a temperature sensitive device connected to control the electric power supply to a magnetic latch holding a body of a neutron absorbing material. The temperature sensitive device is exposed to the reactor coolant so that when the reactor coolant temperature rises above a specific level, the temperature sensitive device will cause deenergization of the magnetic latch to allow the body of neutron absorbing material to enter the reactor core. (author)

  7. Reactor system

    International Nuclear Information System (INIS)

    Miyano, Hiroshi; Narabayashi, Naoshi.

    1990-01-01

    The represent invention concerns a reactor system with improved water injection means to a pressure vessel of a BWR type reactor. A steam pump is connected to a heat removing system pipeline, a high pressure water injection system pipeline and a low pressure water injection system pipeline for injecting water into the pressure vessel. A pump actuation pipeline is disposed being branched from a main steam pump or a steam relieaf pipeline system, through which steams are supplied to actuate the steam pump and supply cooling water into the pressure vessel thereby cooling the reactor core. The steam pump converts the heat energy into the kinetic energy and elevates the pressure of water to a level higher than the pressure of the steams supplied by way of a pressure-elevating diffuser. Cooling water can be supplied to the pressure vessel by the pressure elevation. This can surely inject cooling water into the pressure vessel upon loss of coolant accident or in a case if reactor scram is necessary, without using an additional power source. (I.N.)

  8. Pressurized water reactor system model for control system design and analysis

    International Nuclear Information System (INIS)

    Cooper, K.F.; Cain, J.T.

    1975-01-01

    Satisfactory operation of present generation Pressurized Water Reactor (PWR) Nuclear Power systems requires that several independent and interactive control systems be designed. Since it is not practical to use an actual PWR system as a design tool, a mathematical model of the system must be developed as a design and analysis tool. The model presented has been developed to be used as an aid in applying optimal control theory to design and implement new control systems for PWR plants. To be applicable, the model developed must represent the PWR system in its normal operating range. For safety analysis the operating conditions of the system are usually abnormal and, therefore, the system modeling requirements are different from those for control system design and analysis

  9. MATLAB/SIMULINK platform for simulation of CANDU reactor control system

    International Nuclear Information System (INIS)

    Javidnia, H.; Jiang, J.

    2007-01-01

    In this paper a simulation platform for CANDU reactors' control system is presented. The platform is built on MATLAB/SIMULINK interactive graphical interface. Since MATLAB/SIMULINK are powerful tools to describe systems mathematically, all the subsystems in a CANDU reactor are represented in MATLAB's language and are implemented in SIMULINK graphical representation. The focus of the paper is on the flux control loop of CANDU reactors. However, the ideas can be extended to include other parts in CANDU power plants and the same technique can be applied to other types of nuclear reactors and their control systems. The CANDU reactor model and xenon feedback model are also discussed in this paper. (author)

  10. Analysis and upgrade of instrumentation and control systems for the modernization of research reactors

    International Nuclear Information System (INIS)

    1988-01-01

    This document provides assistance in the review and planning process for the upgrade of instrumentation and control systems (I and C systems) and related safety features of the reactor protection system for research reactors. In the interest of safety a need was realized to evaluate the performance of outdated I and C systems. An advisory group was assembled to develop guidelines and to provide recommendations for the upgrade of I and C systems. The recommendations on I and C systems upgrade contained in this document were developed by the advisory group using as guidelines the established safety criteria and operating standards for research reactors. 24 refs

  11. Reactor instrumentation and control

    International Nuclear Information System (INIS)

    Wach, D.; Beraha, D.

    1980-01-01

    The methods for measuring radiation are shortly reviewed. The instrumentation for neutron flux measurement is classified into out-of-core and in-core instrumentation. The out-of-core instrumentation monitors the operational range from the subcritical reactor to full power. This large range is covered by several measurement channels which derive their signals from counter tubes and ionization chambers. The in-core instrumentation provides more detailed information on the power distribution in the core. The self-powered neutron detectors and the aeroball system in PWR reactors are discussed. Temperature and pressure measurement devices are briefly discussed. The different methods for leak detection are described. In concluding the plant instrumentation part some new monitoring systems and analysis methods are presented: early failure detection methods by noise analysis, acoustic monitoring and vibration monitoring. The presentation of the control starts from an qualitative assessment of the reactor dynamics. The chosen control strategy leads to the definition of the part-load diagram, which provides the set-points for the different control systems. The tasks and the functions of these control systems are described. In additiion to the control, a number of limiting systems is employed to keep the reactor in a safe operating region. Finally, an outlook is given on future developments in control, concerning mainly the increased application of process computers. (orig./RW)

  12. A relay rack for a control and protection system for nuclear reactors

    International Nuclear Information System (INIS)

    Miyata, Yasuyuki; Oda, Noriaki; Akiyama, Toyoshi

    1975-01-01

    It is obvious that all the equipment in the various systems that constitute a nuclear power plant must exhibit the highest levels of reliability, but the reactor control and protection system is of vital importance, and thus it requires a particularly thorough approach, incorporating redundancy, independence and separation. The paper describes the functions, construction and specifications of the relay rack - one of the most important items of equipment for reactor control and protection in a generating facility using a pressurized-water reactor - and it gives details of the extent to which these three requirements are satisfied. (author)

  13. Auxiliary control system of the safety parameters for IPR-R1 reactor

    International Nuclear Information System (INIS)

    Coura, J.G.

    1986-01-01

    This paper deals with the description for the control of three cooling water parameters (conductivity, temperature and the maximum and minimum water levels) as well as the percent power fraction of the nuclear research reactor IPR-R1. In order to keep the reactor in good operation conditions, one permanent and accurate control of the cooling water is needed. The double monitoring of a fourth parameter, part of the original design, the percent power fraction, is obtained through the control of the uncompensated ion chamber current and aims to avoid the operation of the reactor without running the cooling system. (Author) [pt

  14. New digital control system for the operation of the Colombian research reactor IAN-R1

    International Nuclear Information System (INIS)

    Celis del A, L.; Rivero, T.; Bucio, F.; Ramirez, R.; Segovia, A.; Palacios, J.

    2015-09-01

    En 2011, Mexico won the Colombian international tender for the renewal of instrumentation and control of the IAN-R1 Reactor, to Argentina and the United States. This paper presents the design criteria and the development made for the new digital control system installed in the Colombian nuclear reactor IAN-R1, which is based on a redundant and diverse architecture, which provides increased availability, reliability and safety in the reactor operation. This control system and associated instrumentation met all national export requirements, with the safety requirements established by the IAEA as well as the requirements demanded by the Colombian Regulatory Body in nuclear matter. On August 20, 2012, the Colombian IAN-R1 reactor reached its first criticality controlled with the new system developed at Instituto Nacional de Investigaciones Nucleares (ININ). On September 14, 2012, the new control system of the Colombian IAN-R1 reactor was officially handed over to the Colombian authorities, this being the first time that Mexico exported nuclear technology through the ININ. Currently the reactor is operating successfully with the new control system, and has an operating license for 5 years. (Author)

  15. Reactor feedwater control device

    International Nuclear Information System (INIS)

    Koshi, Yuji.

    1993-01-01

    In the device of the present invention, an excess response is not caused in a reactor feed water system even when voids are fluctuated by using an actual water level signal as a reactor water level signal. That is, a standard water level signal and a reactor water level signal are inputted to a comparator. An adder adds water level difference signal outputted from the comparator and mismatch flow rate signal prepared by multiplying the difference between a main steam flow rate signal and a feed water flow rate signal by a mismatch gain. A feed water controller integrates the added signal and outputs flow rate demand signal. A feed water system receives the flow rate demand signal as input. A water level calculation means is disposed to such a device for calculating an actual water level based on the change of coolant possessing amount of the reactor, and the output thereof is defined as a reactor water level signal. With such procedures, excessive elevation of water level of the reactor can be prevented even upon occurrence of void fluctuation phenomenon or the like in the reactor such as upon sole scram operation. Accordingly, plant shut down caused thereby can be avoided safely. (I.S.)

  16. Flow rate control systems for coolants for BWR type reactors

    International Nuclear Information System (INIS)

    Igarashi, Yoko; Kato, Naoyoshi.

    1981-01-01

    Purpose: To increase spontaneous recycling flow rate of coolants in BWR type reactors when the water level in the reactor decreases, by communicating a downcomer with a lower plenum. Constitution: An opening is provided to the back plate disposed at the lower end of a reactor core shroud for communicating a downcomer with a lower plenum, and an ON-OFF valve actuated by an operation rod is provided to the opening. When abnormal water level or pressure in the reactor is detected by a level metal or pressure meter, the operation rod is driven to open the ON-OFF valve, whereby coolants fed from a jet pump partially flows through the opening to increase the spontaneous recycling flow rate of the coolants. This can increase the spontaneous recycling flow rate of the coolants upon spontaneous recycling operation, thereby maintaining the reactor safety and the fuel soundness. (Moriyama, K.)

  17. Contributions to the fundamental optimization and higher reliability of nuclear reactor complex control systems

    International Nuclear Information System (INIS)

    Tapu, C.

    1979-01-01

    The reference control system automatically adjusted to the reactor state and the reliability analysis of the control systems which survive the first defect with given solution for detecting all kinds of defects in the system with parallel redundancy, supply additional data on the high performance and largely available solutions. The results of the paper have been applied in the following cases: The reference control system automatically adapted to the reactor state has been tested with the VVR-S reactor and was patented. The solutions advanced for period and reactivity monitoring by the numerical technique were worked up during the training in France-CEN-Saclay and have been applied on the Ulysse reactor at INSTN (CERN-Saclay). At the same time, the papers published on these subjects have been cited in ''Revue General d'Electricite (February 1976)'' and in ''MCH/MENT 10 - Materiel Electronique Nucleaire pour tableau de commande et de controle (1976)'' issued by CEA France, as marking a progress in the control of nuclear power plants. The reliability analysis of the control system with parallel redundancy was performed during a specialization in France and on the basis of this analysis, the control system for the high flux reactor in Grenoble was selected. (author)

  18. Undergraduate reactor control experiment

    International Nuclear Information System (INIS)

    Edwards, R.M.; Power, M.A.; Bryan, M.

    1992-01-01

    A sequence of reactor and related experiments has been a central element of a senior-level laboratory course at Pennsylvania State University (Penn State) for more than 20 yr. A new experiment has been developed where the students program and operate a computer controller that manipulates the speed of a secondary control rod to regulate TRIGA reactor power. Elementary feedback control theory is introduced to explain the experiment, which emphasizes the nonlinear aspect of reactor control where power level changes are equivalent to a change in control loop gain. Digital control of nuclear reactors has become more visible at Penn State with the replacement of the original analog-based TRIGA reactor control console with a modern computer-based digital control console. Several TRIGA reactor dynamics experiments, which comprise half of the three-credit laboratory course, lead to the control experiment finale: (a) digital simulation, (b) control rod calibration, (c) reactor pulsing, (d) reactivity oscillator, and (e) reactor noise

  19. The control-and-instrumentation system of the IEA zero power reactor and its reliability calculation

    International Nuclear Information System (INIS)

    Peluso, M.A.V.

    1978-01-01

    The control-and instrumentation system for the Instituto de Energia Atomica Zero Power Reactor is described and the design criteria are presented and discussed. The reliability analysis for the reactor protection system was performed using the fault tree method. This was done using a computer code based on the Monte Carlo simulation. That code is an adaptation of the SAFTE-I, for the IBM 360/155 IEA Computer. (Author) [pt

  20. Digital control of research reactors

    International Nuclear Information System (INIS)

    Crump, J.C. III.; Richards, W.J.; Heidel, C.C.

    1991-01-01

    Research reactors provide an important service for the nuclear industry. Developments and innovations used for research reactors can be later applied to larger power reactors. Their relatively inexpensive cost allows research reactors to be an excellent testing ground for the reactors of tomorrow. One area of current interest is digital control of research reactor systems. Digital control systems offer the benefits of implementation and superior system response over their analog counterparts. At McClellan Air Force Base in Sacramento, California, the Stationary Neutron Radiography System (SNRS) uses a 1,000-kW TRIGA reactor for neutron radiography and other nuclear research missions. The neutron radiography beams generated by the reactor are used to detect corrosion in aircraft structures. While the use of the reactor to inspect intact F-111 wings is in itself noteworthy, there is another area in which the facility has applied new technology: the instrumentation and control system (ICS). The ICS developed by General Atomics (GA) contains several new and significant items: (a) the ability to servocontrol on three rods, (b) the ability to produce a square wave, and (c) the use of a software configurator to tune parameters affected by the actual reactor core dynamics. These items will probably be present in most, if not all, future research reactors. They were developed with increased control and overall usefulness of the reactor in mind

  1. Knowledge base expert system control of spatial xenon oscillations in pressurized water reactors

    International Nuclear Information System (INIS)

    Alten, S.

    1992-01-01

    Nuclear reactor operators are required to pay special attention to spatial xenon oscillations during the load-follow operation of pressurized water reactors. They are expected to observe the axial offset of the core, and to estimate the correct time and amount of necessary control action based on heuristic rules given in axial xenon oscillations are knowledge intensive, and heuristic in nature. An expert system, ACES (Axial offset Control using Expert Systems) is developed to implement a heuristic constant axial offset control procedure to aid reactor operators in increasing the plant reliability by reducing the human error component of the failure probability. ACES is written in a production system language, OPS5, based on the forward chaining algorithm. It samples reactor data with a certain time interval in terms of measurable parameters, such as the power, period, and the axial offset of the core. It then processes the core status utilizing a set of equations which are used in a back of the envelope calculations by domain experts. Heuristic rules of ACES identify the control variable to be used among the full and part length control rods and boron concentration, while a knowledge base is used to determine the amount of control. ACES is designed as a set of generic rules to avoid reducing the system into a set of patterns. Instead ACES evaluates the system, determines the necessary corrective actions in terms of reactivity insertion, and provides this reactivity insertion using the control variables. The amount of control action is determined using a knowledge base which consists of the differential rod worth curves, and the boron reactivity worth of a given reactor. Having the reactor dependent parameters in its knowledge base, ACES is applicable to an arbitrary reactor for axial offset control purposes

  2. High-throughput reactor system with individual temperature control for the investigation of monolith catalysts

    Science.gov (United States)

    Dellamorte, Joseph C.; Vijay, Rohit; Snively, Christopher M.; Barteau, Mark A.; Lauterbach, Jochen

    2007-07-01

    A high-throughput parallel reactor system has been designed and constructed to improve the reliability of results from large diameter catalysts such as monoliths. The system, which is expandable, consists of eight quartz reactors, 23.5mm in diameter. The eight reactors were designed with separate K type thermocouples and radiant heaters, allowing for the independent measurement and control of each reactor temperature. This design gives steady state temperature distributions over the eight reactors within 0.5°C of a common setpoint from 50to700°C. Analysis of the effluent from these reactors is performed using rapid-scan Fourier transform infrared (FTIR) spectroscopic imaging. The integration of this technique to the reactor system allows a chemically specific, truly parallel analysis of the reactor effluents with a time resolution of approximately 8s. The capabilities of this system were demonstrated via investigation of catalyst preparation conditions on the direct epoxidation of ethylene, i.e., on the ethylene conversion and the ethylene oxide selectivity. The ethylene, ethylene oxide, and carbon dioxide concentrations were calibrated based on spectra from FTIR imaging using univariate and multivariate chemometric techniques. The results from this analysis showed that the calcination conditions significantly affect the ethylene conversion, with a threefold increase in the conversion when the catalyst was calcined for 3h versus 12h at 400°C.

  3. Nuclear reactor safety systems

    International Nuclear Information System (INIS)

    Ball, R.M.; Roberts, R.C.

    1980-01-01

    A safety system for shutting down a nuclear reactor under overload conditions is described. The system includes a series of parallel-connected computer memory type look-up tables each of which receives data on a particular reactor parameter and in each of which a precalculated functional value for that parameter is stored indicative of the percentage of maximum reactor load that the parameter contributes. The various functional values corresponding to the actual measured parameters are added together to provide a control signal used to shut down the reactor under overload conditions. (U.K.)

  4. Design and application of leakage monitor for reactor and control rod driving system

    International Nuclear Information System (INIS)

    Li, Dongyu; Zou, Yimin; Ling, Qiu; Guo, Lanying

    2009-04-01

    By measuring the number of γ photons produced by the annihilation of the β + particles of 13 N's decay product in the sample air, the nuclide density of 13 N can be obtained, comparing with its density in the reactor coolant, we can get the leakage information of the reactor vessel and control rod driving system, the article describes the cause of improvement in monitoring for leakage of reactor vessel and control rod driving system of Qinshan Second Nuclear Power Plant (PWR reactor), also the determination of monitoring method and system configuration, as well as the main technical index and function. Furthermore, the main parts and its function of the monitor are introduced. After operation for more than four years, it is proved that both the stability and MTBF index of the monitor meet the design, even more, thanks to the improvement of the algorithm, the Compton Effect caused by other nuclide became neglectable, the MDA of the monitor was lowered also. (authors)

  5. New digital control and power protection system of VR 1 training reactor

    International Nuclear Information System (INIS)

    Kropik, M.; Matejka, K.; Juoeickova, M.

    2005-01-01

    The contribution describes the new VR-1 training reactor control and power protection system at the Czech Technical University in Prague. The control system provides safety and control functions, calculates average values of the important variables and sends data and system status to the human-machine interface. The upgraded control system is based on a high quality industrial PC. The operating system of the PC is the Microsoft Windows XP with the real time support RTX of the VentureCom Company. The software was developed according to requirements in MS Visual C. The independent power protection system is a component of the reactor safety (protection) system with high quality and reliability requirements. The digital system is redundant; each channel evaluates the reactor power and the velocity of power changes and provides safety functions. The digital part of the channel is multiprocessor-based. The software was developed with respect to nuclear standards. The software design was coded in the C language regarding the NRC restrictions. Configuration management, verification and validation accompanied the software development. Both systems were thoroughly tested. Firstly, the non active tests were carried out. During these tests, the active core of the reactor was subcritical; the input signals were generated from HPIB and VXI controlled instruments to simulate different operational and safety events. The software for instruments control and tests evaluation utilized Agilent VEE development system. After the successful non active checking, the active tests followed. (author)

  6. Study on the power control system for NPP power unit with the WWER-440 reactor

    International Nuclear Information System (INIS)

    Aleksandrova, N.D.; Naumov, A.V.

    1981-01-01

    Results of model investigations into basic version of the power control systems (PCS) conformably to the WWER-440 NPP power unit are stated. Transient processes in the power unit system when being two PCS versions during perturbations of different parameters: unit power, vapour pressure or position of control rods have been simulated. Investigations into the different PCS versions show that quality of operation of a traditional scheme with a turbine power controller and reactor pressure controller can be significantly improved with the introduction of a high-speed signal of pressure into the reactor controller. The PCS version with the compensation of interrelations between the turbine and reactor controllers constructed according to the same principles as the standard schemes of power units of thermal electric power plant is perspective as well [ru

  7. Design and Analysis of the Power Control System of the Fast Zero Energy Reactor FR-0

    Energy Technology Data Exchange (ETDEWEB)

    Schuh, N J.H.

    1966-12-15

    This report describes the power control by means of the fine-control rod and the design of the control system of the fast zero energy reactor FR-0 located in Studsvik, Sweden. System requirements and some operational conditions were used as design criteria. Manual and automatic control is possible. Variable electronic end-stops for the control rod have been designed, because of the special construction of the reactor and control rod. Noise in the control system caused by the reactor, detector and electronics caused disturbances of the control system at the lower power levels. The noise power-spectrum was measured. Statistical design methods, using the measured noise power spectrum, were used to design filters, which will reduce the influence of the noise at the lower power levels. Root Loci sketches and Bode diagrams were used for stability analyses. The system was simulated on an analogue computer, taking into account even nonlinearities of the control system and noise. Typical cases of reactor operation were simulated and stability analysis performed.

  8. Design and Analysis of the Power Control System of the Fast Zero Energy Reactor FR-0

    International Nuclear Information System (INIS)

    Schuh, N.J.H.

    1966-12-01

    This report describes the power control by means of the fine-control rod and the design of the control system of the fast zero energy reactor FR-0 located in Studsvik, Sweden. System requirements and some operational conditions were used as design criteria. Manual and automatic control is possible. Variable electronic end-stops for the control rod have been designed, because of the special construction of the reactor and control rod. Noise in the control system caused by the reactor, detector and electronics caused disturbances of the control system at the lower power levels. The noise power-spectrum was measured. Statistical design methods, using the measured noise power spectrum, were used to design filters, which will reduce the influence of the noise at the lower power levels. Root Loci sketches and Bode diagrams were used for stability analyses. The system was simulated on an analogue computer, taking into account even nonlinearities of the control system and noise. Typical cases of reactor operation were simulated and stability analysis performed

  9. A remotely controlled CCTV system for nuclear reactor retube operations

    International Nuclear Information System (INIS)

    Stovman, J.A.

    1984-01-01

    This paper describes the CCTV Vault Observation Subsystem (VOS) under development for Ontario Hydro for the Pickering 'A' Nuclear Power Plant Large Scale Retubing program. This subsystem will be used by a supervisor and several operators to observe fuel channel replacement operations following plant shutdown and removal of the fuel bundles. VOS basically comprises 23 monochrome television camera driven circuits, a matrix switcher, 15 monitors, 9 tape recorders and 4 microphone driven sound circuits. Remote control of the camera's zoom lenses and mounts is via a digitally multiplexed control system. Design considerations include viewing requirements, reliability, radiation, redundance, and economic factors

  10. 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.

  11. Performance analyses of the communication networks of a modern supervision and control system of research reactors

    International Nuclear Information System (INIS)

    El-Madbouly, E.I.; Shaat, M.K.; Shokr, A.M.; Elrefaei, G.H.

    2009-01-01

    The functions of the Instrumentation and Control (I and C) system in research reactors, the changes in its design according to the advances in the technology, and the internationally established safety requirements on the design and operational performance of this system are reviewed. The main features of the communication networks commonly used in the Supervision and Control systems (SCS) are presented. A methodology for the performance analysis of the communication networks of computer-based distributed SCS is developed and presented along with discussions. Application of this methodology to a modern SCS of a typical research reactor is illustrated. (orig.)

  12. Performance analyses of the communication networks of a modern supervision and control system of research reactors

    Energy Technology Data Exchange (ETDEWEB)

    El-Madbouly, E.I. [Menoufia Univ., Menouf (Egypt). Faculty of Electronics Engineering; Shaat, M.K.; Shokr, A.M.; Elrefaei, G.H. [Atomic Energy Authority, Abouzabal (Egypt). Egypt Second Research Reactor

    2009-04-15

    The functions of the Instrumentation and Control (I and C) system in research reactors, the changes in its design according to the advances in the technology, and the internationally established safety requirements on the design and operational performance of this system are reviewed. The main features of the communication networks commonly used in the Supervision and Control systems (SCS) are presented. A methodology for the performance analysis of the communication networks of computer-based distributed SCS is developed and presented along with discussions. Application of this methodology to a modern SCS of a typical research reactor is illustrated. (orig.)

  13. Description of the control and safety systems of the RA reactor

    International Nuclear Information System (INIS)

    Popovic, B.; Pesic, M.

    1962-01-01

    This report contains detailed description and scheme of the control and safety system of the RA reactor. It consists of interconnected five systems: for automated regulation; compensation rods; safety rods; power density measurement device; period meter; automated D 2 O level meter in the core. Automated regulation system is divided into two parts: basic system for reactor operation regime at power from 10kW - 10 MW and precise regulation system for operation at set-up power level up to 10 kW which is used occasionally

  14. Gray mode control or a device with added manoeuvrability: a new fine control system for pressurized water reactors

    International Nuclear Information System (INIS)

    Guilmin, J.

    1983-01-01

    The present fine control system for pressurized water reactors (the so-called ''A'' mode) cannot meet the requirements of a variable output in all respects. For this reason specialists are now developing a new fine control system, called gray mode control which is better suited to rapid load variations. The basic principles for control of pressurized water reactors are summarized and, then gray mode control or the device with added manoeuvrability is described. The series of tests carried out in 1981 and 1982 on stage 3 of the Tricastin power station are analyzed and the satisfactory results obtained are presented [fr

  15. Development of a nuclear reactor control system simulator using virtual instruments

    International Nuclear Information System (INIS)

    Pinto, Antonio Juscelino; Mesquita, Amir Zacarias; Lameiras, Fernando Soares

    2011-01-01

    The International Atomic Energy Agency recommends the use of safety and friendly interfaces for monitoring and controlling the operational parameters of the nuclear reactors. This article describes a digital system being developed to simulate the behavior of the operating parameters using virtual instruments. The control objective is to bring the reactor power from its source level (mW) to a full power (kW). It is intended for education of basic reactor neutronic and thermohydraulic principles such as the multiplication factor, criticality, reactivity, period, delayed neutron, control by rods, fuel and coolant temperatures, power, etc. The 250 kW IPR-R1 TRIGA research reactor at Nuclear Technology Development Centre - CDTN was used as reference. TRIGA reactors, developed by General Atomics (GA), are the most widely used research reactor in the world. The simulator system is being developed using the LabVIEW (Laboratory Virtual Instruments Engineering Workbench) software, considering the modern concept of virtual instruments (VI's) using electronic processor and visual interface in video monitor. The main purpose of the system is to provide training tools for instructors and students, allowing navigating by user-friendly operator interface and monitoring tendencies of the operational variables. It will be an interactive tool for training and teaching and could be used to predict the reactor behavior. Some scenarios are presented to demonstrate that it is possible to know the behavior of some variables from knowledge of input parameters. The TRIGA simulator system will allow the study of parameters, which affect the reactor operation, without the necessity of using the facility. (author)

  16. Development of a nuclear reactor control system simulator using virtual instruments

    Energy Technology Data Exchange (ETDEWEB)

    Pinto, Antonio Juscelino; Mesquita, Amir Zacarias; Lameiras, Fernando Soares, E-mail: ajp@cdtn.b, E-mail: amir@cdtn.b, E-mail: fsl@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2011-07-01

    The International Atomic Energy Agency recommends the use of safety and friendly interfaces for monitoring and controlling the operational parameters of the nuclear reactors. This article describes a digital system being developed to simulate the behavior of the operating parameters using virtual instruments. The control objective is to bring the reactor power from its source level (mW) to a full power (kW). It is intended for education of basic reactor neutronic and thermohydraulic principles such as the multiplication factor, criticality, reactivity, period, delayed neutron, control by rods, fuel and coolant temperatures, power, etc. The 250 kW IPR-R1 TRIGA research reactor at Nuclear Technology Development Centre - CDTN was used as reference. TRIGA reactors, developed by General Atomics (GA), are the most widely used research reactor in the world. The simulator system is being developed using the LabVIEW (Laboratory Virtual Instruments Engineering Workbench) software, considering the modern concept of virtual instruments (VI's) using electronic processor and visual interface in video monitor. The main purpose of the system is to provide training tools for instructors and students, allowing navigating by user-friendly operator interface and monitoring tendencies of the operational variables. It will be an interactive tool for training and teaching and could be used to predict the reactor behavior. Some scenarios are presented to demonstrate that it is possible to know the behavior of some variables from knowledge of input parameters. The TRIGA simulator system will allow the study of parameters, which affect the reactor operation, without the necessity of using the facility. (author)

  17. Modernization of turbine control system and reactor control system in Almaraz 1 and 2; MOdernizacion de los sistemas de control de turbina y del reactor en Almaraz 1 y 2

    Energy Technology Data Exchange (ETDEWEB)

    Pulido, C.; Diez, J.; Carrasco, J. A.; Lopez, L.

    2005-07-01

    The replacement of the Turbine Control System and Reactor Control System are part of the Almaraz modernization program for the Instrumentation and Control. For these upgrades Almaraz has selected the Ovation Platform that provides open architecture and easy expansion to other systems, these platforms is highly used in many nuclear and thermal plants around the world. One of the main objective for this project were to minimize the impact on the installation and operation of the plant, for that reason the project is implemented in two phases, Turbine Control upgrade and Reactor Control upgrade. Another important objective was to increase the reliability of the control system making them fully fault tolerant to single failures. The turbine Control System has been installed in Units 1 and 2 while the Reactor Control System will be installed in 2006 and 2007 outages. (Author)

  18. Reactor control system upgrade for the McClellan Nuclear Radiation Center Sacramento, CA

    International Nuclear Information System (INIS)

    Power, M. A.

    1999-01-01

    Argonne National Laboratory is currently developing a new reactor control system for the McClellan Nuclear Radiation Facility. This new control system not only provides the same functionality as the existing control system in terms of graphic displays of reactor process variables, data archival capability, and manual, automatic, pulse and square-wave modes of operation, but adds to the functionality of the previous control system by incorporating signal processing algorithms for the validation of sensors and automatic calibration and verification of control rod worth curves. With the inclusion of these automated features, the intent of this control system is not to replace the operator but to make the process of controlling the reactor easier and safer for the operator. For instance, an automatic control rod calibration method reduces the amount of time to calibrate control rods from days to minutes, increasing overall reactor utilization. The control rod calibration curve, determined using the automatic calibration system, can be validated anytime after the calibration, as long as the reactor power is between 50W and 500W. This is done by banking all of the rods simultaneously and comparing the tabulated rod worth curves with a reactivity computer estimate. As long as the deviation between the tabulated values and the reactivity estimate is within a prescribed error band, then the system is in calibration. In order to minimize the amount of information displayed, only the essential flux-related data are displayed in graphical format on the control screen. Information from the sensor validation methods is communicated to the operators via messages, which appear in a message window. The messages inform the operators that the actual process variables do not correlate within the allowed uncertainty in the reactor system. These warnings, however, cannot cause the reactor to shutdown automatically. The reactor operator has the ultimate responsibility of using this

  19. Nuclear reactor kinetics and control

    International Nuclear Information System (INIS)

    Lewins, J.

    1978-01-01

    A consistent, integrated account of modern developments in the study of nuclear reactor kinetics and the problem of their efficient and safe control. It aims to prepare the student for advanced study and research or practical work in the field. Special features include treatments of noise theory, reliability theory and safety related studies. It covers all aspects of the operation and control of nuclear reactors, power and research and is complete in providing physical data methods of calculation and solution including questions of equipment reliability. The work uses illustrations of the main types of reactors in use in the UK, USA and Europe. Each chapter contains problems and worked examples suitable for course work and study. The subject is covered in chapters, entitled: introductory review; neutron and precursor equations; elementary solutions at low power; linear reactor process dynamics with feedback; power reactor control systems; fluctuations and reactor noise; safety and reliability; nonlinear systems (safety and control); analogue computing. (author)

  20. Automatic start-up system of nuclear reactor based on sequence control technology

    International Nuclear Information System (INIS)

    Zhang Yao; Zhang Dafa; Peng Huaqing

    2009-01-01

    A conceptive design of an automatic start-up system based on the sequence control for the nuclear reactors is given in this paper, so as to solve the problems during the start-up process, such as the long operation time, low automatic control level and high accident rate. The start-up process and its requirements are analyzed in detail at first. Then,the principle, the architecture, the key technologies of the automatic start-up system of nuclear reactors are designed and discussed. With the designed system, the automatic start-up of the nuclear reactor can be realized,the work load of the operator can be reduced,and the safety and efficiency of the nuclear power plant during its start-up can be improved. (authors)

  1. Risk assessment of computer-controlled safety systems for fusion reactors

    International Nuclear Information System (INIS)

    Fryer, M.O.; Bruske, S.Z.

    1983-01-01

    The complexity of fusion reactor systems and the need to display, analyze, and react promptly to large amounts of information during reactor operation will require a number of safety systems in the fusion facilities to be computer controlled. Computer software, therefore, must be included in the reactor safety analyses. Unfortunately, the science of integrating computer software into safety analyses is in its infancy. Combined plant hardware and computer software systems are often treated by making simple assumptions about software performance. This method is not acceptable for assessing risks in the complex fusion systems, and a new technique for risk assessment of combined plant hardware and computer software systems has been developed. This technique is an extension of the traditional fault tree analysis and uses structured flow charts of the software in a manner analogous to wiring or piping diagrams of hardware. The software logic determines the form of much of the fault trees

  2. New approach to control the methanogenic reactor of a two-phase anaerobic digestion system.

    Science.gov (United States)

    von Sachs, Jürgen; Meyer, Ulrich; Rys, Paul; Feitkenhauer, Heiko

    2003-03-01

    A new control strategy for the methanogenic reactor of a two-phase anaerobic digestion system has been developed and successfully tested on the laboratory scale. The control strategy serves the purpose to detect inhibitory effects and to achieve good conversion. The concept is based on the idea that volatile fatty acids (VFA) can be measured in the influent of the methanogenic reactor by means of titration. Thus, information on the output (methane production) and input of the methanogenic reactor is available, and a (carbon) mass balance can be obtained. The control algorithm comprises a proportional/integral structure with the ratio of (a) the methane production rate measured online and (b) a maximum methane production rate expected (derived from the stoichiometry) as a control variable. The manipulated variable is the volumetric feed rate. Results are shown for an experiment with VFA (feed) concentration ramps and for experiments with sodium chloride as inhibitor.

  3. Computerized reactor monitor and control for research reactors

    International Nuclear Information System (INIS)

    Buerger, L.; Vegh, E.

    1981-09-01

    The computerized process control system developed in the Central Research Institute for Physics, Budapest, Hungary, is described together with its special applications at research reactors. The nuclear power of the Hungarian research reactor is controlled by this computerized system, too, while in Lybia many interesting reactor-hpysical calculations are built into the computerized monitor system. (author)

  4. Computer-based regulating control system for the Advanced Test Reactor

    International Nuclear Information System (INIS)

    Johnson, M.R.

    1983-01-01

    This paper describes a new control system which has recently been designed and installed at the Advanced Test Reactor at INEL, replacing an older system that had been in service for some 17 years. Based on modern digital technology, the new system provides improved capability, reliability, and an enhanced man/machine interface that includes comprehensive failure and error messages and voice synthesis. In addition to control functions, and transparent to the operator, the system performs continual on-line checks to sense subsystem failures and takes appropriate automatic action. In the maintenance mode, service technicians can carry on a dialog with the controller to quickly identify faulty components. The operational capabilities of the new system are summarized, and reactor operator training, experience, and acceptance of the system are discussed

  5. Nuclear reactor safety system

    International Nuclear Information System (INIS)

    Ball, R.M.; Roberts, R.C.

    1983-01-01

    The invention provides a safety system for a nuclear reactor which uses a parallel combination of computer type look-up tables each of which receives data on a particular parameter (from transducers located in the reactor system) and each of which produces the functional counterpart of that particular parameter. The various functional counterparts are then added together to form a control signal for shutting down the reactor. The functional counterparts are developed by analysis of experimental thermal and hydraulic data, which are used to form expressions that define safe conditions

  6. Implementation of an expert system for xenon spatial control in pressurized-water reactors

    International Nuclear Information System (INIS)

    Chung, S.K.

    1988-01-01

    Control of the axial xenon oscillations is a knowledge- and experience-intensive activity for reactor operators. To aid reactor operators in the control of axial xenon oscillations, an advisory expert system was developed. A rule-based expert system shell, INSIGHT2+, was used to build the expert system which was interfaced with a microcomputer-based core control model of a pressurized-water reactor, graphic engine, and data base. A core control model described by one-group diffusion theory with moderator temperature and xenon feedbacks was used to develop heuristic control rules and to test the system. Full- and part-length control rods, boron concentration, and coolant inlet temperature were considered as control variables of the core control model. This expert system consists of a search space: the set of possible power level and power shape patterns. The search space was made by combining the following core state variables: the sign of relative power and axial offset (AO) error, sign of the rate of change of power level and AO, and magnitude of relative power and AO error

  7. Reactor water chemistry control

    International Nuclear Information System (INIS)

    Kundu, A.K.

    2010-01-01

    Tarapur Atomic Power Station - 1 and 2 (TAPS) is a twin unit Boiling Water Reactors (BWRs) built in 1960's and operating presently at 160MWe. TAPS -1 and 2 are one of the vintage reactors operating in the world and belongs to earlier generation of BWRs has completed 40 years of successful, commercial and safe operation. In 1980s, both the reactors were de-rated from 660MWth to 530MWth due to leaks in the Secondary Steam Generators (SSGs). In BWR the feed water acts as the primary coolant which dissipates the fission heat and thermalises the fast neutrons generated in the core due to nuclear fission reaction and under goes boiling in the Reactor Pressure Vessel (RPV) to produce steam. Under the high reactor temperature and pressure, RPV and the primary system materials are highly susceptible to corrosion. In order to avoid local concentration of the chemicals in the RPV of BWR, chemical additives are not recommended for corrosion prevention of the system materials. So to prevent corrosion of the RPV and the primary system materials, corrosion resistant materials like stainless steel (of grade SS304, SS304L and SS316LN) is used as the structural material for most of the primary system components. In case of feed water system, main pipe lines are of carbon steel and the heater shell materials are of carbon steel lined with SS whereas the feed water heater tubes are of SS-304. In addition to the choice of materials, another equally important factor for corrosion prevention and corrosion mitigation of the system materials is maintaining highly pure water quality and strict water chemistry regime for both the feed water and the primary coolant, during operation and shutdown of the reactor. This also helps in controlled migration of corrosion product to and from the reactor core and to reduce radiation field build up across the primary system materials. Experience in this field over four decades added to the incorporation of modern techniques in detection of low

  8. Control system design for a 100 MW(th) research reactor

    International Nuclear Information System (INIS)

    Seshadri, S.N.; Ranganath, M.V.; Singh, Manjit.

    1983-01-01

    This paper presents the computer simulation carried out to evolve a suitable analog controller for a 100 MW(th) heavy water moderated research reactor under construction at Trombay. The control action is based on the average neutron flux in the reactor core and the reactivity is controlled by adjusting the moderator level in the calandria. A dual control scheme controlling the inflow as well as the outflow was adopted in order to fully exploit the capabilities of control elements. For reasons of reliability, the system consists of three identical channels enabling safe operation even under one channel failure. Based on the simulation studies a suitable compensation network was incorporated to achieve satisfactory system response. (author)

  9. Fuel cycle facility control system for the Integral Fast Reactor Program

    International Nuclear Information System (INIS)

    Benedict, R.W.; Tate, D.A.

    1993-01-01

    As part of the Integral Fast Reactor (IFR) Fuel Demonstration, a new distributed control system designed, implemented and installed. The Fuel processes are a combination of chemical and machining processes operated remotely. To meet this special requirement, the new control system provides complete sequential logic control motion and positioning control and continuous PID loop control. Also, a centralized computer system provides near-real time nuclear material tracking, product quality control data archiving and a centralized reporting function. The control system was configured to use programmable logic controllers, small logic controllers, personal computers with touch screens, engineering work stations and interconnecting networks. By following a structured software development method the operator interface was standardized. The system has been installed and is presently being tested for operations

  10. AFRRI's conversion to a microprocessor-based reactor instrumentation and control system

    International Nuclear Information System (INIS)

    Moore, Mark L.; Hodgdon, Kenneth M.

    1986-01-01

    The Armed Forces Radiobiology Research Institute (AFRRI) is procuring a state-of- the-art microprocessor-based instrumentation and control system to operate AFRRI's 1 MW (steady-state), 3000 MW (pulse) TRIGA Mark-F reactor. This system will replace the current control console while improving or maintaining the existing operational capabilities and safety characteristics. The new unit will have a 15-year design life using state-of-the-art components

  11. 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

  12. Electromagnetic drive of the control and protection system of a nuclear reactor

    International Nuclear Information System (INIS)

    Zav'yalova, G.I.

    1983-01-01

    The design and operating principle of an electromagnetic drive with a linear synchronous reaction motor are described. At the present time, electromagnetic control mechanisms using linear electric motors are finding increasingly widespread application as drives for the control and protection system of nuclear reactors. In these drives there is a functional mergence of the electromagnetic mechanism with the final control element; these drives, therefore, have advantages over electromechanical drives

  13. 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)

  14. FBR Italian position. The PEC reactor and its control rods system

    International Nuclear Information System (INIS)

    Artioli, C.

    1984-01-01

    The greatest effort in Italy concerning Fast Breeder Reactors is concentrated on the PEC project. This project represents in fact the sole Italian work on the FBR option. Up to now the state of progress of the PEC project is as follows: - Design work on reactor components and related systems: 80% - Civil engineering work: 45% - Components construction: 18%. It has recently been stated that construction of the plant must be finished by 1987. The progressive design of the PEC control system and the supporting research and development programmes are described in this paper

  15. Reactor power control device

    International Nuclear Information System (INIS)

    Ishii, Yoshihiko; Arita, Setsuo; Miyamoto, Yoshiyuki; Fukazawa, Yukihisa; Ishii, Kazuhiko

    1998-01-01

    The present invention provides a reactor power control device capable of enhancing an operation efficiency while keeping high reliability and safety in a BWR type nuclear power plant. Namely, the device of the present invention comprises (1) a means for inputting a set value of a generator power and a set value of a reactor power, (2) a means for controlling the reactor power to either smaller one of the reactor power corresponding to the set value of the generator power and the set value of the reactor power. With such procedures, even if the nuclear power plant is set so as to operate it to make the reactor power 100%, when the generator power reaches the upper limit, the reactor power is controlled with a preference given to the upper limit value of the generator power. Accordingly, safety and reliability are not deteriorated. The operation efficiency of the plant can be improved. (I.S.)

  16. Expert system technology for control integration in nuclear reactors

    International Nuclear Information System (INIS)

    Stabler, E.P. Jr.; Zimmerman, J.J.; Stratton, R.C.

    1986-03-01

    This report describes the role of expert system technology in nuclear power plant operation. The use of computers to assist operator decisions would greatly enhance the safety and efficiency of operation. A description of the necessary operator interfaces, data acquisition and validation, plant status and parameter diagnosis, and system reliability is presented. (FL)

  17. Design and realization experience of Advanced Control Rod Group and Individual Control System (GIC) for VVER-1000 reactors

    International Nuclear Information System (INIS)

    Cerny, V.; Novy, L.; Janour, J.; Ris, M.; Zidek, P.

    1997-01-01

    During the reactor refueling outage of unit 1 of the South Ukrainian nuclear power plant in mid-1996, full replacement of the reactor's group and individual control (GIC) system was performed. The main functions of the GIC system are briefly characterized. The structure of the advanced GIC system is described and shown by means of a diagram. The criteria used in deciding on the upgrading strategy are discussed in some detail. The implementation of the replacement is also dealt with, as is the testing and commissioning of the system. (A.K.)

  18. Recent advances in reactor protection and control system technology

    International Nuclear Information System (INIS)

    Anon.

    1997-01-01

    After a first-generation digital integrated protection system has been installed on all 1300 MWe PWR units in France, a new digital protection system was developed for the 1450 MWe units, using local area networks, fiber optics, Motorola 68000 microprocessors, and a modular design allowing for the design of any system on the basis of around 50 types of standard cards. In 1993, an upgrading program for this equipment was launched in order to reduce costs, in particular software development costs, further improve hardware modularity and facilitate integration and connection to existing equipment. The basic principles of the units are described together with the implementation of computer-aided software engineering (CASE) tools, interfaces with hard-wired equipment, and multiplexed connections. The nuclear instrumentation systems at the Fessenheim and Bugey plants have been renovated with these equipment

  19. Design of the decision aiding system for the control of the research nuclear reactor

    International Nuclear Information System (INIS)

    Adda, F.; Allek, M.; Larbes, C.

    2003-01-01

    Intelligent and decision aiding systems as support to operators are becoming increasingly a necessity in nuclear installations and in nuclear reactors in particular, specially after the Tree Mile Island. Development of new technologies based on linguistic approaches such as fuzzy logic has given rise to much interest during the last years. Fuzzy logic controller (FLC) has many advantage compared to conventional controllers using classical techniques. The aim of the present work is to use a fuzzy logic controller in parallel to actual semi-automatic controller in order to supervise in real time the operation of the research nuclear reactor. The principal of this controller is based on rules which are established previous from experiment using the semi-automatic controller and from the knowledge of the operators. (authors)

  20. Design of the Control System for Engineered Safety Features of KIJANG Research Reactor

    International Nuclear Information System (INIS)

    Kim, Hagtae; Kim, Jun-Yeon; Chae, Hee-Taek

    2015-01-01

    The purpose of this paper is to design an effective control system for the Engineered Safety Features (ESF) of KJRR such as the Safety Residual Heat Removal System (SRHRS) pumps and Siphon Break Valve (SBV) without an Engineered Safety Features-Component Control System (ESF-CCS). This control system is called a 'local motor starter', because this system controls motors in the SRHRS pumps and SBVs by receiving the signal from Reactor Protection System (RPS) and Alternate Protection System (APS) when the differential pressure or pool level reach the set points. In this paper, the design concepts and requirements of the local motor starter based on the design features of KJRR is proposed. An ESF is a safety system that mitigates consequences of the Anticipated Operational Occurrence (AOO) and Design Basis Accident (DBA). The results of this paper are able to be used for the development of control systems for research reactors similar to KJRR. The precondition for such application is to have a few ESFs and conduct simple logic. The proposed control system called a local motor starter is being designed, and a manufacture of the actual systems is expected in the foreseeable future

  1. Closed loop auto control system software for Miniature Neutron Source Reactors (MNSRs)

    International Nuclear Information System (INIS)

    Iqbal, Masood; Qadir, Javed

    2009-01-01

    A closed loop auto control system has been developed and implemented at the Pakistan Research Reactor-2 (PARR-2). For interfacing of the site signals with a computer, a data acquisition card was employed. The system utilizes nine signals and, on their basis, carries out continuous on-line analysis, maintaining safe operation. On-line reactivity measurement is included in this software. Also, it generates control signals for keeping the reactor at the desired power level in auto-mode round-the-clock without human intervention. In case of abnormal conditions for either power or temperature or radiation level, alarms are initiated and if their levels reach beyond prescribed safe operation regime, automatic shutdown sequence is initiated. The control system has been thoroughly tested for various postulated scenarios. The test results have been in good agreement with the expected response. (orig.)

  2. Hardware-in-the-Loop Simulation for the Automatic Power Control System of Research Reactors

    International Nuclear Information System (INIS)

    Fikry, R.M.; Shehata, S.A.; Elaraby, S.M.; Mahmoud, M.I.; Elbardini, M.M.

    2009-01-01

    Designing and testing digital control system for any nuclear research reactor can be costly and time consuming. In this paper, 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 thc 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-Ioop 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. Two kinds of controllers are studied, Proportional derivative (PD) 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

  3. Modelling of reactor control and protection systems in the core simulator program GARLIC

    International Nuclear Information System (INIS)

    Beraha, D.; Lupas, O.; Ploegert, K.

    1984-01-01

    For analysis of the interaction between control and limitation systems and the power distribution in the reactor core, a valuable tool is provided by the joint simulation of the core and the interacting systems. To this purpose, the core simulator GARLIC has been enhanced by models of the systems for controlling and limiting the reactor power and the power distribution in the core as well as by modules for calculating safety related core parameters. The computer-based core protection system, first installed in the Grafenrheinfeld NPP, has been included in the simulation. In order to evaluate the accuracy of GARLIC-simulations, the code has been compared with a design code in the train of a verification phase. The report describes the program extensions and the results of the verification. (orig.) [de

  4. Development of advanced automatic control system for nuclear ship. 2. Perfect automatic operation after reactor scram events

    International Nuclear Information System (INIS)

    Yabuuchi, Noriaki; Nakazawa, Toshio; Takahashi, Hiroki; Shimazaki, Junya; Hoshi, Tsutao

    1997-11-01

    An automatic operation system has been developed for the purpose of realizing a perfect automatic plant operation after reactor scram events. The goal of the automatic operation after a reactor scram event is to bring the reactor hot stand-by condition automatically. The basic functions of this system are as follows; to monitor actions of the equipments of safety actions after a reactor scram, to control necessary control equipments to bring a reactor to a hot stand-by condition automatically, and to energize a decay heat removal system. The performance evaluation on this system was carried out by comparing the results using to Nuclear Ship Engineering Simulation System (NESSY) and the those measured in the scram test of the nuclear ship 'Mutsu'. As the result, it was showed that this system had the sufficient performance to bring a reactor to a hot syand-by condition quickly and safety. (author)

  5. Development of advanced automatic control system for nuclear ship. 2. Perfect automatic operation after reactor scram events

    Energy Technology Data Exchange (ETDEWEB)

    Yabuuchi, Noriaki; Nakazawa, Toshio; Takahashi, Hiroki; Shimazaki, Junya; Hoshi, Tsutao [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1997-11-01

    An automatic operation system has been developed for the purpose of realizing a perfect automatic plant operation after reactor scram events. The goal of the automatic operation after a reactor scram event is to bring the reactor hot stand-by condition automatically. The basic functions of this system are as follows; to monitor actions of the equipments of safety actions after a reactor scram, to control necessary control equipments to bring a reactor to a hot stand-by condition automatically, and to energize a decay heat removal system. The performance evaluation on this system was carried out by comparing the results using to Nuclear Ship Engineering Simulation System (NESSY) and the those measured in the scram test of the nuclear ship `Mutsu`. As the result, it was showed that this system had the sufficient performance to bring a reactor to a hot syand-by condition quickly and safety. (author)

  6. A digitized wide range channel for new instrumentation and control system of PUSPATI TRIGA Reactor (RTP)

    International Nuclear Information System (INIS)

    Zareen Khan Abdul Jalil Khan; Izhar Abu Hussin; Mohd Idris Taib; Nurfarhana Ayuni Joha; Roslan Md Dan

    2010-01-01

    Wide Range Channel is one of very important part of Reactor Instrumentation and Control system. Current system is using all analog system. The main functions of the new system are to provide Wide-log power and Multi-range linear power. The other functions are to provide Percent power and Power rate of change. The linear power level range is up to 125 % and the log power system to cover from below source level to 150 %. The main function of digital signal processor is for pulse shaping, pulse counting and root mean square signal processing. The system employs automatic on-line self diagnostics and calibration verification. (author)

  7. Pulse-density modulation control of chemical oscillation far from equilibrium in a droplet open-reactor system

    OpenAIRE

    Sugiura, Haruka; Ito, Manami; Okuaki, Tomoya; Mori, Yoshihito; Kitahata, Hiroyuki; Takinoue, Masahiro

    2016-01-01

    The design, construction and control of artificial self-organized systems modelled on dynamical behaviours of living systems are important issues in biologically inspired engineering. Such systems are usually based on complex reaction dynamics far from equilibrium; therefore, the control of non-equilibrium conditions is required. Here we report a droplet open-reactor system, based on droplet fusion and fission, that achieves dynamical control over chemical fluxes into/out of the reactor for c...

  8. Construction of the monitoring, processing and logging systems supporting for management, operation and maintenance of the Dalat reactor control system

    International Nuclear Information System (INIS)

    Trinh Dinh Hai; Nguyen Thanh Cuong; Huynh Ton Nghiem; Phan Quoc Minh; Nguyen Duc Tuan; Nguyen Nhi Dien

    2004-01-01

    From 1/2002 to 12/2003, we implemented successfully a project, entitled 'Construction of the monitoring, processing and logging systems supporting for management, operation and maintenance of the Dalat reactor control system' under the assistance of the Ministry of Science and Technology. Its main results such as Testing Apparatus based on microcontroller for all functional boards of the Control Logic System of the Reactor Control System (RCS). Technical support CD - ROM for Process Instrumentation System, software for logging automatically information from important systems of the RCS through LAN, program for failure management of Process Instrumentation System have been playing an important role for observation, operation support, maintenance of the RCS. Through this project, the implementation group has grown up rapidly. The control and instrumentation group has been provided with some modern equipment, electronic components, and materials for maintenance work and research development in the years to come. This paper presents typical results and discussions. (author)

  9. 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)

  10. The resonance absorption controlled reactor

    International Nuclear Information System (INIS)

    Caro, R.

    1977-01-01

    In this report a new method of reactor control based on tho isotopic moderator composition variation is studied, taking as a reference a D 2 O/H 2 O system. With this method an spectacular increment in the burn-up degree and a sensible reduction of the conventional control system is obtained. An important part of this work has been the detailed analysis of the parameters affecting the neutron spectrum in a heterogeneous reactor. (Author) 50 refs

  11. The resonance absorption controlled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Caro, R

    1977-07-01

    In this report a new method of reactor control based on tho isotopic moderator composition variation is studied, taking as a reference a D{sub 2}O/H{sub 2}O system. With this method an spectacular increment in the burn-up degree and a sensible reduction of the conventional control system is obtained. An important part of this work has been the detailed analysis of the parameters affecting the neutron spectrum in a heterogeneous reactor. (Author) 50 refs.

  12. Design requirements of instrumentation and control systems for next generation reactor

    International Nuclear Information System (INIS)

    Koo, In Soo; Lee, Byung Sun; Park, Kwang Hyun; Park, Heu Yoon; Lee, Dong Young; Kim, Jung Taek; Hwang, In Koo; Chung, Chul Hwan; Hur, Seop; Kim, Chang Hoi; Na, Nan Ju

    1994-03-01

    In this report, the basic design requirements of Instrumentation and Control systems for next generation reactor are described, which are top-tier level, to support the advanced I and C systems. It contains the requirements in accordance with the plant reliability, the plant performance, the operator's aid functions, the features for maintenance and testing, licensing issues for I and C systems. Advanced I and C systems are characterized such as the application of the digital and the human engineering technologies. To development of this requirements, the I and C systems for the foreign passive and the evolutionary types of reactor and the domestic conventional reators were reviewed and anlysed. At the detail design stage, these requirements will be used for top-tier requirements. To develop the detail design requirements in the future, more quantitive and qualitive analyses are need to be added. (Author) 44 refs

  13. Design requirements of instrumentation and control systems for next generation reactor

    Energy Technology Data Exchange (ETDEWEB)

    Koo, In Soo; Lee, Byung Sun; Park, Kwang Hyun; Park, Heu Yoon; Lee, Dong Young; Kim, Jung Taek; Hwang, In Koo; Chung, Chul Hwan; Hur, Seop; Kim, Chang Hoi; Na, Nan Ju [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1994-03-01

    In this report, the basic design requirements of Instrumentation and Control systems for next generation reactor are described, which are top-tier level, to support the advanced I and C systems. It contains the requirements in accordance with the plant reliability, the plant performance, the operator`s aid functions, the features for maintenance and testing, licensing issues for I and C systems. Advanced I and C systems are characterized such as the application of the digital and the human engineering technologies. To development of this requirements, the I and C systems for the foreign passive and the evolutionary types of reactor and the domestic conventional reators were reviewed and anlysed. At the detail design stage, these requirements will be used for top-tier requirements. To develop the detail design requirements in the future, more quantitive and qualitive analyses are need to be added. (Author) 44 refs.

  14. Design concept of control system for cryogenic distillation columns of fusion reactor

    International Nuclear Information System (INIS)

    Yamanishi, Toshihiko; Okuno, Kenji

    1993-09-01

    Control systems were designed for cryogenic distillation columns in the main fuel cycle and the breeder blanket interface systems of fusion reactors. Three basic control modes were proposed for the column whose top product was more important; the column whose bottom product is more important; and the column having a feed back stream. The key component in the important product stream was selected for each column, and the analysis method for measurement of this key component was discussed. Some of the columns need the gas chromatography as the analysis instrument of the control system. The time required for the measurement of product purity by the gas chromatography considerably affects the stability of the control system. A significant conclusion is that permissible time is about 20 min. It is possible to complete the measurement within 20 minute by the gas chromatography. The gas chromatography is applicable for the control system of the column. (author)

  15. Automated systems help prevent operator error during [reactor] I and C [instrumentation and control] testing

    International Nuclear Information System (INIS)

    Courcoux, R.

    1989-01-01

    On a nuclear steam supply system, even a minor failure can involve actuation of the whole reactor protection system (RPS). To reduce the likelihood of human error leading to unwanted trips during the maintenance of instrumentation and control systems, Framatome has been developing and installing various automated testing systems. Such automated systems are particularly helpful when periodic tests with a potential for RPS actuation have to be carried out, or when the test is on the critical path for the refuelling outage. The Sensitive Channel Programme described is an example of the sort of work that has been done. (author)

  16. 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

  17. Control of xenon spatial oscillations during load follow of nuclear reactor via robust servo systems

    International Nuclear Information System (INIS)

    Ukai, Hiroyuki; Yada, Yukihiro; Iwazumi, Tetsuo; Morita, Yoshifumi.

    1990-01-01

    This paper investigates the control problem of xenon spatial oscillations in the axial direction during load following operations of a nuclear reactor. The system model is described by a one-group diffusion equation with xenon and power feedbacks and iodine-xenon dynamic equations and controlled by full-length and part-length control rods. In order to achieve the control purpose we formulate the control model as the design problem of robust servo systems for distributed parameter reactor systems. Hence the total thermal power and the axial offset are chosen as outputs to be controlled. The control law is designed based upon finite-dimensional systems which are constructed by linearizing around steady states, approximating by the Galerkin approximate method and reducing dimensions via the singular perturbation method. From a computational point of view a simple computational algorithm to obtain an approximate solution of the steady state neutron balance is developed via the perturbation method. Some results of numerical simulations are represented to show effectiveness of the theory developed in this paper. Particularly it is shown that the designed servo systems are robust against model errors with the linearization and the model truncation. (author)

  18. Plant level of automated control system at a NPP with RBMK reactor

    International Nuclear Information System (INIS)

    Vorob'ev, V.P.; Gorbunov, V.P.; Dmitriev, V.M.; Litvin, A.S.

    1987-01-01

    The functional structure of plant level automated control system (ACS) at NPP with RBMK-1000 reactors, its binding with the on-line control system of higher and lower levels, as well as engineering requirements to software and recommendations on composition of hardware components, are considered. NPP ACS is an organizational-engineering system consisting of computer facilities and binding aimed at solving management, economical, organizational and physical-engineering problems to control NPP more effectively. The system carries out data acquisition, preliminary processing, analysis, transmission and representation for users to accept solutions for NPP operation by operative and management personnel. The main aim of integrated NPP ACS is the control development and increase of NPP economical efficiency, the increase of electric and heat energy production, the optimization of the production distribution between units, the development of production and economic NPP control

  19. Model Reduction Using Proper Orthogonal Decomposition and Predictive Control of Distributed Reactor System

    Directory of Open Access Journals (Sweden)

    Alejandro Marquez

    2013-01-01

    Full Text Available This paper studies the application of proper orthogonal decomposition (POD to reduce the order of distributed reactor models with axial and radial diffusion and the implementation of model predictive control (MPC based on discrete-time linear time invariant (LTI reduced-order models. In this paper, the control objective is to keep the operation of the reactor at a desired operating condition in spite of the disturbances in the feed flow. This operating condition is determined by means of an optimization algorithm that provides the optimal temperature and concentration profiles for the system. Around these optimal profiles, the nonlinear partial differential equations (PDEs, that model the reactor are linearized, and afterwards the linear PDEs are discretized in space giving as a result a high-order linear model. POD and Galerkin projection are used to derive the low-order linear model that captures the dominant dynamics of the PDEs, which are subsequently used for controller design. An MPC formulation is constructed on the basis of the low-order linear model. The proposed approach is tested through simulation, and it is shown that the results are good with regard to keep the operation of the reactor.

  20. Comparison of control systems applied to the handling of radioactive reactor components

    International Nuclear Information System (INIS)

    Robinson, C.; Harris, E.G.; Dyer, P.C.; Williams, J.G.B.

    1985-01-01

    The first generation of nuclear power stations have individual reactors each incorporating complete facilities for servicing components and refuelling. In the later designs, each power station has two reactors which are connected by a central block. This central block contains one set of facilities to service both reactors, but to improve the station capability, some of these are to be replicated. The central block incorporates a hoist well which was used during construction for the accessing of complete components. On completion of this work, the physical size of the hoist well is such as to permit the incorporation of additional facilities if these are shown to be operationally and economically desirable. Since a number of years of power operation has elapsed, the advantages of back-fitting to existing fuel-handling facilities has been illustrated. Since the mechanical arrangements and operating procedures are substantially similar for both the original and new handling facilities, the paper will illustrate the control systems provided for each. The configuration of the system is arranged to have two channels of control which complies with the current standard requirements in the United Kingdom. These requirements are more stringent than when the existing facility was designed and constructed, as described in the relevant sections of the paper. The new system has been designed and is being manufactured to comply with the Central Electricity Generating Board standard for nuclear fuel route interlock and control systems. (author)

  1. Advanced control system for the Integral Fast Reactor fuel pin processor

    International Nuclear Information System (INIS)

    Lau, L.D.; Randall, P.F.; Benedict, R.W.; Levinskas, D.

    1993-01-01

    A computerized control system has been developed for the remotely-operated fuel pin processor used in the Integral Fast Reactor Program, Fuel Cycle Facility (FCF). The pin processor remotely shears cast EBR- reactor fuel pins to length, inspects them for diameter, straightness, length, and weight, and then inserts acceptable pins into new sodium-loaded stainless-steel fuel element jackets. Two main components comprise the control system: (1) a programmable logic controller (PLC), together with various input/output modules and associated relay ladder-logic associated computer software. The PLC system controls the remote operation of the machine as directed by the OCS, and also monitors the machine operation to make operational data available to the OCS. The OCS allows operator control of the machine, provides nearly real-time viewing of the operational data, allows on-line changes of machine operational parameters, and records the collected data for each acceptable pin on a central data archiving computer. The two main components of the control system provide the operator with various levels of control ranging from manual operation to completely automatic operation by means of a graphic touch screen interface

  2. Reactor control device

    International Nuclear Information System (INIS)

    Kameda, Akiyuki.

    1979-01-01

    Purpose: To enable three types of controls, that is, level control, scram control and excess reactivity control required for a reactor by a same mechanism by feeding neutron absorber liquid and pressure control gas to several blind pipes provided in the reactor core. Constitution: A plurality of blind pipes are disposed spaced apart in a reactor core and connected by way of injection pipes to a neutron absorber liquid tank. A pressure regulator is connected to the blind pipes, to which pressure control gas is supplied. The neutron absorber liquid used herein consists of sodium, potassium or their alloy, or mercury as a basic substance incorporated with one or more selected from boron, tantalum, rhenium, europium or their compounds. The level control, scram control and excess reactivity control can be attained by moderating the pressure changes in the pressure control gas or by regulating the fluctuation in the liquid level. (Horiughi, T.)

  3. Results of neutron physics analyses of WWER-440 cores with modified reactor protection and control systems

    International Nuclear Information System (INIS)

    Lehmann, M.; Pecka, M.; Rocek, J.; Zalesky, K.

    1993-12-01

    Detailed results are given of neutron physics analyses performed to assess the efficiency and acceptability of modifications of the WWER-440 core protection and control system; the modifications have been proposed with a view to increasing the proportion of mechanical control in the compensation of reactivity effects during reactor unit operation in the variable load mode. The calculations were carried out using the modular MOBY-DICK macrocode system together with the SMV42G36 library of two-group parametrized diffusion constants, containing corrections which allow new-design WWER-440 fuel assemblies to be discriminated. (J.B). 37 tabs., 18 figs., 5 refs

  4. Improvement of remote control system of automatic ultrasonic equipment for inspection of reactor pressure vessel

    International Nuclear Information System (INIS)

    Cheong, Yong Moo; Jung, H. K.; Joo, Y. S.; Koo, K. M.; Hyung, H.; Sim, C. M.; Gong, U. S.; Kim, S. H.; Lee, J. P.; Rhoo, H. C.; Kim, M. S.; Ryoo, S. K.; Choi, C. H.; Oh, K. I.

    1999-12-01

    One of the important issues related to the nuclear safety is in-service inspection of reactor pressure vessel (RPV). A remote controlled automatic ultrasonic method is applied to the inspection. At present the automatic ultrasonic inspection system owned by KAERI is interrupted due to degradation of parts. In order to resume field inspection new remote control system for the equipment was designed and installed to the existing equipment. New ultrasonic sensors and their modules for RPV inspection were designed and fabricated in accordance with the new requirements of the inspection codes. Ultrasonic sensors were verified for the use in the RPV inspection. (author)

  5. Improvement of remote control system of automatic ultrasonic equipment for inspection of reactor pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Cheong, Yong Moo; Jung, H. K.; Joo, Y. S.; Koo, K. M.; Hyung, H.; Sim, C. M.; Gong, U. S.; Kim, S. H.; Lee, J. P.; Rhoo, H. C.; Kim, M. S.; Ryoo, S. K.; Choi, C. H.; Oh, K. I

    1999-12-01

    One of the important issues related to the nuclear safety is in-service inspection of reactor pressure vessel (RPV). A remote controlled automatic ultrasonic method is applied to the inspection. At present the automatic ultrasonic inspection system owned by KAERI is interrupted due to degradation of parts. In order to resume field inspection new remote control system for the equipment was designed and installed to the existing equipment. New ultrasonic sensors and their modules for RPV inspection were designed and fabricated in accordance with the new requirements of the inspection codes. Ultrasonic sensors were verified for the use in the RPV inspection. (autho0008.

  6. Design and adjustment on test bed of replacing subassembly machine control system for China experimental fast reactor

    International Nuclear Information System (INIS)

    Dong Shengguo; Ma Hongsheng; Zhao Lixia

    2008-01-01

    The present research concerns in the design and adjustment of replacing sub- assembly machine control system of China Experimental Fast Reactor. The design of replacing subassembly machine control system adopts some electric equipments, such as programmable controllers, digital DC drivers. The designed control system was adjusted on the test bed. The results indicate that the operation of the control system is steady and reliable, and designed control system can meet the needs of the design specification. (authors)

  7. Theoretical study of an on-off automatic control system for a nuclear reactor

    International Nuclear Information System (INIS)

    Menezes, J.; Jover, P.

    1964-01-01

    The automatic control system designed for a high flux nuclear reactor is of the 'constant speed-dead zone' type, in which the control rod is run at normal speed in the required direction when the error signal overrides a preset level. The study of the closed loop absolute stability was carried out with the describing function method. An analog computer study yielded the optimal values of the setting parameters (relay hysteresis, motor response time), which lead to a minimization of the control steps frequency when the reactivity varies slowly with the xenon poisoning. (authors) [fr

  8. Design of power auto-regulating system's high reliability controller for 200 MW nuclear heating reactor

    International Nuclear Information System (INIS)

    An Zhencai; Liu Longzhi; Chen Yuan

    1996-01-01

    The paper mainly introduces power auto-regulating system's high reliability controller for 200 MW Nuclear Heating Reactor. The controller is implemented with excellent performance 16 bit single chip microcomputer 8097. Master controller and 10 digit samplers are blocked. Each and every block's hardware is identical. These blocks communicate each other through 8 bit BUS and operate synchronously by united clock and reset signal and are designed with three redundancies. The identity comparison principle through two-out-of three is also introduced. The test proves that designing scheme is feasible

  9. Nuclear reactor control column

    International Nuclear Information System (INIS)

    Bachovchin, D.M.

    1982-01-01

    The nuclear reactor control column comprises a column disposed within the nuclear reactor core having a variable cross-section hollow channel and containing balls whose vertical location is determined by the flow of the reactor coolant through the column. The control column is divided into three basic sections wherein each of the sections has a different cross-sectional area. The uppermost section of the control column has the greatest crosssectional area, the intermediate section of the control column has the smallest cross-sectional area, and the lowermost section of the control column has the intermediate cross-sectional area. In this manner, the area of the uppermost section can be established such that when the reactor coolant is flowing under normal conditions therethrough, the absorber balls will be lifted and suspended in a fluidized bed manner in the upper section. However, when the reactor coolant flow falls below a predetermined value, the absorber balls will fall through the intermediate section and into the lowermost section, thereby reducing the reactivity of the reactor core and shutting down the reactor

  10. Pulse-density modulation control of chemical oscillation far from equilibrium in a droplet open-reactor system.

    Science.gov (United States)

    Sugiura, Haruka; Ito, Manami; Okuaki, Tomoya; Mori, Yoshihito; Kitahata, Hiroyuki; Takinoue, Masahiro

    2016-01-20

    The design, construction and control of artificial self-organized systems modelled on dynamical behaviours of living systems are important issues in biologically inspired engineering. Such systems are usually based on complex reaction dynamics far from equilibrium; therefore, the control of non-equilibrium conditions is required. Here we report a droplet open-reactor system, based on droplet fusion and fission, that achieves dynamical control over chemical fluxes into/out of the reactor for chemical reactions far from equilibrium. We mathematically reveal that the control mechanism is formulated as pulse-density modulation control of the fusion-fission timing. We produce the droplet open-reactor system using microfluidic technologies and then perform external control and autonomous feedback control over autocatalytic chemical oscillation reactions far from equilibrium. We believe that this system will be valuable for the dynamical control over self-organized phenomena far from equilibrium in chemical and biomedical studies.

  11. First application of the Bailey Control Micro-Z system. Control of sodium purity in the Super Phenix reactor

    International Nuclear Information System (INIS)

    Hureau, J.-C.

    1979-01-01

    The system for testing the purity of sodium in the Super Phenix reactor at Crey-Malville is one of the very first applications of the 'μz' Bailey Control system which, as is generally known, uses a microprocessor on each regulator. This characteristic is cleverly exploited here for discontinuous regulation, in stages, of an analog parameter: temperature, with a quite remarkable flexibility of use and an extremely reduced volume of equipment [fr

  12. Reactor control device

    International Nuclear Information System (INIS)

    Araki, Takao; Inoue, Toyokazu.

    1981-01-01

    Purpose: To protect the reactor floor by alleviating the shock imparted to the reactor floor by a dropped control rod when a wire rope accidentally breaks. Constitution: A control rod is hung by wire rope from a control rod drive, and shock absorbers are mounted at the upper and lower portions of the control rod. The outer diameter of the upper shock absorber is made larger than the inner diameter of a control rod inserting hole formed in the reactor core. If the control rod drops, the upper absorber is stopped at the upper tapered portion of the inserting hole. Thus, the dropping energy of the control rod can be sufficiently absorbed by the upper and lower shock absorbers. (Kamimura, M.)

  13. Computerized reactor power regulation with logarithmic controller

    International Nuclear Information System (INIS)

    Gossanyi, A.; Vegh, E.

    1982-11-01

    A computerized reactor control system has been operating at a 5 MW WWR-SM research reactor in the Central Research Institute for Physics, Budapest, for some years. This paper describes the power controller used in the SPC operating mode of the system, which operates in a 5-decade wide power range with +-0.5% accuracy. The structure of the controller easily limits the minimal reactor period and produces a reactor transient with constant period if the power demand changes. (author)

  14. Reactor power control device

    International Nuclear Information System (INIS)

    Doi, Kazuyori.

    1981-01-01

    Purpose: To automatically control the BWR type reactor power by simple and short-time searching the load pattern nearest to the required pattern at a nuclear power plant side. Constitution: The reactor power is automatically regulated by periodical modifying of coefficients fitting to a reactor core model, according as a required load pattern. When a load requirement pattern is given, a simulator estimates the total power change and the axial power distribution change from a xenon density change output calculated by a xenon dynamic characteristic estimating device, and a load pattern capable of being realized is searched. The amount to be recirculated is controlled on the basis of the load patteren thus searched, and the operation of the BWR type reactor is automatically controlled at the side of the nuclear power plant. (Kamimura, M.)

  15. Nuclear reactor power control device

    International Nuclear Information System (INIS)

    Koshi, Yuji; Sakata, Akira; Karatsu, Hiroyuki.

    1987-01-01

    Purpose: To control abrupt changes in neutron fluxes by feeding back a correction signal obtained from a deviation between neutron fluxes and heat fluxes for changing the reactor core flow rate to a recycling flow rate control system upon abrupt power change of a nuclear reactor. Constitution: In addition to important systems, that is, a reactor pressure control system and a recycling control system in the power control device of a BWR type power plant, a control circuit for feeding back a deviation between neutron fluxes and heat fluxes to a recycling flow rate control system is disposed. In the suppression circuit, a deviation signal is prepared in an adder from neutron flux and heat flux signals obtained through a primary delay filter. The deviation signal is passed through a dead band and an advance/delay filter into a correction signal, which is adapted to be fed back to the recycling flow rate control system. As a result, the reactor power control can be conducted smoothly and it is possible to effectively suppress the abrupt change or over shoot of the neutron fluxes and abrupt power change. (Kamimura, M.)

  16. A multi-agent design for a pressurized water reactor (P.W.R.) control system

    International Nuclear Information System (INIS)

    Aimar-Lichtenberger, M.

    1999-01-01

    This PhD work is in keeping with the complex industrial process control. The starting point is the analysis of control principles in a Pressurized Water Reactor (P.W.R). In order to cope with the limits of the present control procedures, a new control organisation by objectives and means is defined. This functional organisation is based on the state approach and is characterized by the parallel management of control functions to ensure the continuous control of the installation essential variables. With regard to this complex system problematic, we search the most adapted computer modeling. We show that a multi-agent system approach brings an interesting answer to manage the distribution and parallelism of control decisions and tasks. We present a synthetic study of multi-agent systems and their application fields.The choice of a multi-agent approach proceeds with the design of an agent model. This model gains experiences from other applications. This model is implemented in a computer environment which combines the mechanisms of an object language with Prolog. We propose in this frame a multi-agent modeling of the control system where each function is represented by an agent. The agents are structured in a hierarchical organisation and deal with different abstraction levers of the problem. Following a prototype process, the validation is realized by an implementation and by a coupling to a reactor simulator. The essential contributions of an agent approach turn on the mastery of the system complexity, the openness, the robustness and the potentialities of human-machine cooperation. (author)

  17. Tritium control in fusion reactor materials: A model for Tritium Extracting System

    Energy Technology Data Exchange (ETDEWEB)

    Zucchetti, Massimo [DENERG, Politecnico di Torino (Italy); Utili, Marco, E-mail: marco.utili@enea.it [ENEA UTIS – C.R. Brasimone, Bacino del Brasimone, Camugnano, BO (Italy); Nicolotti, Iuri [DENERG, Politecnico di Torino (Italy); Ying, Alice [University of California Los Angeles (UCLA), Los Angeles, CA (United States); Franza, Fabrizio [Karlsruhe Institute of Technology, Karlsruhe (Germany); Abdou, Mohamed [University of California Los Angeles (UCLA), Los Angeles, CA (United States)

    2015-10-15

    Highlights: • A modeling work has been performed to address these issues in view of its utilization for the TES (Tritium Extraction System), in the case of the HCPB TBM and for a Molecular sieve as adsorbent material. • A computational model has been setup and tested in this paper. • The results of experimental measurement of fundamental parameters such as mass transfer coefficients have been implemented in the model. • It turns out the capability to model the extraction process of gaseous tritium compounds and to estimate the breakthrough curves of the two main tritium gaseous species (H2 and HT). - Abstract: In fusion reactors, tritium is bred by lithium isotopes inside the blanket and then extracted. However, tritium can contaminate the reactor structures, and can be eventually released into the environment. Tritium in reactor components should therefore be kept under close control throughout the fusion reactor lifetime, bearing in mind the risk of accidents, the need for maintenance and the detritiation of dismantled reactor components before their re-use or disposal. A modeling work has been performed to address these issues in view of its utilization for the TES (Tritium Extraction System), in the case of the HCPB TBM and for a molecular sieve as adsorbent material. A computational model has been setup and tested. The results of experimental measurement of fundamental parameters such as mass transfer coefficients have been implemented in the model. It turns out the capability of the model to describe the extraction process of gaseous tritium compounds and to estimate the breakthrough curves of the two main tritium gaseous species (H2 and HT).

  18. Tritium control in fusion reactor materials: A model for Tritium Extracting System

    International Nuclear Information System (INIS)

    Zucchetti, Massimo; Utili, Marco; Nicolotti, Iuri; Ying, Alice; Franza, Fabrizio; Abdou, Mohamed

    2015-01-01

    Highlights: • A modeling work has been performed to address these issues in view of its utilization for the TES (Tritium Extraction System), in the case of the HCPB TBM and for a Molecular sieve as adsorbent material. • A computational model has been setup and tested in this paper. • The results of experimental measurement of fundamental parameters such as mass transfer coefficients have been implemented in the model. • It turns out the capability to model the extraction process of gaseous tritium compounds and to estimate the breakthrough curves of the two main tritium gaseous species (H2 and HT). - Abstract: In fusion reactors, tritium is bred by lithium isotopes inside the blanket and then extracted. However, tritium can contaminate the reactor structures, and can be eventually released into the environment. Tritium in reactor components should therefore be kept under close control throughout the fusion reactor lifetime, bearing in mind the risk of accidents, the need for maintenance and the detritiation of dismantled reactor components before their re-use or disposal. A modeling work has been performed to address these issues in view of its utilization for the TES (Tritium Extraction System), in the case of the HCPB TBM and for a molecular sieve as adsorbent material. A computational model has been setup and tested. The results of experimental measurement of fundamental parameters such as mass transfer coefficients have been implemented in the model. It turns out the capability of the model to describe the extraction process of gaseous tritium compounds and to estimate the breakthrough curves of the two main tritium gaseous species (H2 and HT).

  19. Control system of an anaerobia reactor used in the treatment of the Industrial residual waters

    International Nuclear Information System (INIS)

    Duque, Mauricio; Giraldo, Eugenio; Bello Frank

    1995-01-01

    The technology of the anaerobia digestion, has had a wide acceptance in the Colombian means for the treatment of industrial residual waters, especially for the economic advantages that it present and the good purification results. The technology of the anaerobia digestion for the treatment of residual waters, is based in the conversion of the organic matter present in the polluted waters, in methane and carbon dioxide. These two gases are removed of the reactor by means of special structures of gathering. Microorganisms that are sensitive to the changes of the pH mediate the conversion of the organic matter to CH4 and CO2. Therefore, the control on the pH is necessary for a correct behavior of the reactor. At the moment many industries are implementing plans of contamination control, that involve treatment of residual waters for means anaerobia. The present investigation is part of a wide work program in the technology of the anaerobia digestion. It is looked for to develop a monitored system and automatic control of reactors discharge anaerobia appraises, in a combined effort among the departments of Civil and Electric Engineering of the Andes University

  20. Monitoring and control of the Rossendorf research reactor using a microcomputerized automation system

    International Nuclear Information System (INIS)

    Ba weg, F.; Enkelmann, W.; Klebau, J.

    1982-01-01

    A decentral hierarchic information system (HIS) is presented, which has been developed for monitoring and control of the Rossendorf Research Reactor RFR, but which may also be considered the prototype of a digital automation system (AS) to be used in power stations. The functions integrated in the HIS are as follows: process monitoring, process control, and use of a specialized industrial robot for control of charging and discharging of the materials to be irradiated. The AS is realized on the basis of the process computer system PRA 30 (A 6492) developed in the GDR and including a computer K 1630 and the intelligent process terminals ursadat 5000 connected by a fast serial interface (IFLS). (author)

  1. Reactor core control device

    International Nuclear Information System (INIS)

    Sano, Hiroki

    1998-01-01

    The present invention provides a reactor core control device, in which switching from a manual operation to an automatic operation, and the control for the parameter of an automatic operation device are facilitated. Namely, the hysteresis of the control for the operation parameter by an manual operation input means is stored. The hysteresis of the control for the operation parameter is collected. The state of the reactor core simulated by an operation control to which the collected operation parameters are manually inputted is determined as an input of the reactor core state to the automatic input means. The record of operation upon manual operation is stored as a hysteresis of control for the operation parameter, but the hysteresis information is not only the result of manual operation of the operation parameter. This is results of operation conducted by a skilled operator who judge the state of the reactor core to be optimum. Accordingly, it involves information relevant to the reactor core state. Then, it is considered that the optimum automatic operation is not deviated greatly from the manual operation. (I.S.)

  2. Control of WWER-440 nuclear reactor

    International Nuclear Information System (INIS)

    Wagner, K.; Drab, F.; Grof, V.

    1978-01-01

    The V-1 reactor control systems are described. The data acquisition and processing system fulfils four main functions, ie., reactor start-up and power increase to 10% of the rated power, automatic power control within 3% and 110% of the rated power, reactivity compensation, and reactor protection. The automatic control system ensures constant steam pressure maintained with an accuracy of +-0.05 MPa. Reactivity compensation and spatial power distribution is mainly safeguarded by boric acid control. The V-1 reactor protection system has four levels of accident protection depending on the gravity of the failure. The philosophy of automation of the V-1 reactor control and protection system is based on autonomous automatic controlers and on the direct control of the individual sets and technological equipment. In conclusion, development trends are briefly outlined of control and protection systems of light water reactor power plants. (Z.M.)

  3. Dynamic Modeling and Control of Nuclear Reactors Coupled to Closed-Loop Brayton Cycle Systems using SIMULINKTM

    International Nuclear Information System (INIS)

    Wright, Steven A.; Sanchez, Travis

    2005-01-01

    The operation of space reactors for both in-space and planetary operations will require unprecedented levels of autonomy and control. Development of these autonomous control systems will require dynamic system models, effective control methodologies, and autonomous control logic. This paper briefly describes the results of reactor, power-conversion, and control models that are implemented in SIMULINK TM (Simulink, 2004). SIMULINK TM is a development environment packaged with MatLab TM (MatLab, 2004) that allows the creation of dynamic state flow models. Simulation modules for liquid metal, gas cooled reactors, and electrically heated systems have been developed, as have modules for dynamic power-conversion components such as, ducting, heat exchangers, turbines, compressors, permanent magnet alternators, and load resistors. Various control modules for the reactor and the power-conversion shaft speed have also been developed and simulated. The modules are compiled into libraries and can be easily connected in different ways to explore the operational space of a number of potential reactor, power-conversion system configurations, and control approaches. The modularity and variability of these SIMULINK TM models provides a way to simulate a variety of complete power generation systems. To date, both Liquid Metal Reactors (LMR), Gas Cooled Reactors (GCR), and electric heaters that are coupled to gas-dynamics systems and thermoelectric systems have been simulated and are used to understand the behavior of these systems. Current efforts are focused on improving the fidelity of the existing SIMULINK TM modules, extending them to include isotopic heaters, heat pipes, Stirling engines, and on developing state flow logic to provide intelligent autonomy. The simulation code is called RPC-SIM (Reactor Power and Control-Simulator)

  4. Advanced Instrumentation, Information, and Control Systems Technologies Research in Support of Light Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hallbert, Bruce P.; Kenneth, Thomas [Idaho National Laboratory, Idaho (United States)

    2014-08-15

    The Advanced Instrumentation, Information, and Control (II and C) Systems Technologies Pathway conducts targeted research and development (R and D) to address aging and reliability concerns with the legacy instrumentation and control and related information systems of the U.S. operating light water reactor (LWR) fleet. This work involves two major goals to ensure that legacy analog II and C systems are not life-limiting issues for the LWR fleet, and to implement digital II and C technology in a manner that enables broad innovation and business improvement in the nuclear power plant operating model. Resolving long-term operational concerns with the II and C systems contributes to the long-term sustainability of the LWR fleet, which is vital to the nation's energy and environmental security.

  5. Advanced Instrumentation, Information, and Control Systems Technologies Research in Support of Light Water Reactors

    International Nuclear Information System (INIS)

    Hallbert, Bruce P.; Kenneth, Thomas

    2014-01-01

    The Advanced Instrumentation, Information, and Control (II and C) Systems Technologies Pathway conducts targeted research and development (R and D) to address aging and reliability concerns with the legacy instrumentation and control and related information systems of the U.S. operating light water reactor (LWR) fleet. This work involves two major goals to ensure that legacy analog II and C systems are not life-limiting issues for the LWR fleet, and to implement digital II and C technology in a manner that enables broad innovation and business improvement in the nuclear power plant operating model. Resolving long-term operational concerns with the II and C systems contributes to the long-term sustainability of the LWR fleet, which is vital to the nation's energy and environmental security

  6. Aging assessment of the boiling-water reactor (BWR) standby liquid control system

    International Nuclear Information System (INIS)

    Orton, R.D.; Johnson, A.B.; Buckley, G.D.; Larson, L.L.

    1992-10-01

    Pacific Northwest Laboratory conducted a Phase I aging assessment of the standby liquid control (SLC) system used in boiling-water reactors. The study was based on detailed reviews of SLC system component and operating experience information obtained from the Nuclear Plant Reliability Database System, the Nuclear Document System, Licensee Event Reports, and other databases. Sources dealing with sodium pentaborate, borates, boric acid, and the effects of environment and corrosion in the SLC system were reviewed to characterize chemical properties and corrosion characteristics of borated solutions. The leading aging degradation concern to date appears to be setpoint drift in relief valves, which has been discovered during routine surveillance and is thought to be caused by mechanical wear. Degradation was also observed in pump seals and internal valves. In general, however, the results of the Phase I study suggest that age-related degradation of SLC systems has not been serious

  7. Use of in-core self-powered detectors in the nuclear reactor control system

    International Nuclear Information System (INIS)

    Mitel'man, M.G.; Andreev, L.G.; Batenin, I.V.

    1975-01-01

    The paper describes the results of experiments conducted at the Obninsk atomic power station on the establishment of an integrated system for the control and automatic regulation of the distribution of energy liberation. The pick-ups used for this system are direct-charge detectors with rhodium emitters. The system, which consists of 12 integral DPZ-7 detectors distributed evenly on 2 mutually perpendicular diamaters and 2 assemblies, each containing 4 1n direct-charge detectors, monitors the distribution of neutron flow density according to core height. Increased signals from the detectors were fed into the summation block, in which the outgoing signal was also adjusted. The half-life of the rhodium was calculated and the non-inertia of the entire system ascertained. The total operating time of the system for the automatic control of reactor power is approximately 1 h. The influence of interference was not determined. During the experiment, the power of the reactor remained constant, i.e. its regulation by the direct-charge detectors inside the core was effective. (author)

  8. BWR reactor management system

    International Nuclear Information System (INIS)

    Makino, Kakuji; Kawamura, Atsuo; Yoshioka, Ritsuo; Neda, Toshikatsu.

    1979-01-01

    It is necessary to grasp the delicate state of operation in reactor cores in view of the control of burn-up and power output at the time of the operation management of BWRs. Enormous labor has been required for the collection, processing and evaluation of the data. It is desirable to obtain the safer, more efficient and faster method of operation control by predicting the states in cores including the change of xenon and reflecting them to operation plans as well as by tracing with high accuracy the past burn-up history for a long period. At present, the on-line evaluation of the states in cores is carried out with the process computers attached to respective units, but the amount of data required for core operation management of high degree far exceeds their capacity. From such viewpoints, the research and development on the reactor management system were carried out. The data processing concerning core operation management is performed with newly installed computers utilizing the data from existing process computers, and the operation of reactor cores, the qualitative improvement of management works, labor saving, and fast, efficient operation control are feasible with it. This system was installed in an actual plant in October, 1977. The composition of the system, the prediction of the change in local output distribution accompanying control rod operation, the prediction of the change in the states in cores due to the flow rate of coolant, and the function of collecting plant data are explained. (Kako, I.)

  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. Reactor power control device

    International Nuclear Information System (INIS)

    Watanabe, Mitsutaka

    1997-01-01

    Hardware of an analog nuclear instrumentation system is reformed, a function generator is added to a setting calculation circuit of the nuclear instrumentation system, and each of setting lines of the nuclear instrumentation system is set in parallel with an upper limit curve in an operation region defined by a second order or third order equation. Upon transient change of abnormal power elevation during operation, scram signals are generated by power change in the same state as 100% rated operation due to elevation of reactor thermal power. Since the operation limit value relative to transient change due to power elevation can be made substantially equal with the same as that upon rated operation, the operation limit value for partial power operation state can be kept substantially the same level as that upon rated operation. When transition change caused by abnormal control rod withdrawal occurs during operation, a control rod withdrawal inhibition signal can ensure the power elevation width equal with that upon rated power operation, and since the withdrawal inhibition signal is generated in substantially the same withdrawing state, the operation limit value relative to a partial power operation state can be kept at the same level as that during rated operation. (N.H.)

  11. Implementation of nuclear power plant simulation in start-up commissioning of reactor control system

    International Nuclear Information System (INIS)

    Yang Zongwei; Huang Tieming; Feng Guangyu; Luan Zhenhua; Lin Meng; Zhu Lizhi

    2009-01-01

    Based on the nuclear power thermal-hydraulic model, Labview graphical programming language and virtual instrument data acquisition technology, this paper describes a dedicate test platform to solve the problem that the reactor control system (RRC) can not be evaluated and analyzed far before the actual startup of the unit. By connecting the test platform to the nuclear Digital Control System (DCS), the step-by-step closed-looped test and global function test of RRC system were performed, the dynamic validation and logical function demonstration for RRC were realized, and a lot of configuration mistakes of RRC and nonconformity were solved. The test for unit 3 of Ling'ao phase II has proved that the implementation of nuclear power plant simulation in the start-up commissioning of RRC can greatly reduce the risk of normal power operation and great transient tests, with which the term of startup for overall unit test can be greatly shortened. (authors)

  12. Reactivity control system of a passively safe thorium breeder pebble bed reactor

    International Nuclear Information System (INIS)

    Wols, F.J.; Kloosterman, J.L.; Lathouwers, D.; Hagen, T.H.J.J. van der

    2014-01-01

    Highlights: • A worth of over 15,000 pcm ensures achieving long-term cold shutdown in thorium PBR. • Control rod worth in side reflector is insufficient due to low-power breeder zone. • 20 control rods, just outside the driver zone, can achieve long-term cold shutdown. • BF 3 gas can be inserted for reactor shutdown, but only in case of emergency. • Perturbation theory accurately predicts absorber gas worth for many concentrations. - Abstract: This work investigates the neutronic design of the reactivity control system for a 100 MW th passively safe thorium breeder pebble bed reactor (PBR), a conceptual design introduced previously by the authors. The thorium PBR consists of a central driver zone of 100 cm radius, surrounded by a breeder zone with 300 cm outer radius. The fissile content of the breeder zone is low, leading to low fluxes in the radial reflector region. Therefore, a significant decrease of the control rod worth at this position is anticipated. The reactivity worth of control rods in the side reflector and at alternative in-core positions is calculated using different techniques, being 2D neutron diffusion, perturbation theory and more accurate 3D Monte Carlo models. Sensitivity coefficients from perturbation theory provide a first indication of effective control rod positions, while the 2D diffusion models provide an upper limit on the reactivity worth achievable at a certain radial position due to the homogeneous spreading of the absorber material over the azimuthal domain. Three dimensional forward calculations, e.g. in KENO, are needed for an accurate calculation of the total control rod worth. The two dimensional homogeneous calculations indicate that the reactivity worth in the radial reflector is by far insufficient to achieve cold reactor shutdown, which requires a control rod worth of over 15 000 pcm. Three dimensional heterogeneous KENO calculations show that placing 20 control rods just outside the driver channel, between 100 cm

  13. Reactivity control system of a passively safe thorium breeder pebble bed reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wols, F.J., E-mail: f.j.wols@tudelft.nl; Kloosterman, J.L.; Lathouwers, D.; Hagen, T.H.J.J. van der

    2014-12-15

    Highlights: • A worth of over 15,000 pcm ensures achieving long-term cold shutdown in thorium PBR. • Control rod worth in side reflector is insufficient due to low-power breeder zone. • 20 control rods, just outside the driver zone, can achieve long-term cold shutdown. • BF{sub 3} gas can be inserted for reactor shutdown, but only in case of emergency. • Perturbation theory accurately predicts absorber gas worth for many concentrations. - Abstract: This work investigates the neutronic design of the reactivity control system for a 100 MW{sub th} passively safe thorium breeder pebble bed reactor (PBR), a conceptual design introduced previously by the authors. The thorium PBR consists of a central driver zone of 100 cm radius, surrounded by a breeder zone with 300 cm outer radius. The fissile content of the breeder zone is low, leading to low fluxes in the radial reflector region. Therefore, a significant decrease of the control rod worth at this position is anticipated. The reactivity worth of control rods in the side reflector and at alternative in-core positions is calculated using different techniques, being 2D neutron diffusion, perturbation theory and more accurate 3D Monte Carlo models. Sensitivity coefficients from perturbation theory provide a first indication of effective control rod positions, while the 2D diffusion models provide an upper limit on the reactivity worth achievable at a certain radial position due to the homogeneous spreading of the absorber material over the azimuthal domain. Three dimensional forward calculations, e.g. in KENO, are needed for an accurate calculation of the total control rod worth. The two dimensional homogeneous calculations indicate that the reactivity worth in the radial reflector is by far insufficient to achieve cold reactor shutdown, which requires a control rod worth of over 15 000 pcm. Three dimensional heterogeneous KENO calculations show that placing 20 control rods just outside the driver channel

  14. Radiation control monitoring system on the High Temperature Engineering Test Reactor

    International Nuclear Information System (INIS)

    Minowa, Y.; Nakazawa, T.; Sato, K.; Kikuchi, H.; Nomura, T.

    1999-01-01

    Radiation control monitoring system of the HTTR is divided into three subsystems; exhaust monitoring equipment, room air monitoring equipment, dose equivalent rate monitoring equipment. The exhaust monitoring equipment consists of exhaust gas monitors, exhaust dust monitors, and a tritium and carbon sampling device at normal operation of the reactor. Accident gas monitors are also provided for the emergency. The tritium and carbon sampling device uses cupper oxide as a oxidizer, and ethanol amine as a sampling materials which collects continuously tritium and carbon in dust during about one month and is measured by a liquid scintillation counter. The accident gas monitors consist of two channels, for a low and a high range. The high range-gas monitor consists of two ionization chambers: one encloses argon gas and the other encloses xenon gas. Average energy of various gamma-rays, hence, accident exposure dose of the public can be estimated with the comparison of the sensitivity of two kinds of ionization chambers. The dose equivalent rate monitoring equipment consists of silicon semiconductor detectors for gamma-ray, a ionization chamber for gamma-ray, a BF 3 counter for neutron, and accident area monitors which are located in the reactor container. The message of 'check dose !' or 'temporary evacuation !' can be send to the workers in the reactor with a light and a sound. A computer system collects the radiation monitoring data every 10 sec cycle and accumulates them in a server computer. The leakage and the dispersion of helium gas must be taken into account on the radiation control monitoring system of the HTTR. (Suetake, M.)

  15. Dynamic model for the control system simulation and design of a 200 MW nuclear heating reactor

    International Nuclear Information System (INIS)

    Zhang Yuai; Liu Longzhi; Ma Changwen

    1999-01-01

    The author develops a nonlinear dynamic model used in a wide range control system simulation for a 200 MW Nuclear Heating Reactor (NHR-200). Besides a one-point neutron kinetics equation and temperature feedback based on the lumped fuel and coolant temperature, which are the usual methods used in modeling of PWR, two other factors are also considered in order to suit the wide range operation. The first consideration is the natural circulation in the primary loop because it affects the heat transfer coefficients in the core and in the primary heat exchanger (PHE). The second consideration is the flow rate variation in the secondary loop which leads to some nonlinear properties. The simulation results show that the model is accurate enough for control system simulation. Some model reduction basis can be obtained through the dynamic analysis

  16. The Programmable Logic Controller and its application in nuclear reactor systems

    Energy Technology Data Exchange (ETDEWEB)

    Palomar, J.; Wyman, R. [Lawrence Livermore National Lab., CA (United States)

    1993-09-01

    This document provides recommendations to guide reviewers in the application of Programmable Logic Controllers (PLCS) to the control, monitoring and protection of nuclear reactors. The first topics addressed are system-level design issues, specifically including safety. The document then discusses concerns about the PLC manufacturing organization and the protection system engineering organization. Supplementing this document are two appendices. Appendix A summarizes PLC characteristics. Specifically addressed are those characteristics that make the PLC more suitable for emergency shutdown systems than other electrical/electronic-based systems, as well as characteristics that improve reliability of a system. Also covered are PLC characteristics that may create an unsafe operating environment. Appendix B provides an overview of the use of programmable logic controllers in emergency shutdown systems. The intent is to familiarize the reader with the design, development, test, and maintenance phases of applying a PLC to an ESD system. Each phase is described in detail and information pertinent to the application of a PLC is pointed out.

  17. The Programmable Logic Controller and its application in nuclear reactor systems

    International Nuclear Information System (INIS)

    Palomar, J.; Wyman, R.

    1993-09-01

    This document provides recommendations to guide reviewers in the application of Programmable Logic Controllers (PLCS) to the control, monitoring and protection of nuclear reactors. The first topics addressed are system-level design issues, specifically including safety. The document then discusses concerns about the PLC manufacturing organization and the protection system engineering organization. Supplementing this document are two appendices. Appendix A summarizes PLC characteristics. Specifically addressed are those characteristics that make the PLC more suitable for emergency shutdown systems than other electrical/electronic-based systems, as well as characteristics that improve reliability of a system. Also covered are PLC characteristics that may create an unsafe operating environment. Appendix B provides an overview of the use of programmable logic controllers in emergency shutdown systems. The intent is to familiarize the reader with the design, development, test, and maintenance phases of applying a PLC to an ESD system. Each phase is described in detail and information pertinent to the application of a PLC is pointed out

  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. Account of requirements for modernization in VPBER-600 enhanced safety reactor instrumentation and control system development

    International Nuclear Information System (INIS)

    Shashkin, S.L.; Pobedonostsev, A.B.; Drumov, V.V.; Chudin, A.G.

    1993-01-01

    Nuclear power plant (NPP) with VPBER-600 reactor is a station of new generation. The specified term of reactor plant operation is 60 years and taking into account that the proposed term of starting the first power unit is on the turn of centuries one can definitely state that for Russia conditions VPBER-600 is a plant of 21 century. Such far removed term for NPP now in the stage of development as it can seem does not put the problems of modernization as first order tasks. But open-quotes...who does not think about future lives in the past.close quotes It is that the NPP instrumentation and control (I ampersand C) systems are in the most degree subjected to the influence of factors which favor their modifications. These factors can be arbitrarily divided into two groups: (1) inner factors, i.e. changes (failures, aging, etc) in I ampersand C components as well as changes dictated by technological reasons (change of equipment composition, control algorithms, operation modes); (2) outer factors, i.e. intensive development of information technologies and rapid improvement of electronic components. This presentation addresses the problem of modernization of the safety instrumentation for this next generation facility, and the research effort it will entail. The system is designed to allow for modernization, and the relatively easy adoption of new instrumentation and technology as it becomes available

  20. Comparative assessment of instrumentation and control (I and C) system architectures for research reactors

    International Nuclear Information System (INIS)

    Khalil, Rah Man; Heo, Gyun Young; Son, Han Seong; Kim, Young Ki; Park, Jae Kwan

    2012-01-01

    Application of digital I and C has increased in nuclear industry since last two decades but lack of experience, innovative and naive nature of technology and insufficient failure information raised questions on its use. The issues has been highlighted due to the use of digital I and C which were not relevant to analog. These are the potential weakness of digital systems for Common Cause Failure, threat to system security and reliability due to inter channel communication, need for highly integrated control room and difficulty to assess the digital I and C reliability. In the existing scenario, HANARO and JRTR have hybrid I and C systems (digital plus analog) whereas OPAL is fully digitalized. In order to authenticate the choice of fully digital I and C architecture for research reactor, it is required to perform assessment from risk point of view, cyber security as well other issues. The architecture assessment method and restrictions are discussed in the next part of article

  1. Comparative assessment of instrumentation and control (I and C) system architectures for research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Khalil, Rah Man; Heo, Gyun Young [Kyung Hee Univ., Seoul (Korea, Republic of); Son, Han Seong [Joongbu Univ., Chungnam (Korea, Republic of); Kim, Young Ki; Park, Jae Kwan [KAERI, Daejeon (Korea, Republic of)

    2012-10-15

    Application of digital I and C has increased in nuclear industry since last two decades but lack of experience, innovative and naive nature of technology and insufficient failure information raised questions on its use. The issues has been highlighted due to the use of digital I and C which were not relevant to analog. These are the potential weakness of digital systems for Common Cause Failure, threat to system security and reliability due to inter channel communication, need for highly integrated control room and difficulty to assess the digital I and C reliability. In the existing scenario, HANARO and JRTR have hybrid I and C systems (digital plus analog) whereas OPAL is fully digitalized. In order to authenticate the choice of fully digital I and C architecture for research reactor, it is required to perform assessment from risk point of view, cyber security as well other issues. The architecture assessment method and restrictions are discussed in the next part of article.

  2. Fast-acting nuclear reactor control device

    International Nuclear Information System (INIS)

    Kotlyar, O.M.; West, P.B.

    1993-01-01

    A fast-acting nuclear reactor control device is described for controlling a safety control rod within the core of a nuclear reactor, the reactor controlled by a reactor control system, the device comprising: a safety control rod drive shaft and an electromagnetic clutch co-axial with the drive shaft operatively connected to the safety control rod for driving and positioning the safety control rod within or without the reactor core during reactor operation, the safety rod being oriented in a substantially vertical position to allow the rod to fall into the reactor core under the influence of gravity during shutdown of the reactor; the safety control rod drive shaft further operatively connected to a hydraulic pump such that operation of the drive shaft simultaneously drives and positions the safety control rod and operates the hydraulic pump such that a hydraulic fluid is forced into an accumulator, filling the accumulator with oil for the storage and supply of primary potential energy for safety control rod insertion such that the release of potential energy in the accumulator causes hydraulic fluid to flow through the hydraulic pump, converting the hydraulic pump to a hydraulic motor having speed and power capable of full length insertion and high speed driving of the safety control rod into the reactor core; a solenoid valve interposed between the hydraulic pump and the accumulator, said solenoid valve being a normally open valve, actuated to close when the safety control rod is out of the reactor during reactor operation; and further wherein said solenoid opens in response to a signal from the reactor control system calling for shutdown of the reactor and rapid insertion of the safety control rod into the reactor core, such that the opening of the solenoid releases the potential energy in the accumulator to place the safety control rod in a safe shutdown position

  3. Reactor control device

    International Nuclear Information System (INIS)

    Kinoshita, Mitsuo.

    1991-01-01

    Heretofore, since the aimed value of a reactor power has been determined only based on the deviation between a temperature variation coefficient and the aimed value, it involves a problem that a region for finely determining a control constant is complicated. In view of the above, in the present invention, an approximate value of the aimed reactor power value is determined based on the aimed value for the temperature variation coefficient, then a compensation value for the aimed power value is determined based on the deviation between the temperature variation coefficient and the aimed value and, further, the aimed power value is determined based on the approximate value and the compensation value. Control elements are automatically operated so that the power follows the aimed value after determining the aimed value. Then, since the aimed reactor power value is controlled finely so that the responsiveness of the temperature variation coefficient is satisfactory and the temperature variation coefficient agrees with the aimed value, the stability for the control of temperature variation coefficient is satisfactory. That is, high performance control is enabled by a simple control algorithm to reduce the number of the steps for the design and the device control. (N.H.)

  4. Microcomputer-based equipment-control and data-acquisition system for fission-reactor reactivity-worth measurements

    International Nuclear Information System (INIS)

    McDowell, W.P.; Bucher, R.G.

    1980-01-01

    Material reactivity-worth measurements are one of the major classes of experiments conducted on the Zero Power research reactors (ZPR) at Argonne National Laboratory. These measurements require the monitoring of the position of a servo control element as a sample material is positioned at various locations in a critical reactor configuration. In order to guarantee operational reliability and increase experimental flexibility for these measurements, the obsolete hardware-based control unit has been replaced with a microcomputer based equipment control and data acquisition system. This system is based on an S-100 bus, dual floppy disk computer with custom built cards to interface with the experimental system. To measure reactivity worths, the system accurately positions samples in the reactor core and acquires data on the position of the servo control element. The data are then analyzed to determine statistical adequacy. The paper covers both the hardware and software aspects of the design

  5. Microcomputer-based equipment-control and data-acquisition system for fission-reactor reactivity-worth measurements

    Energy Technology Data Exchange (ETDEWEB)

    McDowell, W.P.; Bucher, R.G.

    1980-01-01

    Material reactivity-worth measurements are one of the major classes of experiments conducted on the Zero Power research reactors (ZPR) at Argonne National Laboratory. These measurements require the monitoring of the position of a servo control element as a sample material is positioned at various locations in a critical reactor configuration. In order to guarantee operational reliability and increase experimental flexibility for these measurements, the obsolete hardware-based control unit has been replaced with a microcomputer based equipment control and data acquisition system. This system is based on an S-100 bus, dual floppy disk computer with custom built cards to interface with the experimental system. To measure reactivity worths, the system accurately positions samples in the reactor core and acquires data on the position of the servo control element. The data are then analyzed to determine statistical adequacy. The paper covers both the hardware and software aspects of the design.

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  7. 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.)

  8. Home-made refurbishment of the instrumentation and control system of the TRIGA reactor of the University of Pavia

    International Nuclear Information System (INIS)

    Borio di Tigliole, A.; Cagnazzo, M.; Magrotti, G.; Manera, S.; Salvini, A.; Musitelli, G.; Nardo, R.

    2008-01-01

    The Instrumentation and Control (I and C) System of the TRIGA reactor of the University of Pavia was dated and, in order to grant a safe and continuous reactor operation for the future, it became necessary to substitute or to upgrade the system. Since the substitution of the I and C system with a new-made one was very difficult to be performed due to long authorization procedures, an home-made refurbishment was planned. Using commercial components of high quality, almost a complete substitution, channel-by-channel, of the I and C system was realized without changing the operating and safety logics. The system includes: - the Reactor Linear Power Channel and Chart Recorder; - the Reactor Percent Power Safety Channel; - the High Voltage and Low Voltage Power Supply; - the Automatic Reactor Power Control; - the Fuel Elements and Cooling-Water Temperatures Measuring Channels; - the Water Conductivity Measuring Channel. The refurbished I and C system shows a very good operational behavior and reliability and will assure a continuous operation of the reactor for the future

  9. RETRAN code analysis of Tsuruga-2 plant chemical volume control system (CVCS) reactor coolant leakage incident

    International Nuclear Information System (INIS)

    Kawai, Hiroshi

    2002-01-01

    In the Chemical Volume Control System (CVCS) reactor primary coolant leakage incident, which occurred in Tsuruga-2 (4-loop PWR, 3,423 MWt, 1,160 MWe) on July 12, 1999, it took about 14 hours before the leakage isolation. The delayed leakage isolation and a large amount of leakage have become a social concern. Effective procedure modification was studied. Three betterments were proposed based on a qualitative analysis to reduce the pressure and temperature of the primary loop as fast as possible by the current plant facilities while maintaining enough subcooling of the primary loop. I analyzed the incident with RETRAN code in order to quantitatively evaluate the leakage reduction when these betterments are adopted. This paper is very new because it created a typical analysis method for PWR plant behavior during plant shutdown procedure which conventional RETRAN transient analyses rarely dealt with. Also the event time is very long. To carry out this analysis successfully, I devised new models such as an Residual Heat Removal System (RHR) model etc. and simplified parts of the conventional model. Based on the analysis results, I confirmed that leakage can be reduced by about 30% by adopting these betterments. Then the Japan Atomic Power Company (JAPC) modified the operational procedure for reactor primary coolant leakage events adopting these betterments. (author)

  10. Coordination of motor systems of the control bars of TRIGA Mark III reactor, through the use of a PLC

    International Nuclear Information System (INIS)

    Castro R, P. R.

    2016-01-01

    The use of programmable logic controllers (PLCs) has a wide field in the realization of automatic systems, since is sought that the form of control is easy for any user. In this work, the TRIGA Mark III reactor of Instituto Nacional de Investigaciones Nucleares (ININ) is intended to operate with a programming code in PLC for the automation of the control rods, having uniform wear according to the power required by the user. On the other hand, is proposed to develop an HMI graphical interface for communication via Ethernet, allowing supervision during the reactor operation process and greater protection of operators during reactor startup. The accuracy of the new actuators, as well as their durability, will allow a good performance of the reactor for many years to come. (Author)

  11. TRIGA reactor main systems

    International Nuclear Information System (INIS)

    Boeck, H.; Villa, M.

    2007-01-01

    This module describes the main systems of low power (<2 MW) and higher power (≥2 MW) TRIGA reactors. The most significant difference between the two is that forced reactor cooling and an emergency core cooling system are generally required for the higher power TRIGA reactors. However, those TRIGA reactors that are designed to be operated above 3 MW also use a TRIGA fuel that is specifically designed for those higher power outputs (3 to 14 MW). Typical values are given for the respective systems although each TRIGA facility will have unique characteristics that may only be determined by the experienced facility operators. Due to the inherent wide scope of these research reactor facilities construction and missions, this training module covers those systems found at most operating TRIGA reactor facilities but may also discuss non-standard equipment that was found to be operationally useful although not necessarily required. (author)

  12. Design of a decoupled AP1000 reactor core control system using digital proportional–integral–derivative (PID) control based on a quasi-diagonal recurrent neural network (QDRNN)

    International Nuclear Information System (INIS)

    Wei, Xinyu; Wang, Pengfei; Zhao, Fuyu

    2016-01-01

    Highlights: • We establish a disperse dynamic model for AP1000 reactor core. • A digital PID control based on QDRNN is used to design a decoupling control system. • The decoupling performance is verified and discussed. • The decoupling control system is simulated under the load following operation. - Abstract: The control system of the AP1000 reactor core uses the mechanical shim (MSHIM) strategy, which includes a power control subsystem and an axial power distribution control subsystem. To address the strong coupling between the two subsystems, an interlock between the two subsystems is used, which can only alleviate but not eliminate the coupling. Therefore, sometimes the axial offset (AO) cannot be controlled tightly, and the flexibility of load-following operation is limited. Thus, the decoupling of the original AP1000 reactor core control system is the focus of this paper. First, a two-node disperse dynamic model is established for the AP1000 reactor core to use PID control. Then, a digital PID control system based on a quasi-diagonal recurrent neural network (QDRNN) is designed to decouple the original system. Finally, the decoupling of the control system is verified by the step signal and load-following condition. The results show that the designed control system can decouple the original system as expected and the AO can be controlled much more tightly. Moreover, the flexibility of the load following is increased.

  13. Design of a decoupled AP1000 reactor core control system using digital proportional–integral–derivative (PID) control based on a quasi-diagonal recurrent neural network (QDRNN)

    Energy Technology Data Exchange (ETDEWEB)

    Wei, Xinyu, E-mail: xyuwei@mail.xjtu.edu.cn; Wang, Pengfei, E-mail: pengfeixiaoli@yahoo.cn; Zhao, Fuyu, E-mail: fuyuzhao_xj@163.com

    2016-08-01

    Highlights: • We establish a disperse dynamic model for AP1000 reactor core. • A digital PID control based on QDRNN is used to design a decoupling control system. • The decoupling performance is verified and discussed. • The decoupling control system is simulated under the load following operation. - Abstract: The control system of the AP1000 reactor core uses the mechanical shim (MSHIM) strategy, which includes a power control subsystem and an axial power distribution control subsystem. To address the strong coupling between the two subsystems, an interlock between the two subsystems is used, which can only alleviate but not eliminate the coupling. Therefore, sometimes the axial offset (AO) cannot be controlled tightly, and the flexibility of load-following operation is limited. Thus, the decoupling of the original AP1000 reactor core control system is the focus of this paper. First, a two-node disperse dynamic model is established for the AP1000 reactor core to use PID control. Then, a digital PID control system based on a quasi-diagonal recurrent neural network (QDRNN) is designed to decouple the original system. Finally, the decoupling of the control system is verified by the step signal and load-following condition. The results show that the designed control system can decouple the original system as expected and the AO can be controlled much more tightly. Moreover, the flexibility of the load following is increased.

  14. Research and development on the application of advanced control technologies to advanced nuclear reactor systems: A US national perspective

    International Nuclear Information System (INIS)

    White, J.D.; Monson, L.R.; Carrol, D.G.; Dayal, Y.

    1989-01-01

    Control system designs for nuclear power plants are becoming more advanced through the use of digital technology and automation. This evolution is taking place because of: (1) the limitations in analog based control system performance and maintenance and availability and (2) the promise of significant improvement in plant operation and availability due to advances in digital and other control technologies. Digital retrofits of control systems in US nuclear plants are occurring now. Designs of control and protection systems for advanced LWRs are based on digital technology. The use of small inexpensive, fast, large-capacity computers in these designs is the first step of an evolutionary process described in this paper. Under the sponsorship of the US Department of Energy (DOE), Oak Ridge National Laboratory, Argonne National Laboratory, GE Nuclear Energy and several universities are performing research and development in the application of advances in control theory, software engineering, advanced computer architectures, artificial intelligence, and man-machine interface analysis to control system design. The target plant concept for the work described in this paper is the Power Reactor Inherently Safe Module reactor (PRISM), an advanced modular liquid metal reactor concept. This and other reactor designs which provide strong passive responses to operational upsets or accidents afford good opportunities to apply these advances in control technology. 18 refs., 5 figs

  15. Description of the control and safety systems of the RA reactor; Opis sistema za upravljanje i sigurnosnu zastitu RA

    Energy Technology Data Exchange (ETDEWEB)

    Popovic, B; Pesic, M [Institute of Nuclear Sciences Boris Kidric, Reaktor RA, Vinca, Beograd (Yugoslavia)

    1962-07-01

    This report contains detailed description and scheme of the control and safety system of the RA reactor. It consists of interconnected five systems: for automated regulation; compensation rods; safety rods; power density measurement device; period meter; automated D{sub 2}O level meter in the core. Automated regulation system is divided into two parts: basic system for reactor operation regime at power from 10kW - 10 MW and precise regulation system for operation at set-up power level up to 10 kW which is used occasionally.

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

    International Nuclear Information System (INIS)

    Gunther, W.; Sullivan, K.

    1991-01-01

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

  17. Design and implementation of the control system for the new console of TRIGA-3-Salazar Reactor

    International Nuclear Information System (INIS)

    Gonzalez M, J.L.

    1994-01-01

    TRIGA-3-Salazar Reactor was set in operation in 1968 and the aging of its components has cause the increasing in the maintenance. In the presence of this, it becomes necessary to replace the reactor console using new technologies, considering the incorporation of a personal computer. The aim of this work is the design and construction of the equipment interfaces as well as the digital computer program for the automation and control of the TRIGA-3-Salazar Reactor by means of a personal computer. (Author)

  18. reactor power control using fuzzy logic

    International Nuclear Information System (INIS)

    Ahmed, A.E.E.

    2001-01-01

    power stabilization is a critical issue in nuclear reactors. convention pd- controller is currently used in egypt second testing research reactor (ETRR-2). two fuzzy controllers are proposed to control the reactor power of ETRR-2 reactor. the design of the first one is based on a set of linguistic rules that were adopted from the human operators experience. after off-line fuzzy computations, the controller is a lookup table, and thus, real time controller is achieved. comparing this f lc response with the pd-controller response, which already exists in the system, through studying the expected transients during the normal operation of ETRR-2 reactor, the simulation results show that, fl s has the better response, the second controller is adaptive fuzzy controller, which is proposed to deal with system non-linearity . The simulation results show that the proposed adaptive fuzzy controller gives a better integral square error (i se) index than the existing conventional od controller

  19. Neutron behavior, reactor control, and reactor heat transfer. Volume four

    International Nuclear Information System (INIS)

    Anon.

    1986-01-01

    Volume four covers neutron behavior (neutron absorption, how big are nuclei, neutron slowing down, neutron losses, the self-sustaining reactor), reactor control (what is controlled in a reactor, controlling neutron population, is it easy to control a reactor, range of reactor control, what happens when the fuel burns up, controlling a PWR, controlling a BWR, inherent safety of reactors), and reactor heat transfer (heat generation in a nuclear reactor, how is heat removed from a reactor core, heat transfer rate, heat transfer properties of the reactor coolant)

  20. Water chemical control of the TRIGA IPR-R1 reactor primary cooling system

    International Nuclear Information System (INIS)

    Auler, Lucia M.L.A; Chaves, Renata D.A.; Palmieri, Helena E.L.; Menezes, Maria Angela de B.C.; Oliveira, Paulo F.; Kastner, Geraldo F.; Damazio, Ilza; Fagundes, Oliene dos R.; Cintra, Maria Olivia C.; Andrade, Geraldo V. de; Amaral, Angela M.; Franco, Milton B.; Fortes, Flavio; Gomes, Nilton Carlos; Vidal, Andrea; Maretti Junior, Fausto; Knupp, Eliana A.N.; Souza, Wagner de; Guedes, Joao B.; Furtado, Renato C.S.

    2013-01-01

    The TRIGA Mark I IPR-R1 reactor located at CDTN/CNEN has been in operation and contributed to research and with services to society since 1960. Is has been used in several activities such as nuclear power plant operation, graduate and post-graduate training courses, isotope production, and as an analytical irradiation tool of different types of samples. Among the several structural and operational safety requirements is the chemical quality control of the primary circuit cooling water. The aim of this work was to check the cooling water quality from the pool reactor. A water sampling plan was proposed (May, 2011 - June, 2012) and presents the results obtained in this period. The natural radioactivity level as gross alpha and gross beta activity and other chemical parameters (pH and electric conductivity) of the samples were analyzed. Some instrumental techniques were used: potentiometric methods (pH), conductometric methods (electrical conductivity, EC) and gross α and gross β proportional counting system). (author)

  1. A study on implementation of dynamic safety system in programmable logic controller for pressurized water reactor

    International Nuclear Information System (INIS)

    Kim, Ung Soo

    1997-02-01

    The dynamic safety system (DSS) is a computer based reactor protection system that has dynamic self-testing feature and fail-safe nature inherently. The inherent dynamic self-testing feature and fail-safe design provide a high level of reliability and low spurious trip rate. We can also reduce the time and human efforts to maintain the system by virtue of those features. Therefore, the application of the DSS to PWR has many advantages. The DSS has been applied only to advanced gas-cooled reactor (AGR) in the UK. In order to apply the DSS for PWR, the DSS has to be modified because there exist many differences between PWR and AGR for which the DSS was tested and installed. These differences are trip algorithms, monitored parameters, trip logics, and other conditions. In this study, the DSS algorithm is modified for PWR first. The modified DSS has several new features : 1) The modified DSS tests and processes time-dependent parameters, while the original DSS does not. 2) It has flexibility for handling several types of voting logic but the original DSS handles the only one type of voting - 2 out of 4 coincidence logic. Then, in this study, the modified DSS is implemented in programmable logic controller (PLC) using the ladder logic. Finally, the modified DSS is tested in two ways in this work : 1) The manual test is performed using direct input through the human computer interface (HCI) system. 2) The scenario based test is performed using input from the FISA-2/WS simulator. From the test results, it is shown that the modified DSS operates correctly in all conditions

  2. Reactor water level control device

    International Nuclear Information System (INIS)

    Utagawa, Kazuyuki.

    1993-01-01

    A device of the present invention can effectively control fluctuation of a reactor water level upon power change by reactor core flow rate control operation. That is, (1) a feedback control section calculates a feedwater flow rate control amount based on a deviation between a set value of a reactor water level and a reactor water level signal. (2) a feed forward control section forecasts steam flow rate change based on a reactor core flow rate signal or a signal determining the reactor core flow rate, to calculate a feedwater flow rate control amount which off sets the steam flow rate change. Then, the sum of the output signal from the process (1) and the output signal from the process (2) is determined as a final feedwater flow rate control signal. With such procedures, it is possible to forecast the steam flow rate change accompanying the reactor core flow rate control operation, thereby enabling to conduct preceding feedwater flow rate control operation which off sets the reactor water level fluctuation based on the steam flow rate change. Further, a reactor water level deviated from the forecast can be controlled by feedback control. Accordingly, reactor water level fluctuation upon power exchange due to the reactor core flow rate control operation can rapidly be suppressed. (I.S.)

  3. Development of monitoring, control and protection systems of nuclear power reactors in the USSR: Status and trends

    International Nuclear Information System (INIS)

    Kondrat'ev, V.V.; Mikhailov, M.N.; Prozorov, V.K.; Chudin, A.G.

    1992-01-01

    In 1989-90 the Supervision Body approved the new safety regulations for nuclear power plants. The operating plants do not always completely satisfy them and a great amount of work to develop operating power units up to the conditions required is necessary. This paper describes briefly the main changes made in monitoring, control and protection systems of nuclear power reactors to increase the reactor safety. The following fields are presented in the paper: General status of the NPP control and safety systems and instrumentation in USSR; sensors; electronic equipment; actuators improvement; qualification tests. (author). 3 figs

  4. Pressurized water reactor systems

    International Nuclear Information System (INIS)

    Meyer, P.J.

    1975-01-01

    Design and mode of operation of the main PWR components are described: reactor core, pressure vessel and internals, cooling systems with pumps and steam generators, ancillary systems, and waste processing. (TK) [de

  5. Simplified Models for Analysis and Design of the Control System Main Loops of CAREM Reactor

    International Nuclear Information System (INIS)

    Etchepareborda, Andres; Flury, Celso

    2000-01-01

    The target of this work is to show a few models developed for control analysis and design of the reactor CAREM's main control loops within a broad range of power (between 40 % and 100%).By one side, it is shown the main features of a analytic model programed in MATLAB.This model is based on fitting steady state points at different power levels of the CAREM's RETRAN model.By the other side, it is shown linear models of black-box type denoting the perturbed behavior of the system for each level power point.These models are identified from temporal responses of CAREM's RETRAN model to perturbed input signals over the different steady power level points.Then the dynamics of these models are verified contrasting the temporal responses of the RETRAN model versus the responses of the MATLAB model and the identified models, in each steady power level point.Also are contrasting the frequency response of the linearization of MATLAB model versus the frequency response of the identified models, in each steady power level point.Either the MATLAB model as the identified models are good enough for the control analysis and design of the three main control loops.The MATLAB model has a few differences against the RETRAN model in the primary pressure output variable, that it must be taken into account in the design of this control loop if this model is used.The aim of these models is to represent in a satisfactory way the dynamics of the plant for a later control analysis and design of the control loops in a frequency range between 0.01 rad/seg and 0.3 rad/seg, and a power range between 40 % and 100 %

  6. New digital control system for the operation of the Colombian research reactor IAN-R1; Nuevo sistema de control digital para la operacion del reactor de investigacion Colombiano IAN-R1

    Energy Technology Data Exchange (ETDEWEB)

    Celis del A, L.; Rivero, T.; Bucio, F.; Ramirez, R.; Segovia, A.; Palacios, J., E-mail: lina.celis@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2015-09-15

    En 2011, Mexico won the Colombian international tender for the renewal of instrumentation and control of the IAN-R1 Reactor, to Argentina and the United States. This paper presents the design criteria and the development made for the new digital control system installed in the Colombian nuclear reactor IAN-R1, which is based on a redundant and diverse architecture, which provides increased availability, reliability and safety in the reactor operation. This control system and associated instrumentation met all national export requirements, with the safety requirements established by the IAEA as well as the requirements demanded by the Colombian Regulatory Body in nuclear matter. On August 20, 2012, the Colombian IAN-R1 reactor reached its first criticality controlled with the new system developed at Instituto Nacional de Investigaciones Nucleares (ININ). On September 14, 2012, the new control system of the Colombian IAN-R1 reactor was officially handed over to the Colombian authorities, this being the first time that Mexico exported nuclear technology through the ININ. Currently the reactor is operating successfully with the new control system, and has an operating license for 5 years. (Author)

  7. Design, assembly, and initial use of a digital system for the closed-loop control of a nuclear research reactor

    International Nuclear Information System (INIS)

    Kwok, K.S.; Bernard, J.A.; Lanning, D.D.

    1991-01-01

    In this paper the design, implementation, and initial testing of a multiple-computer/single-task system for the closed-loop control of a nuclear research reactor is described. A major advantage of the multiple-computer approach is that generic safety-related software that remains invariant can be separated from the control law software that is updated as plant procedures change. This facilitates software validation. Also, this approach allows both real-time operation and high numerical throughput. System compatibility was achieved through design of a special passive back plane which enabled the otherwise incompatible components to be operated in an integrated system. This multiple-computer system, which was designated as the Advanced Control Computer System (ACCS), has been installed on the 5-MWt MIT Research Reactor. In addition to a description of both the system and its associated hardware and software interfaces, experimental results are presented from its initial trials

  8. Computerized reactor monitor and control for nuclear reactors

    International Nuclear Information System (INIS)

    Buerger, L.

    1982-01-01

    The analysis of a computerized process control system developed by Transelektro-KFKI-Videoton (Hangary) for a twenty-year-old research reactor in Budapest and or a new one in Tajura (Libya) is given. The paper describes the computer hardware (R-10) and the implemented software (PROCESS-24K) as well as their applications at nuclear reactors. The computer program provides for man-machine communication, data acquisition and processing, trend and alarm analysis, the control of the reactor power, reactor physical calculations and additional operational functions. The reliability and the possible further development of the computerized systems which are suitable for application at reactors of different design are also discussed. (Sz.J.)

  9. A control system for industrial plant (e.g. a pressurized water reactor)

    International Nuclear Information System (INIS)

    Spiller, C.R.L.

    1990-01-01

    A control system for an industrial plant eg. a pressurised water nuclear reactor, comprises a plurality of instrument sets and a plurality of logic sets. The instrument sets have a number of sensors which detect parameters (temperature, pressure vibration) of the industrial plant, and have two serial link controllers which supply the output signals from each sensor in the instrument set sequentially to the logic sets via conductors. The logic sets have a number of auto select logic circuits, each of which selects data from the sensors from one of the instrument sets, and a synchroniser ensures that the output signals from the sensors detecting the same parameter are supplied to a voting logic circuit at the same time. The voting logic circuit performs a voting function on the output signals to produce a series of high reliability signals which are converted to parallel high reliability signals by a series to a parallel converter. The high reliability signals are supplied to a fault logic shutdown circuit which controls the operation of shutdown mechanisms for the industrial plant. (author)

  10. The micro-processor controlled process radiation monitoring system for reactor safety systems

    International Nuclear Information System (INIS)

    Mizuno, K.; Noguchi, A.; Kumagami, S.; Gotoh, Y.; Kumahara, T.; Arita, S.

    1986-01-01

    Digital computers are soon expected to be applied to various real-time safety and safety-related systems in nuclear power plants. Hitachi is now engaged in the development of a micro-processor controlled process radiation monitoring system, which operates on digital processing methods employed with a log ratemeter. A newly defined methodology of design and test procedures is being applied as a means of software program verification for these safety systems. Recently implemented micro-processor technology will help to achieve an advanced man-machine interface and highly reliable performance. (author)

  11. Nuclear reactor control assembly

    International Nuclear Information System (INIS)

    Negron, S.B.

    1991-01-01

    This patent describes an assembly for providing global power control in a nuclear reactor having the core split into two halves. It comprises a disk assembly formed from at least two disks each machined with an identical surface hole pattern such that rotation of one disk relative to the other causes the hole pattern to open or close, the disk assembly being positioned substantially at the longitudinal center of and coaxial with the core halves; and means for rotating at least one of the disks relative to the other

  12. Simulation-based expert system for nuclear reactor control and diagnostics. Progress report

    International Nuclear Information System (INIS)

    Lee, J.C.; Martin, W.R.

    1986-01-01

    This research concerns the development of artificial intelligence (AI) techniques suitable for application to the diagnostics and control of nuclear power plant systems. The overall objective of the current effort is to build a prototype simulation-based expert system for diagnosing accidents in nuclear reactors. The system is being designed to analyze plant data heuristically using fuzzy logic to form a set of hypotheses about a particular transient. Hypothesis testing, fault magnitude estimation and transient analysis is performed using simulation programs to model plant behavior. An adaptive learning technique has been developed for achieving accurate simulations of plant dynamics using low-order physical models of plant components. The results of the diagnostics and simulation analysis of the plant transient are to be analyzed by an expert system for final diagnoses and control guidance. To date, significant progress has been made toward achieving the primary goals of this project. Based on a critical safety functions approach, an overall design for the nuclear plant expert system has been developed. The methodology for performing diagnostic reasoning on plant signals has been developed and the algorithms implemented and tested. A methodology for utilizing the information contained in the physical models of plant components has also been developed. This work included the derivation of a unique Kalman filtering algorithm for using power plant data to systematically improve on-line simulations through the judicious adjustment of key model parameters. A few simulation models of key plant components have been developed and implemented to demonstrate the method on a realistic accident scenario. The chosen transient is a loss of feed flow exasperated by a stuck open relief valve, similar to the initiating event of the Three Mile Island Unit 2 accident in 1979

  13. Reactor Safety Assessment System

    International Nuclear Information System (INIS)

    Sebo, D.E.; Bray, M.A.; King, M.A.

    1987-01-01

    The Reactor Safety Assessment System (RSAS) is an expert system under development for the United States Nuclear Regulatory Commission (USNRC). RSAS is designed for use at the USNRC Operations Center in the event of a serious incident at a licensed nuclear power plant. RSAS is a situation assessment expert system which uses plant parametric data to generate conclusions for use by the NRC Reactor Safety Team. RSAS uses multiple rule bases and plant specific setpoint files to be applicable to all licensed nuclear power plants in the United States. RSAS currently covers several generic reactor categories and multiple plants within each category

  14. Reactor safety assessment system

    International Nuclear Information System (INIS)

    Sebo, D.E.; Bray, M.A.; King, M.A.

    1987-01-01

    The Reactor Safety Assessment System (RSAS) is an expert system under development for the United States Nuclear Regulatory Commission (USNRC). RSA is designed for use at the USNRC Operations Center in the event of a serious incident at a licensed nuclear power plant. RSAS is a situation assessment expert system which uses plant parametric data to generate conclusions for use by the NRC Reactor Safety Team. RSAS uses multiple rule bases and plant specific setpoint files to be applicable to all licensed nuclear power plants in the United States. RSAS currently covers several generic reactor categories and multiple plants within each category

  15. Light Water Reactor Sustainability Program Advanced Instrumentation, Information, and Control Systems Technologies Technical Program Plan for 2013

    Energy Technology Data Exchange (ETDEWEB)

    Hallbert, Bruce [Idaho National Laboratory (INL), Idaho Falls, ID (United States); Thomas, Ken [Idaho National Laboratory (INL), Idaho Falls, ID (United States)

    2014-09-01

    Reliable instrumentation, information, and control (II&C) systems technologies are essential to ensuring safe and efficient operation of the U.S. light water reactor (LWR) fleet. These technologies affect every aspect of nuclear power plant (NPP) and balance-of-plant operations. In 1997, the National Research Council conducted a study concerning the challenges involved in modernization of digital instrumentation and control systems in NPPs. Their findings identified the need for new II&C technology integration.

  16. Light Water Reactor Sustainability Program Advanced Instrumentation, Information, and Control Systems Technologies Technical Program Plan for FY 2016

    Energy Technology Data Exchange (ETDEWEB)

    Hallbert, Bruce Perry [Idaho National Lab. (INL), Idaho Falls, ID (United States); Thomas, Kenneth David [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-10-01

    Reliable instrumentation, information, and control (II&C) systems technologies are essential to ensuring safe and efficient operation of the U.S. light water reactor (LWR) fleet. These technologies affect every aspect of nuclear power plant (NPP) and balance-of-plant operations. In 1997, the National Research Council conducted a study concerning the challenges involved in modernization of digital instrumentation and control systems in NPPs. Their findings identified the need for new II&C technology integration.

  17. Method for calculating the forces and deformations in the fast reactor fuel assembly accounting for the effects of reactor control system elements and shutdown

    International Nuclear Information System (INIS)

    Likhachev, Yu.I.; Vashlyaev, Yu.N.; Kravchenko, I.N.

    1980-01-01

    Methods for calculating deformations and interaction forces of heat-generating assemblies (HGA) of fast reactor core with account for the effect of control and protection system (CPS) elements at the reactor operation and change of interaction efforts between HGA at the reactor shutdown, are described. The results of testing the suggested methods on example of estimate of HGA behaviour of the BN-350 reactor are presented. For estimating the effect of CPS elements on HGA bending the sector model has been used. It is assumed that HGA deformation inside each sector is independent of HGA deformation of other sectors. A higher calculation accuracy is attained by means of laying out of sectors into regions of preferable influence of emergency protection elements and compensating packets. When determining deformation and interaction efforts between HGA caused by temperature change in the course of shutdown it is supposed that the HGA deformation is purely elastic. The methods described are realized in the form of ABRI-CPS and ABRI-HOL programs written in FORTRAN for the BESM-6 computer. The results of HGA calculations of the BN-350 reactor core show that CPS elements decrease contact efforts in the middle of the central packet, increase contact efforts in the peak of the central packet, increase contact efforts in the peaks of packets from the eight row to the periphery and increase contact efforts in the middles of packets from the 5th to 9th row [ru

  18. Reactor protection systems for the Replacement Research Reactor, ANSTO

    International Nuclear Information System (INIS)

    Morris, C.R.

    2003-01-01

    The 20-MW Replacement Research Reactor Project which is currently under construction at ANSTO will have a combination of a state of the art triplicated computer based reactor protection system, and a fully independent, and diverse, triplicated analogue reactor protection system, that has been in use in the nuclear industry, for many decades. The First Reactor Protection System (FRPS) consists of a Triconex triplicated modular redundant system that has recently been approved by the USNRC for use in the USA?s power reactor program. The Second Reactor Protection System is a hardwired analogue system supplied by Foxboro, the Spec 200 system, which is also Class1E qualified. The FRPS is used to drop the control rods when its safety parameter setpoints have been reached. The SRPS is used to drain the reflector tank and since this operation would result in a reactor poison out due to the time it would take to refill the tank the FRPS trip setpoints are more limiting. The FRPS and SRPS have limited hardwired indications on the control panels in the main control room (MCR) and emergency control centre (ECC), however all FRPS and SRPS parameters are capable of being displayed on the reactor control and monitoring system (RCMS) video display units. The RCMS is a Foxboro Series I/A control system which is used for plant control and monitoring and as a protection system for the cold neutron source. This paper will provide technical information on both systems, their trip logics, their interconnections with each other, and their integration into the reactor control and monitoring system and control panels. (author)

  19. Simulation study of multi-step model algorithmic control of the nuclear reactor thermal power tracking system

    International Nuclear Information System (INIS)

    Shi Xiaoping; Xu Tianshu

    2001-01-01

    The classical control method is usually hard to ensure the thermal power tracking accuracy, because the nuclear reactor system is a complex nonlinear system with uncertain parameters and disturbances. A sort of non-parameter model is constructed with the open-loop impulse response of the system. Furthermore, a sort of thermal power tracking digital control law is presented using the multi-step model algorithmic control principle. The control method presented had good tracking performance and robustness. It can work despite the existence of unmeasurable disturbances. The simulation experiment testifies the correctness and effectiveness of the method. The high accuracy matching between the thermal power and the referenced load is achieved

  20. Reactor system on barge

    International Nuclear Information System (INIS)

    Hayashi, Kingo; Yamada, Nobuyuki

    1987-01-01

    Floating electrical power plants or power plant barges add new dimensions to utility planners and agencies in the world. Intrinsically safe and economical reactors (ISER) employ steel reactor pressure vessels, which significantly reduce the weight as compared with PIUS, and provide siting versatility including barge-mounted plants. In this paper, the outline of power plant barges and barge-mounted ISERs is described. Besides their mobility, power plant barges have the salient advantages such as short delivery time and better quality control due to the outfitting in shipyards. These power plant barges may be temporarily moored or permanently grounded in shallow water at the centers of industrial complexes or the suitable areas adjacent to them, and satisfy the increasing needs for electric power. A cost-effective and technically perfect barge positioning system should be designed to meet the specific requirement for the location and its condition. Offshore siting away from coast may be applicable only to large plants of 1,000 MWe or more, and inshore siting and coastal or river siting are considered for an ISER-200 barge-mounted plant. The system of a barge-mounted ISER plant is discussed in the case of a floating type and the type on a seismic base isolator. (Kako, I.)

  1. Pneumatic transport systems for TRIGA reactors

    International Nuclear Information System (INIS)

    Bolton, John A.

    1970-01-01

    Main parameters and advantages of pneumatically operated systems, primarily those operated by gas pressure are discussed. The special irradiation ends for the TRIGA reactor are described. To give some idea of the complexity of some modern systems, the author presents the large system currently operating at the National Bureau of Standards in Washington. In this system, 13 stations are located throughout the radiochemistry laboratories and three irradiation ends are located in the reactor, which is a 14-megawatt unit. The system incorporates practically every fail-safe device possible, including ball valves located on all capsule lines entering the reactor area, designed to close automatically in the event of a reactor scram, and at that time capsules within the reactor would be diverted by means of switches located on the inside of the reactor wall. The whole system is under final control of a permission control panel located in the reactor control room. Many other safety accessories of the system are described

  2. Computerized supervision and control system for movement at the RP-10 reactor control rods bank; Sistema de supervicion y mando computarizado para el movimiento en banco de las barras de control del Reactor RP-10

    Energy Technology Data Exchange (ETDEWEB)

    Padilla M, C E

    1998-07-01

    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)

  3. Life extension activities and modernization strategies for instrumentation ampersand control systems of research and power reactors in India

    International Nuclear Information System (INIS)

    Chaganty, S.P.; Bairi, B.R.

    1993-01-01

    Based on three and half decades of experience gained in the operation and maintenance of Instrumentation and Control Systems of nuclear reactors in India, specific investigations were made to understand various aspects of aging. The analysis of the failure rates of various instruments, plant outage figures and obsolescence of components have necessitated the replacement of instrumentation to improve the reliability and performance. The aging models available were used to determine the extent of performance degradation and to formulate maintenance strategies. The nuclear instrumentation of the aging research reactors at Bhabha Atomic Research Centre (BARC) has been replaced with high reliability equipment using modern integrated circuits. This has resulted in an improvement in the mean time between failure (MTBF) by a factor of five. The neutronic instrumentation of Fast Breeder Test Reactor (FBTR) at Madras is currently being upgraded with the introduction of microprocessor based safety units for reactivity computation and online testing of safety logic with Fine Impulse Technique. The operating experience has also indicated the necessity of developing online surveillance methods and status monitoring of various systems to detect aging. Online cable insulation measurement technique and noise analysis methods for vibration monitoring have been developed. Campbell method of signal processing has been successfully used in extending the useful life of Local Power Range monitors in the Boiling Water Reactor at Tarapur. In order to improve reliability, accuracy and provide efficient man machine interface, microprocessor based systems with online testing features have been installed in power reactors. These include the high performance reactor regulating system and centralised radiation monitoring systems commissioned at Kakrapara power station. The paper describes the above systems and the modernization strategies for nuclear instrumentation and control

  4. Control for nuclear reactor

    International Nuclear Information System (INIS)

    Ash, E.B.; Bernath, L.; Facha, J.V.

    1980-01-01

    A nuclear reactor is provided with several hydraulically-supported spherical bodies having a high neutron absorption cross section, which fall by gravity into the core region of the reactor when the flow of supporting fluid is shut off. (auth)

  5. RETRAN code analysis of Tsuruga-2 plant chemical volume control system (CVCS) reactor coolant leakage incident

    International Nuclear Information System (INIS)

    Kawai, H.

    2001-01-01

    JAPC purchased RETRAN, a program for transient thermal hydraulic analysis of complex fluid flow system, from the U.S. Electric Power Research Institute in 1992. Since then, JAPC has been utilizing RETRAN to evaluate safety margins of actual plant operation, in coping with troubles (investigating trouble causes and establishing countermeasures), and supporting reactor operation (reviewing operational procedures etc.). In this paper, a result of plant analysis performed on a CVCS reactor primary coolant leakage incident which occurred at JAPC's Tsuruga-2 plant (4-loop PWR, 3423 MWt, 1160 MW) on July 12 of 1999 and, based on the result, we made a plan to modify our operational procedure for reactor primary coolant leakage events in order to make earlier plant shutdown and this reduced primary coolant leakage. (author)

  6. Complete automation of nuclear reactors control

    International Nuclear Information System (INIS)

    Weill, J.

    1955-01-01

    The use of nuclear reactor for energy production induces the installation of automatic control systems which need to be safe enough and can adapt to the industrial scale of energy production. These automatic control systems have to insure the constancy of power level and adjust the power produced to the energy demand. Two functioning modes are considered: nuclear plant connected up to other electric production systems as hydraulic or thermic plants or nuclear plants functioning on an independent network. For nuclear plants connected up with other production plants, xenon poisoning and operating cost lead to keep working at maximum power the nuclear reactors. Thus, the power modulation control system will not be considered and only start-up control, safety control, and control systems will be automated. For nuclear power plants working on an independent network, the power modulation control system is needed to economize fuel. It described the automated control system for reactors functioning with constant power: a power measurement system constituted of an ionization chamber and a direct-current amplifier will control the steadfastness of the power produced. For reactors functioning with variable power, the automated power control system will allow to change the power and maintain it steady with all the necessary safety and will control that working conditions under P max and R max (maximum power and maximum reactivity). The effects of temperature and xenon poisoning will also be discussed. Safety systems will be added to stop completely the functioning of the reactor if P max is reached. (M.P.)

  7. Study on computer-aided control system design platform of 10MW high temperature gas-cooled test reactor

    International Nuclear Information System (INIS)

    Feng Yan; Shi Lei; Sun Yuliang; Luo Shaojie

    2004-01-01

    the 10 MW high temperature gas-cooled test reactor (HTR-10) is the first modular pebble bed reactor built in China, which needs to be researched on engineering design, control study, safety analysis and operator training. An integrated system for simulation, control design and online assistance of the HTR-10 (HTRSIMU) has been developed by the Institute of Nuclear Energy Technology (INET) of Tsinghua University. The HTRSIMU system is based on a high-speed local area network, on which a computer-aided control system design platform (CDP) is developed and combined with the simulating subsystem in order to provide a visualized and convenient tool for the HTR-10 control system design. The CDP has friendly man-machine interface and good expansibility, in which eighteen types of control items are integrated. These control items are divided into two types: linear and non-linear control items. The linear control items include Proportion, Integral, Differential, Inertial, Leed-lag, Oscillation, Pure-lag, Common, PID and Fuzzy, while the non-linear control items include Saturation, Subsection, Insensitive, Backlash, Relay, Insensi-Relay, Sluggish-Relay and Insens-Slug. The CDP provides a visualized platform for control system modeling and the control loop system can be automatically generated and graphically simulated. Users can conveniently design control loop, modify control parameters, study control method, and analyze control results just by clicking mouse buttons. This kind of control system design method can provide a powerful tool and good reference for the actual system operation for HTR-10. A control scheme is also given and studied to demonstrate the functions of the CDP in this article. (author)

  8. Practice in development and utilization of program-technical complex (PTK) in in-reactor control systems

    International Nuclear Information System (INIS)

    Gribov, A.A.; Kuzil, A.C.; Padun, S.P.; Surnachov, S.I.; Jakovlev, G.V.

    2001-01-01

    Experience with the development and utilization of the program-technical complex PTK 'KRUIZ' is analyzed in the paper. A peculiarity of PTK is the orientation on acquisition, processing and diagnostics of signals from in-reactor sensors (thermocouples and SPD). The PTK 'KRUIZ' represents a new generation of tools open for further development, oriented specifically on the use in in-reactor control systems in modernized and built power units of the WWER type. In the PTK 'KRUIZ', methods, models and algorithms proved in nuclear power plants are used accounting for the utilization of up to date technical tools and systematic technical solutions. Experience with the use of basic elements of the PTK 'KRUIZ' at existing WWER reactors including peculiarities of temperature control in nuclear power plants are also dealt within the paper. (Authors)

  9. Design of a nuclear reactor cooperative controller

    International Nuclear Information System (INIS)

    Alang-Rashid, N.K.; Heger, A.S.

    1991-01-01

    This paper describes the development of a fuzzy logic controller software package and explores the feasibility of its use in nuclear reactor operation. The controller complements reactor operator actions, and the operators can override the controller decisions. Techniques of providing learning capability to the controller are also being investigated to improve the reasoning and control skill of the controller. The fuzzy logic controller is implemented in C language and its overall structure is shown. The heart of the systems consists of a fuzzifier, a rule interpreter, and a defuzzifier. The controller is designed as a stand-alone package that can be interfaced to a simulated model of a nuclear reactor. Since no model is an accurate representation of the actual process being modeled, some tuning must be performed to use the controller in an actual reactor. This is accomplished using the learning feature of the controller

  10. Cyber security level assignment for research reactor digital instrumentation and control system architecture using concept of defense in depth

    International Nuclear Information System (INIS)

    Shin, Jin Soo; Heo, Gyun Young; Son, Han Seong; Kim, Young Ki; Park, Jaek Wan

    2012-01-01

    Due to recent aging of the analog instrumentation of many nuclear power plants (NPPs) and research reactors, the system reliability decreases while maintenance and testing costs increase. In addition, it is difficult to find the substitutable analog equipment s due to obsolescence. Therefore, the instrumentation and control (I and C) systems have changed from analog system to digital system due to these facts. With the introduction of digital systems, research reactors are forced to care for the problem of cyber attacks because I and C systems have been digitalized using networks or communication systems. Especially, it is more issued at research reactors due to the accessibility of human resources. In the real world, an IBM researcher has been successful in controlling the software by penetrating a NPPs network in U.S. on July 2008 and acquiring the control right of nuclear facilities after one week. Moreover, the malignant code called 'stuxnet' impaired the nearly 1,000 centrifugal separators in Iran according to an IAEA report. The problem of cyber attacks highlights the important of cyber security, which should be emphasized. Defense.in.depth (DID) is a significant concept for the cyber security to work properly. DID institutes and maintains a hardy program for critical digital asset (CDA) by implementing multiple security boundaries. In this work, we assign cyber security levels to a typical digital I and C system using DID concept. This work is very useful in applying the concept of DID to nuclear industry with respect to cyber security

  11. Cyber security level assignment for research reactor digital instrumentation and control system architecture using concept of defense in depth

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Jin Soo; Heo, Gyun Young [Kyung Hee University, Seoul (Korea, Republic of); Son, Han Seong [Joongbu Univ., Chungnam (Korea, Republic of); Kim, Young Ki; Park, Jaek Wan [KAERI, Daejeon (Korea, Republic of)

    2012-10-15

    Due to recent aging of the analog instrumentation of many nuclear power plants (NPPs) and research reactors, the system reliability decreases while maintenance and testing costs increase. In addition, it is difficult to find the substitutable analog equipment s due to obsolescence. Therefore, the instrumentation and control (I and C) systems have changed from analog system to digital system due to these facts. With the introduction of digital systems, research reactors are forced to care for the problem of cyber attacks because I and C systems have been digitalized using networks or communication systems. Especially, it is more issued at research reactors due to the accessibility of human resources. In the real world, an IBM researcher has been successful in controlling the software by penetrating a NPPs network in U.S. on July 2008 and acquiring the control right of nuclear facilities after one week. Moreover, the malignant code called 'stuxnet' impaired the nearly 1,000 centrifugal separators in Iran according to an IAEA report. The problem of cyber attacks highlights the important of cyber security, which should be emphasized. Defense.in.depth (DID) is a significant concept for the cyber security to work properly. DID institutes and maintains a hardy program for critical digital asset (CDA) by implementing multiple security boundaries. In this work, we assign cyber security levels to a typical digital I and C system using DID concept. This work is very useful in applying the concept of DID to nuclear industry with respect to cyber security.

  12. Investigation of the Bilibin reactor operation in the regime of automatic power and frequency control in isolated power system

    International Nuclear Information System (INIS)

    Sankovskij, G.A.; Molochkov, V.I.; Dolgov, V.V.; Soldatov, G.E.; Minashin, M.E.

    1981-01-01

    The results of experimental investigations of the power unit operation of the Bilibin nuclear power and heating plant (BNPHP) in the regime of automatic power and frequency control in an isolated power system are presented. The BNPHP comprises four similar power units. Each unit includes a steam generating setup - the channel water-graphite reactor with tubular fuel elements with natural circulation of boiling water at all the power levels as well as a turbosetup with two heat selectors and a turbogenerator. The turbine operates on dry saturated steam (with intermediate separation) which is brought from the drum-separator of the reactor natural circulation circuit. The BNPHP operates according to the controller schedule since the start-up of the first power unit. The BNPHP unit power varifies within the 50-100% range 3-4 times per day (by the number of maxima in the schedule of the power system loadings). Two design flowsheets of the unit power control and dynamic characteristics of the system for both vatiants are considered. It is concluded that both investigated automatic control systems are seviceable and deviations of the reactor parameters within the transients are not dangerous for heat release from the core. The plant is better shielded from external mainly short-term perturbations coming from the power system when the system operates in accordance with the first variant of the flowsheet [ru

  13. Nuclear reactor monitoring system

    International Nuclear Information System (INIS)

    Drummond, C.N.; Bybee, R.T.; Mason, F.L.; Worsham, H.J.

    1976-01-01

    The invention pertains to an improved monitoring system for the neutron flux in a nuclear reactor. It is proposed to combine neutron flux detectors, a thermoelement, and a background radiation detector in one measuring unit. The spatial arrangement of these elements is fixed with great exactness; they are enclosed by an elastic cover and are brought into position in the reactor with the aid of a bent tube. The arrangement has a low failure rate and is easy to maintain. (HP) [de

  14. Coordination of motor systems of the control bars of TRIGA Mark III reactor, through the use of a PLC; Coordinacion de los sistemas motrices de las barras de control del reactor Triga Mark III, mediante el uso de un PLC

    Energy Technology Data Exchange (ETDEWEB)

    Castro R, P. R.

    2016-07-01

    The use of programmable logic controllers (PLCs) has a wide field in the realization of automatic systems, since is sought that the form of control is easy for any user. In this work, the TRIGA Mark III reactor of Instituto Nacional de Investigaciones Nucleares (ININ) is intended to operate with a programming code in PLC for the automation of the control rods, having uniform wear according to the power required by the user. On the other hand, is proposed to develop an HMI graphical interface for communication via Ethernet, allowing supervision during the reactor operation process and greater protection of operators during reactor startup. The accuracy of the new actuators, as well as their durability, will allow a good performance of the reactor for many years to come. (Author)

  15. Reactor feedwater system

    International Nuclear Information System (INIS)

    Kagaya, Hiroyuki; Tominaga, Kenji.

    1993-01-01

    In a simplified water type reactor using a gravitationally dropping emergency core cooling system (ECCS), the present invention effectively prevents remaining high temperature water in feedwater pipelines from flowing into the reactor upon occurrence of abnormal events. That is, (1) upon LOCA, if a feedwater pipeline injection valve is closed, boiling under reduced pressure of the remaining high temperature water occurs in the feedwater pipelines, generated steams prevent the remaining high temperature water from flowing into the reactor. Accordingly, the reactor is depressurized rapidly. (2) The feedwater pipeline injection valve is closed and a bypassing valve is opened. Steams generated by boiling under reduced pressure of the remaining high temperature water in the feedwater pipelines are released to a condensator or a suppression pool passing through bypass pipelines. As a result, the remaining high temperature water is prevented from flowing into the reactor. Accordingly, the reactor is rapidly depressurized and cooled. It is possible to accelerate the depressurization of the reactor by the method described above. Further, load on the depressurization valve disposed to a main steam pipe can be reduced. (I.S.)

  16. Study on simulation, control and online assistance integrated system of 10 MW high temperature gas-cooled test reactor

    International Nuclear Information System (INIS)

    Luo, S.; Shi, L.; Zhu, S.

    2004-01-01

    In order to provide a convenient tool for engineering designed, safety analysis, operator training and control system design of the high temperature gas-cooled test reactor (HTR), an integrated system for simulation, control and online assistance of the HTR-10 has been designed and is still under development by the Institute of Nuclear Energy Technology (INET) of Tsinghua University in China. The whole system is based on a network environment and includes three subsystems: the simulation subsystem (SIMUSUB), the visualized control designed subsystem (VCDSUB) and the online assistance subsystem (OASUB). The SIMUSUB consists of four parts: the simulation calculating server (SCS), the main control client (MCC), the data disposal client (DDC) and the results graphic display client (RGDC), all of which can communicate with each other via network. The SIMUSUB is intended to analyze and calculate the physical processes of the reactor core, the main loop system and the stream generator, etc., as well as to simulate the normal operation and transient accidents, and the result data can be graphically displayed through the RGDC dynamically. The VCDSUB provides a platform for control system modeling where the control flow systems can be automatically generated and graphically simulated. Based on the data from the field bus, the OASUB provides some of the reactor core parameter, which are difficult to measure. This whole system can be used as an educational tool to understand the design and operational characteristics of the HTR-10, and can also provide online supports for operators in the main control room, or as a convenient powerful tool for the control system design. (authors)

  17. The power control system of the Siemens-KWU nuclear power station of the PWR [pressurized water reactors] type

    International Nuclear Information System (INIS)

    Huber, Horacio

    1989-01-01

    Starting with the first nuclear power plant constructed by Siemens AG of the pressurized light water reactor line (PWR), the Obrigheim Nuclear Power Plant (340 MWe net), until the recently constructed plants of 1300 MWe (named 'Konvoi'), the design of the power control system of the plant was continuously improved and optimized using the experience gained in the operation of the earlier generations of plants. The reactor power control system of the Siemens - KWU nuclear power plants is described. The features of this design and of the Siemens designed heavy water power plants (PHWR) Atucha I and Atucha II are mentioned. Curves showing the behaviour of the controlled variables during load changes obtained from plant tests are also shown. (Author) [es

  18. Reactor protection system

    International Nuclear Information System (INIS)

    Fairbrother, D.B.; Lesniak, L.M.; Orgera, E.G.

    1977-10-01

    The report describes the reactor protection system (RPS-II) designed for use on Babcock and Wilcox 145-, later 177-, and 205-fuel assembly pressurized water reactors. In this system, relays in the trip logic have been replaced by solid state devices. A calculating module for the low DNBR, pump status, and offset trip functions has replaced the overpower trip (based on flow and imbalance), the power/RC pump trip, and the variable low-pressure trip. Included is a description of the changes from the present Oconee-type reactor protection system (RPS-I), a functional and hardware description of the calculating module, a description of the software programmed in the calculating module, and a discussion of the qualification program conducted to ensure that the degree of protection provided by RPS-II is not less than that provided by previously licensed systems supplied by B and W

  19. Design strategy for control of inherently safe reactors

    International Nuclear Information System (INIS)

    Chisholm, G.H.

    1984-01-01

    Reactor power plant safety is assured through a combination of engineered barriers to radiation release (e.g., reactor containment) in combination with active reactor safety systems to shut the reactor down and remove decay heat. While not specifically identified as safety systems, the control systems responsible for continuous operation of plant subsystems are the first line of defense for mitigating radiation releases and for plant protection. Inherently safe reactors take advantage of passive system features for decay-heat removal and reactor shutdown functions normally ascribed to active reactor safety systems. The advent of these reactors may permit restructuring of the present control system design strategy. This restructuring is based on the fact that authority for protection against unlikely accidents is, as much as practical, placed upon the passive features of the system instead of the traditional placement upon the PPS. Consequently, reactor control may be simplified, allowing the reliability of control systems to be improved and more easily defended

  20. Measurements and simulations of electromagnetic perturbations on neutronic and thermodynamic control systems of nuclear reactors

    International Nuclear Information System (INIS)

    Lecompte, J.-C.; Buisson, Jacques.

    1981-01-01

    Neutronics and thermodynamics measuring devices realize the control and nuclear safety of the reactor and the core. They cause the emergency stop of the reactor if the informations given by the sensors are out of the normal operation tolerances. Defining and assuring a protection level sufficiently high against electromagnetic disturbances is essential for avoiding those measuring devices of giving erroneous indications in the environment where they are located. To obtain this result, one procceds according to the following three steps: a) In situ measurements of the ambient level of disturbances (oscilloscope analysis, statistical measurements). b) Measurement in laboratory of the immunity against the disturbances caused by every constituent of the device (electronic cables, connectors) and then the complete device. c) In situ simulation on the devices of the perturbations caused by the measured ambient level [fr

  1. The nuclear reactor systems

    International Nuclear Information System (INIS)

    Bacher, P.

    2008-01-01

    This paper describes the various nuclear reactor systems, starting with the Generation II, then the present development of the Generation III and the stakes and challenges of the future Generation IV. Some have found appropriate to oppose reactor systems or generations one to another, especially by minimizing the enhancements of generation III compared to generation II or by expecting the earth from generation IV (meaning that generation III is already obsolete). In the first part of the document (chapter 2), some keys are given to the reader to develop its proper opinion. Chapter 3 describes more precisely the various reactor systems and generations. Chapter 4 discusses the large industrial manoeuvres around the generation III, and the last chapter gives some economical references, taking into account, for the various means of power generation, the impediments linked to climate protection

  2. Reactor water level control device

    International Nuclear Information System (INIS)

    Hiramatsu, Yohei.

    1980-01-01

    Purpose: To increase the rapid response of the waterlevel control converting a reactor water level signal into a non-linear type, when the water level is near to a set value, to stabilize the water level reducting correlatively the reactor water level variation signal to stabilize greatly from the set value, and increasing the variation signal. Constitution: A main vapor flow quality transmitter detects the vapor flow generated in a reactor and introduced into a turbine. A feed water flow transmitter detects the quantity of a feed water flow from the turbine to the reactor, this detected value is sent to an addition operating apparatus. On the other hand, the power signal of the reactor water level transmitter is sent to the addition operating apparatus through a non-linear water level signal converter. The addition operation apparatus generates a signal for requesting the feed water flow quantity from both signals. Upon this occasion, the reactor water level signal converter makes small the reactor water level variation when the reactor level is close the set value, and when the water level deviates greatly from the set value, the reactor water level variation is made large thereby to improve the rapid response of the reactor coater level control. (Yoshino, Y.)

  3. 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

  4. Nuclear reactor refueling system

    International Nuclear Information System (INIS)

    Wade, E.E.

    1978-01-01

    A system for transferring fuel assemblies between a nuclear reactor core and a fuel storage area while the fuel assembies remain completely submerged in a continuous body of coolant is described. The system comprises an in-vessel fuel transfer machine located inside the reactor vessel and an ex-vessel fuel transfer machine located in a fuel storage tank. The in-vessel fuel transfer machine comprises two independently rotatable frames with a pivotable fuel transfer apparatus disposed on the lower rotatable frame. The ex-vessel fuel transfer machine comprises one frame with a pivotable fuel transfer apparatus disposed thereon. The pivotable apparatuses are capable of being aligned with each other to transfer a fuel assembly between the reactor vessel and fuel storage tank while the fuel assembly remains completely submerged in a continuous body of coolant. 9 claims, 7 figures

  5. Technology Transfer Programme In Reactor Digital Instrumentation And Control System (REDICS) Project: Knowledge, Experiences And Future Expectations

    International Nuclear Information System (INIS)

    Nurfarhana Ayuni Joha; Mohamad Puad Abu; Izhar Abu Hussin; Ridzuan Abdul Mutalib; Zareen Khan Abdul Jalil Khan; Mohd Khairulezwan Abdul Manan; Mohd Sabri Minhat; Mohd Idris Taib

    2013-01-01

    The PUSPATI TRIGA MARK II research reactor in Malaysia was commissioned in 1982. After 31 years of operation, Nuclear Malaysia is taking an approach for a better research and development in nuclear radiations as well as the technical services that provided. Reactor TRIGA PUSPATI (RTP) is currently upgrading its control console from analogue to digital system. The Reactor Digital Instrumentation and Control System (ReDICS) project is done on cooperation with Korea Atomic Energy Research Institute (KAERI), Korea including the technical part from the design stage until commissioning as well as the Technology Transfer Program (TTP). TTP in this ReDICS project is a part of Human Resource and System Development Program. It was carried out from the design stage until the commissioning of the system. It covers all subjects related to the design on the digital system and the requirements for the operation of RTP. The objective of this paper is to share the knowledge and experiences gained through this ReDICS project. This paper will also discuss the future expectations from this ReDICS project for Nuclear Malaysia and its personnel, as well as to the country. (author)

  6. ADVANCED CONTROL FOR A ETHYLENE REACTOR

    Directory of Open Access Journals (Sweden)

    Dumitru POPESCU

    2017-06-01

    Full Text Available The main objective of this work is the design and implementation of control solutions for petrochemical processes, namely the control and optimization of a pyrolysis reactor, the key-installation in the petrochemical industry. We present the technological characteristics of this petrochemical process and some aspects about the proposed control system solution for the ethylene plant. Finally, an optimal operating point for the reactor is found, considering that the process has a nonlinear multi-variable structure. The results have been implemented on an assembly of pyrolysis reactors on a petrochemical platform from Romania.

  7. Design and verification of computer-based reactor control system modification at Bruce-A candu nuclear generating station

    International Nuclear Information System (INIS)

    Basu, S.; Webb, N.

    1995-01-01

    The Reactor Control System at Bruce-A Nuclear Generating Station is going through some design modifications, which involve a rigorous design process including independent verification and validation. The design modification includes changes to the control logic, alarms and annunciation, hardware and software. The design (and verification) process includes design plan, design requirements, hardware and software specifications, hardware and software design, testing, technical review, safety evaluation, reliability analysis, failure mode and effect analysis, environmental qualification, seismic qualification, software quality assurance, system validation, documentation update, configuration management, and final acceptance. (7 figs.)

  8. Heat Transfer Salts for Nuclear Reactor Systems - Chemistry Control, Corrosion Mitigation, and Modeling

    International Nuclear Information System (INIS)

    Anderson, Mark; Sridharan, Kumar; Morgan, Dane; Peterson, Per; Calderoni, Pattrick; Scheele, Randall; Casekka, Andrew; McNamara, Bruce

    2015-01-01

    The concept of a molten salt reactor has existed for nearly sixty years. Previously all work was done during a large collaborative effort at Oak Ridge National Laboratory, culminating in a research reactor which operated for 15,000 hours without major error. This technical success has garnished interest in modern, high temperature, reactor schemes. Research using molten fluoride salts for nuclear applications requires a steady supply of high grade molten salts. There is no bulk supplier of research grade fluoride salts in the world, so a facility which could provide all the salt needed for testing at the University of Wisconsin had to be produced. Two salt purification devices were made for this purpose, a large scale purifier, and a small scale purifier, each designed to clean the salts from impurities and reduce their corrosion potential. As of now, the small scale has performed with flibe salt, hydrogen, and hydrogen fluoride, yielding clean salt. This salt is currently being used in corrosion testing facilities at the Massachusetts Institute of Technology and the University of Wisconsin. Working with the beryllium based salts requires extensive safety measures and health monitoring to prevent the development of acute or chronic beryllium disease, two pulmonary diseases created by an allergic reaction to beryllium in the lungs. Extensive health monitoring, engineering controls, and environment monitoring had to be set up with the University of Wisconsin department of Environment, Health and Safety. The hydrogen fluoride required for purification was also an extreme health hazard requiring thoughtful planning and execution. These dangers have made research a slow and tedious process. Simple processes, such as chemical handling and clean-up, can take large amounts of ingenuity and time. Other work has complemented the experimental research at Wisconsin to advance high temperature reactor goals. Modeling work has been performed in house to re

  9. Heat Transfer Salts for Nuclear Reactor Systems - Chemistry Control, Corrosion Mitigation, and Modeling

    Energy Technology Data Exchange (ETDEWEB)

    Anderson, Mark [Univ. of Wisconsin, Madison, WI (United States); Sridharan, Kumar [Univ. of Wisconsin, Madison, WI (United States); Morgan, Dane [Univ. of Wisconsin, Madison, WI (United States); Peterson, Per [Univ. of Wisconsin, Madison, WI (United States); Calderoni, Pattrick [Univ. of Wisconsin, Madison, WI (United States); Scheele, Randall [Univ. of Wisconsin, Madison, WI (United States); Casekka, Andrew [Univ. of Wisconsin, Madison, WI (United States); McNamara, Bruce [Univ. of Wisconsin, Madison, WI (United States)

    2015-01-22

    The concept of a molten salt reactor has existed for nearly sixty years. Previously all work was done during a large collaborative effort at Oak Ridge National Laboratory, culminating in a research reactor which operated for 15,000 hours without major error. This technical success has garnished interest in modern, high temperature, reactor schemes. Research using molten fluoride salts for nuclear applications requires a steady supply of high grade molten salts. There is no bulk supplier of research grade fluoride salts in the world, so a facility which could provide all the salt needed for testing at the University of Wisconsin had to be produced. Two salt purification devices were made for this purpose, a large scale purifier, and a small scale purifier, each designed to clean the salts from impurities and reduce their corrosion potential. As of now, the small scale has performed with flibe salt, hydrogen, and hydrogen fluoride, yielding clean salt. This salt is currently being used in corrosion testing facilities at the Massachusetts Institute of Technology and the University of Wisconsin. Working with the beryllium based salts requires extensive safety measures and health monitoring to prevent the development of acute or chronic beryllium disease, two pulmonary diseases created by an allergic reaction to beryllium in the lungs. Extensive health monitoring, engineering controls, and environment monitoring had to be set up with the University of Wisconsin department of Environment, Health and Safety. The hydrogen fluoride required for purification was also an extreme health hazard requiring thoughtful planning and execution. These dangers have made research a slow and tedious process. Simple processes, such as chemical handling and clean-up, can take large amounts of ingenuity and time. Other work has complemented the experimental research at Wisconsin to advance high temperature reactor goals. Modeling work has been performed in house to re

  10. Nuclear reactor trip system

    International Nuclear Information System (INIS)

    Cook, B.M.

    1982-01-01

    Each parameter of the processes of a nuclear reactor and components operatively associated with it is monitored by a set of four like sensors. A trip system normally operates on a ''two out four'' configuration; i.e., to trip the reactor it is necessary that at least two sensors of a set sense an off-normal parameter. This assumes that all sensors are in normal operating condition. However, when a sensor is in test or is subject to maintenance or is defective or disabled, the ''two out of four''configuration would be reduced to a ''one out of three'' configuration because the affected sensor is taken out of service. This would expose the system to the possibility that a single sensor failure, which may be spurious, will cause a trip of the reactor. To prevent this, it is necessary that the affected sensor be bypassed. If only one sensor is bypassed, the system operates on a ''two out of three'' configuration. With two sensors bypassed, the sensing of an off-normal parameter by a third sensor trips the reactor. The by-pass circuit also disables the circuit coupling the by-passed sensor to the trip circuit. (author)

  11. French experience in the programmed systems for nuclear reactor control and protection

    International Nuclear Information System (INIS)

    Jover, P.

    1986-03-01

    The analysis of incidents during the start-up of the first nuclear power plant 1300 MWe has made possible to obtain good performances evaluation of the two computerized control and protection systems: the protection system (SPIN) and the logic control system (CONTROBLOC). The results of this experiment have shown that the objectives have been attained [fr

  12. Design project of the dosimetry control system in the independent CO2 loop for cooling the samples irradiated in the RA reactor vertical experimental channels, Vol. V

    International Nuclear Information System (INIS)

    1964-01-01

    Design project of the dosimetry control system in the independent CO 2 loop for cooling the samples irradiated in the RA reactor vertical experimental channels includes the following: calculations of CO 2 gas activity, design of the dosimetry control system, review of the changes that should be done in the RA reactor building for installing the independent CO 2 loop, specification of the materials with cost estimation, engineering drawings of the system [sr

  13. 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.)

  14. Nuclear reactor power supply system

    International Nuclear Information System (INIS)

    Cook, B.M.

    1982-01-01

    The redundant signals from the sensor assemblies measuring the process parameters of a nuclear reactor power supply are transmitted each in its turn to a protection system which operates to actuate the protection apparatus for signals indicating off-process conditions. Each sensor assembly includes a number of like sensors measuring the same parameters. The sets of process signals derived from the sensor assemblies are each in its turn transmitted from the protection system to the control system which impresses control signals on the reactor or its components to counteract the tendency for conditions to drift off-normal status requiring operation of the protection system. A parameter signal selector prevents a parameter signal which differs from the other parameter signals of the set by more than twice the allowable variation from passing to the control system. Test signals are periodically impressed by a test unit on a selected pair of a selection unit and control channels. This arrangement eliminates the possibility that a single component failure which may be spurious will cause an inadvertent trip of the reactor during test. (author)

  15. Expert system for assisting the repair operations on the control racks of the control rods assembly in a 900 MW PWR type reactor

    International Nuclear Information System (INIS)

    Monnier, B.; Doutre, J.L.; Franco, A.

    1990-01-01

    The expert system presented was developed for assisting the repair operations on the control equipment of the control rod assembly in a PWR type reactor. The expert system allows the representation of expert knowledge and diagnostic reasoning. The objective of the expert system is to achieve the most precise diagnostic and localizing of the breakdown elements, by processing the data acquired during breakdown. The development steps, the structure and the applications of the expert system are summarized. The expert system operates in an IBM PC equipped with a AMAIA 8 Mo card. A time schedule of 18 months is predicted [fr

  16. An intelligent nuclear reactor core controller for load following operations, using recurrent neural networks and fuzzy systems

    International Nuclear Information System (INIS)

    Boroushaki, M.; Ghofrani, M.B.; Lucas, C.; Yazdanpanah, M.J.

    2003-01-01

    In the last decade, the intelligent control community has paid great attention to the topic of intelligent control systems for nuclear plants (core, steam generator...). Papers mostly used approximate and simple mathematical SISO (single-input-single-output) model of nuclear plants for testing and/or tuning of the control systems. They also tried to generalize theses models to a real MIMO (multi-input-multi-output) plant, while nuclear plants are typically of complex nonlinear and multivariable nature with high interactions between their state variables and therefore, many of these proposed intelligent control systems are not appropriate for real cases. In this paper, we designed an on-line intelligent core controller for load following operations, based on a heuristic control algorithm, using a valid and updatable recurrent neural network (RNN). We have used an accurate 3-dimensional core calculation code to represent the real plant and to train the RNN. The results of simulation show that this intelligent controller can control the reactor core during load following operations, using optimum control rod groups manoeuvre and variable overlapping strategy. This methodology represents a simple and reliable procedure for controlling other complex nonlinear MIMO plants, and may improve the responses, comparing to other control systems

  17. Reactor system safety assurance

    International Nuclear Information System (INIS)

    Mattson, R.J.

    1984-01-01

    The philosophy of reactor safety is that design should follow established and conservative engineering practices, there should be safety margins in all modes of plant operation, special systems should be provided for accidents, and safety systems should have redundant components. This philosophy provides ''defense in depth.'' Additionally, the safety of nuclear power plants relies on ''safety systems'' to assure acceptable response to design basis events. Operating experience has shown the need to study plant response to more frequent upset conditions and to account for the influence of operators and non-safety systems on overall performance. Defense in depth is being supplemented by risk and reliability assessment

  18. 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)

  19. Digital control for nuclear reactors - lessons learned

    International Nuclear Information System (INIS)

    Bernard, J.A.; Aviles, B.N.; Lanning, D.D.

    1992-01-01

    Lessons learned during the course of the now decade-old MIT program on the digital control of nuclear reactors are enumerated. Relative to controller structure, these include the importance of a separate safety system, the need for signal validation, the role of supervisory algorithms, the significance of command validation, and the relevance of automated reasoning. Relative to controller implementation, these include the value of nodal methods to the creation of real-time reactor physics and thermal hydraulic models, the advantages to be gained from the use of real-time system models, and the importance of a multi-tiered structure to the simultaneous achievement of supervisory, global, and local control. Block diagrams are presented of proposed controllers and selected experimental and simulation-study results are shown. In addition, a history is given of the MIT program on reactor digital control

  20. Time-optimal control of reactor power

    International Nuclear Information System (INIS)

    Bernard, J.A.

    1987-01-01

    Control laws that permit adjustments in reactor power to be made in minimum time and without overshoot have been formulated and demonstrated. These control laws which are derived from the standard and alternate dynamic period equations, are closed-form expressions of general applicability. These laws were deduced by noting that if a system is subject to one or more operating constraints, then the time-optimal response is to move the system along these constraints. Given that nuclear reactors are subject to limitations on the allowed reactor period, a time-optimal control law would step the period from infinity to the minimum allowed value, hold the period at that value for the duration of the transient, and then step the period back to infinity. The change in reactor would therefore be accomplished in minimum time. The resulting control laws are superior to other forms of time-optimal control because they are general-purpose, closed-form expressions that are both mathematically tractable and readily implanted. Moreover, these laws include provisions for the use of feedback. The results of simulation studies and actual experiments on the 5 MWt MIT Research Reactor in which these time-optimal control laws were used successfully to adjust the reactor power are presented

  1. Fabrication of cermet bearings for the control system of a high temperature lithium cooled nuclear reactor

    Science.gov (United States)

    Yacobucci, H. G.; Heestand, R. L.; Kizer, D. E.

    1973-01-01

    The techniques used to fabricate cermet bearings for the fueled control drums of a liquid metal cooled reference-design reactor concept are presented. The bearings were designed for operation in lithium for as long as 5 years at temperatures to 1205 C. Two sets of bearings were fabricated from a hafnium carbide - 8-wt. % molybdenum - 2-wt. % niobium carbide cermet, and two sets were fabricated from a hafnium nitride - 10-wt. % tungsten cermet. Procedures were developed for synthesizing the material in high purity inert-atmosphere glove boxes to minimize oxygen content in order to enhance corrosion resistance. Techniques were developed for pressing cylindrical billets to conserve materials and to reduce machining requirements. Finishing was accomplished by a combination of diamond grinding, electrodischarge machining, and diamond lapping. Samples were characterized in respect to composition, impurity level, lattice parameter, microstructure and density.

  2. Application of Decomposition Methodology to Solve Integrated Process Design and Controller Design Problems for Reactor-Separator-Recycle System

    DEFF Research Database (Denmark)

    Abd.Hamid, Mohd-Kamaruddin; Sin, Gürkan; Gani, Rafiqul

    2010-01-01

    This paper presents the integrated process design and controller design (IPDC) for a reactor-separator-recycle (RSR) system and evaluates a decomposition methodology to solve the IPDC problem. Accordingly, the IPDC problem is solved by decomposing it into four hierarchical stages: (i) pre...... the design of a RSR system involving consecutive reactions, A B -> C and shown to provide effective solutions that satisfy design, control and cost criteria. The advantage of the proposed methodology is that it is systematic, makes use of thermodynamic-process knowledge and provides valuable insights......-analysis, (ii) design analysis, (iii) controller design analysis, and (iv) final selection and verification. The methodology makes use of thermodynamic-process insights and the reverse design approach to arrive at the final process-controller design decisions. The developed methodology is illustrated through...

  3. Methodology for advanced control rooms assessment of nuclear reactors: case study using Laboratory of Human System Interface (LABIHS)

    International Nuclear Information System (INIS)

    Carvalho, Eduardo Ferro; Verboonen, Monique; Carvalho, Bruno Batista de

    2005-01-01

    A control room of a nuclear reactor is a complex system that controls a thermodynamic process used to produce electric energy. The operators interact with the control room through interfaces and several monitoring stations. These interfaces present significant implications for the safety of the nuclear power plant, once they influence the activities of the operators, affect the way how operators receive information related with the status from the main systems and determine the necessary requirements so that the operators understand and supervise the main parameters. This article intends to present the methodology and the results of the evaluation carried through in the advanced control room of a compact simulator, that uses as reference a nuclear plant PWR of the Westinghouse. The structure used for evaluation of the simulator is formed by the guideline of human factors of the NRC, the NUREG 700, checklist, questionnaires and the analysis of the operator's activity. (author)

  4. Study of the application of advanced control systems to fusion experiments and reactors. Final report

    International Nuclear Information System (INIS)

    1974-05-01

    The work accomplished to date toward the formulation of an advanced control system concept for large-scale magnetically confined thermonuclear fusion devices is summarized. The work was concentrated in three major areas: (1) general control studies and identification of control issues, (2) exploration of possible direct interactions with AEC National Laboratories, and (3) identification of simulation requirements to support control studies. (U.S.)

  5. Nuclear reactor sealing system

    International Nuclear Information System (INIS)

    McEdwards, J.A.

    1983-01-01

    A liquid metal-cooled nuclear reactor sealing system is disclosed. The nuclear reactor includes a vessel sealed at its upper end by a closure head. The closure head comprises at least two components, one of which is rotatable; and the two components define an annulus therebetween. The sealing system includes at least a first and second inflatable seal disposed in series in an upper portion of the annulus. The system further includes a dip seal extending into a body of insulation located adjacent a bottom portion of the closure head. The dip seal comprises a trough formed by a lower portion of one of the components, and a seal blade pendently supported from the other component and extending downwardly into the trough. A body of liquid metal is contained in the trough which submerges a portion of the seal blade. The seal blade is provided with at least one aperture located above the body of liquid metal for providing fluid communication between the annulus intermediate the dip seal and the inflatable seals, and a body of cover gas located inside the vessel. There also is provided means for introducing a purge gas into the annulus intermediate the inflatable seals and the seal blade. The purge gas is introduced in an amount sufficient to substantially reduce diffusion of radioactive cover gas or sodium vapor up to the inflatable seals. The purge gas mixes with the cover gas in the reactor vessel where it can be withdrawn from the vessel for treatment and recycle to the vessel

  6. Some problems of software development for the plant-level automated control system of NPPs with the RBMK reactors

    International Nuclear Information System (INIS)

    Gorbunov, V.P.; Egorov, A.K.; Isaev, N.V.; Saprykin, E.M.

    1987-01-01

    Problems on development and operation of automated control system (ACS) software of NPPs with the RBMK reactors are discussed. The ES computer with large on-line storage (not less than 1 Mbite) and fast response (not less than 300.000 of operations per a second) should enter the ACS composition. Several program complexes are used in the NPP ACS. The programs collected into the EhNERGIYa library are used to provide central control system operation. The information-retrival system called the Fuel file is used to automate NPP fuel motion account, as well as to estimate efficiency of fuel application, to carry out calculations of a fuel component of electric and heat energy production cost. The automated information system for unit operation efficiency analysis, which solves both plant and unit-level problems, including engineering and economical factors and complexing of operation parameter bank, is under trial operation

  7. Technique of nuclear reactors controls

    International Nuclear Information System (INIS)

    Weill, J.

    1953-12-01

    This report deal about 'Techniques of control of the nuclear reactors' in the goal to achieve the control of natural uranium reactors and especially the one of Saclay. This work is mainly about the measurement into nuclear parameters and go further in the measurement of thermodynamic variables,etc... putting in relief the new features required on behalf of the detectors because of their use in the thermal neutrons flux. In the domain of nuclear measurement, we indicate the realizations and the results obtained with thermal neutron detectors and for the measurement of ionizations currents. We also treat the technical problem of the start-up of a reactor and of the reactivity measurement. We give the necessary details for the comprehension of all essential diagrams and plans put on, in particular, for the reactor of Saclay. (author) [fr

  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. Safety evaluation for instrumentation and control system upgrading project of Malaysian TRIGA MARK II PUSPATI Research reactor

    International Nuclear Information System (INIS)

    Ridha Roslan; Nik Mohd Faiz Khairuddin

    2013-01-01

    Full-text: Malaysian TRIGA MARK II research reactor has been in safe operation since its first criticality in 1982. The reactor is licensed to be operated by Malaysian Nuclear Agency to perform training and research development related activities. Due to its extensive operation since last three decades, the option of modifications for safety and safety-related item and component become a necessary to replace the outdated equipment to a stat-of-art, reliable technologies. This paper will present the current regulatory activities performed by Atomic Energy Licensing Board (AELB) to ensure the upgrading of analogue to digital instrumentation and control system is implemented in safe manner. The review activity includes documentation review, manufacturer quality audit and on-site inspection for commissioning. The review performed by AELB is based on The International Atomic Energy Agency (IAEA) Safety Requirements NS-R-4, entitled Safety of Research Reactors. During this endeavour, AELB seeks technical cooperation from Korea Institute of Nuclear Safety (KINS), the nuclear experts organization of the country of origin of the instrumentation and control technology. The regulatory activity is still on-going and is expected to be completed by issuance of Authorization for Restart on December 2013. (author)

  10. Thermionic nuclear reactor systems

    International Nuclear Information System (INIS)

    Kennel, E.B.

    1986-01-01

    Thermionic nuclear reactors can be expected to be candidate space power supplies for power demands ranging from about ten kilowatts to several megawatts. The conventional ''ignited mode'' thermionic fuel element (TFE) is the basis for most reactor designs to date. Laboratory converters have been built and tested with efficiencies in the range of 7-12% for over 10,000 hours. Even longer lifetimes are projected. More advanced capabilities are potentially achievable in other modes of operation, such as the self-pulsed or unignited diode. Coupled with modest improvements in fuel and emitter material performance, the efficiency of an advanced thermionic conversion system can be extended to the 15-20% range. Advanced thermionic power systems are expected to be compatible with other advanced features such as: (1) Intrinsic subcritically under accident conditions, ensuring 100% safety upon launch abort; (2) Intrinsic low radiation levels during reactor shutdown, allowing manned servicing and/or rendezvous; (3) DC to DC power conditioning using lightweight power MOSFETS; and (4) AC output using pulsed converters

  11. Operating experiences with programmable logic controller (PLC) system of Indian Pressurised Heavy Water Reactors (PHWR)

    International Nuclear Information System (INIS)

    Ughade, A.V.; Singh, Ranjeet; Bhattacharya, P.K.; Kulkarni, R.K.; Chandra, Umesh

    2005-01-01

    PLC system was introduced for the first time in Kaiga-1,2 and RAPS-3,4 Nuclear Power Plants (NPPs) for Station Logic Control of Non Safety Related (NSR) and Safety related (SR) systems. However, the safety system logics are still relay based. The experience on the deployment of PLC system, which is computer-based, has brought out various implementation issues. This paper give details of such experiences, the solutions emerged and applied for plants under operation/construction. (author)

  12. Multivariable Feedback Control of Nuclear Reactors

    Directory of Open Access Journals (Sweden)

    Rune Moen

    1982-07-01

    Full Text Available Multivariable feedback control has been adapted for optimal control of the spatial power distribution in nuclear reactor cores. Two design techniques, based on the theory of automatic control, were developed: the State Variable Feedback (SVF is an application of the linear optimal control theory, and the Multivariable Frequency Response (MFR is based on a generalization of the traditional frequency response approach to control system design.

  13. Model for spatial synthesis of automated control system of the GCR type reactor; Model za prostornu sintezu sistema automatskog upravljanja reaktora GCR tipa

    Energy Technology Data Exchange (ETDEWEB)

    Lazarevic, B; Matausek, M [Institut za nuklearne nauke ' Boris Kidric' , Vinca, Belgrade (Yugoslavia)

    1966-07-01

    This paper describes the model which was developed for synthesis of spatial distribution of automated control elements in the reactor. It represents a general reliable mathematical model for analyzing transition states and synthesis of the automated control and regulation systems of GCR type reactors. One-dimensional system was defined under assumption that the time dependence of parameters of the neutron diffusion equation are identical in the total volume of the reactor and that spatial distribution of neutrons is time independent. It is shown that this assumption is satisfactory in case of short term variations which are relevant for safety analysis.

  14. A distributed hierarchical architecture of expert systems for supervisory control of multimodular nuclear reactors

    International Nuclear Information System (INIS)

    Otaduy, P.J.; Brittain, C.R.; Rovere, L.A.; Gove, N.B.

    1991-01-01

    A hierarchical supervisory control architecture has being implemented at ORNL to coordinate the controllers of a multimodular nuclear plant. The supervisory controller form a network of distributed expert system interfaced with a real-time simulation of the plant, the plant's automatic controllers, and the human operator. The main goal of the supervisory controllers is to maintain the plant operating within safety envelopes while optimizing availability, minimizing stress to components and operators, and facilitating operations. Representative rules implementing strategies for situation dependent reassignment of process goals by embedding diagnostics into the control philosophy are discussed. It should noted that the control philosophies here described use the ALMR concept for illustration purposes and are not part of the official ALMR design at this time. 3 refs., 1 fig

  15. The symbol coding language for the BUTs processor of in-core reactor control systems

    International Nuclear Information System (INIS)

    Vorob'ev, D.M.; Golovanov, M.N.; Levin, G.L.; Parfenova, T.K.; Filatov, V.P.

    1978-01-01

    A symbolic coding language is described; it has been developed for automation of making up programs for in-core control systems. The systems use the ideology of the CAMAC-VECTOR system and include the BUTs-20 processor. The symbolic coding language has been developed as a programming language of the ASSEMBLER type. Operators of instructions and pseudo-instructions, the rules of reading in the text of the source program, and operator record formats are considered

  16. The hierarchy of essential CANDU reactor control functions in a distributed system

    International Nuclear Information System (INIS)

    Mercier, P.

    1980-01-01

    Control functions in CANDU nuclear generating stations are programmed within two centralized and redundant minicomputers while safety functions are covered by conventional analog systems. This set-up is a product of standards, economic and technical considerations which are now being modified by the maturing of microprocessors, the progress in digital communications and the development of mathematical process models. Starting from the control and safety systems installed in Gentilly-2, this paper analyses trends that will affect the implementation of essential control functions within a distributed system. In particular, it emphasizes the characteristics of future software systems that must be built-in in order to comply with important operational requirements of nuclear generating stations. (auth) [fr

  17. Nuclear reactor control apparatus

    International Nuclear Information System (INIS)

    Sridhar, B.N.

    1983-01-01

    Nuclear reactor safety rod release apparatus comprises a ring which carries detents normally positioned in an annular recess in outer side of the rod, the ring being held against the lower end of a drive shaft by magnetic force exerted by a solenoid carried by the drive shaft. When the solenoid is de-energized, the detent-carrying ring drops until the detents contact a cam surface associated with the lower end of the drive shaft, at which point the detents are cammed out of the recess in the safety rod to release the rod from the drive shaft. In preferred embodiments of the invention, an additional latch is provided to release a lower portion of a safety rod under conditions that may interfere with movement of the entire rod

  18. Reactor safety systems

    International Nuclear Information System (INIS)

    Kafka, P.

    1975-01-01

    The spectrum of possible accidents may become characterized by the 'maximum credible accident', which will/will not happen. Similary, the performance of safety systems in a multitude of situations is sometimes simplified to 'the emergency system will/will not work' or even 'reactors are/ are not safe'. In assessing safety, one must avoid this fallacy of reducing a complicated situation to the simple black-and-white picture of yes/no. Similarly, there is a natural tendency continually to improve the safety of a system to assure that it is 'safe enough'. Any system can be made safer and there is usually some additional cost. It is important to balance the increased safety against the increased costs. (orig.) [de

  19. Systems for controlling the electric power of a boiling water reactor power station

    International Nuclear Information System (INIS)

    Fukunishi, Koyu; Kiyokawa, Kazuhiro.

    1975-01-01

    Object: To achieve automatic increase and decrease of electric output in accordance with a predetermined rate of increase or decrease in output when the power output is raised or lowered. Structure: An electric output signal from an atomic power plant is led to a differentiating circuit through a smoothing circuit to produce a signal for rate of change of time, and an error signal between this signal and a preset signal produced from a circuit for a preset rate of change of output with time is supplied to an analog adjuster through a limiter. In this way, the flow rate in the reactor core is adjusted by a speed controller to obtain an output of a predetermined rate of increase. The difference signal between the electric output signal and a desired value signal is passed through an absolute circuit to a comparator circuit for comparison with a predetermined threshold value setting signal. The output signal of the comparator is used to operate a relay to open the contact so as to prevent an increase or decrease in the output beyond the required level. (Kamimura, M.)

  20. Core cooling system for reactor

    International Nuclear Information System (INIS)

    Kondo, Ryoichi; Amada, Tatsuo.

    1976-01-01

    Purpose: To improve the function of residual heat dissipation from the reactor core in case of emergency by providing a secondary cooling system flow channel, through which fluid having been subjected to heat exchange with the fluid flowing in a primary cooling system flow channel flows, with a core residual heat removal system in parallel with a main cooling system provided with a steam generator. Constitution: Heat generated in the core during normal reactor operation is transferred from a primary cooling system flow channel to a secondary cooling system flow channel through a main heat exchanger and then transferred through a steam generator to a water-steam system flow channel. In the event if removal of heat from the core by the main cooling system becomes impossible due to such cause as breakage of the duct line of the primary cooling system flow channel or a trouble in a primary cooling system pump, a flow control valve is opened, and steam generator inlet and outlet valves are closed, thus increasing the flow rate in the core residual heat removal system. Thereafter, a blower is started to cause dissipation of the core residual heat from the flow channel of a system for heat dissipation to atmosphere. (Seki, T.)

  1. A study on the fusion reactor - Design study for tokamak control system

    Energy Technology Data Exchange (ETDEWEB)

    Ko, In Soo; Nam, Kung Won; Cho, Moo Hyun; Kim, Ji Hwa; Kim, Jae Myung; Kim, Sung Chul; Lee, Ki Sun [Pohang University of Science and Technology, Pohang (Korea, Republic of)

    1996-09-01

    Based on the experience accumulated during the construction of Pohang Light Source, we strongly suggest that the control system must have a hierarchical structure linked with multiple ethernets. Some UNIX-based workstation are recommended in the top layer where KSTAR operators are faced with several control/monitoring windows. As a supervisors to individual subsystem, realtime computers such as VME system are recommended. Choice of realtime operating system is up to the actual developer. However, we strongly suggest to use Lynx which is UNIX-based realtime OS and is getting more attentions after its introduction in the market. Graphic interface for operators will be developed by adopting development toolkit such as RTWorks and EPICS. We also recommend to use EPICS because it is introduced by Los Alamos National Laboratory, U. S. A. and several laboratories around world are now adopting it to their control system. This report is also pointing out several key issues relating the system development. Among them are the cabling plan, signal list, and noise suppressions. It is also suggested that early decision of communication protocol will help to integrated control system in the future. 11 refs., 5 tabs., 4 figs. (author)

  2. A prototype expert system 'SMART' for water chemistry control in reactor water circuits

    International Nuclear Information System (INIS)

    Rangarajan, S.; Narasimhan, S.V.

    1998-01-01

    The operational safety of a power plant depends mainly on the material compatibility of the system materials with the environment. However, for an operating plant, the material is almost fixed and hence one can improve the safety by controlling the surrounding environment. From the economy point of view, the plant availability factor as well as plant life extension (PLEX) are important considerations and these necessitate a systematic approach for continuous parametric monitoring, rapid data analysis and diagnosis for controlling the water chemistry regime. A prototype expert system 'SMART' was developed in BASIC language. The expert system consists of four modules. The DATA HANDLER module controls all the data handling functions and graphical display of the data parameters. It also generates weekly and monthly reports of the water chemistry data. The DATA INTERPRETER module compares the experimental data with the theoretically calculated values and predicts the presence of impurity ingress in the system. The CHEMISTRY EXPERT contains the knowledge base about the various sub-systems. All the water chemistry specifications are translated in the form of IF... THEN.. rules and are stored in this module. The expert system inferences with the forward chain reasoning mechanism to identify the diagnostic parameters by consulting the knowledge base and applying the appropriate rules. The ACTION EXPERT module collects all the diagnostic parameters and suggests the operator, the remedial actions/counter measures that should be taken immediately. This rule based system can be expanded to accommodate different water chemistry regimes. (author)

  3. 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.

  4. Special power supply and control system for the gas-cooled fast reactor-core flow test loop

    International Nuclear Information System (INIS)

    Hudson, T.L.

    1981-09-01

    The test bundle in the Gas-Cooled Fast Reactor-Core Flow Test Loop (GCFR-CFTL) requires a source of electrical power that can be controlled accurately and reliably over a wide range of steady-state and transient power levels and skewed power distributions to simulate GCFR operating conditions. Both ac and dc power systems were studied, and only those employing silicon-controlled rectifiers (SCRs) could meet the requirements. This report summarizes the studies, tests, evaluations, and development work leading to the selection. it also presents the design, procurement, testing, and evaluation of the first 500-kVa LMPL supply. The results show that the LMPL can control 60-Hz sine wave power from 200 W to 500 kVA

  5. Reactor cooling system

    International Nuclear Information System (INIS)

    Kato, Etsuji.

    1979-01-01

    Purpose: To eliminate cleaning steps in the pipelines upon reactor shut-down by connecting a filtrating and desalting device to the cooling system to thereby always clean up the water in the pipelines. Constitution: A filtrating and desalting device is connected to the pipelines in the cooling system by way of drain valves and a check valve. Desalted water is taken out from the exit of the filtrating and desalting device and injected to one end of the cooling system pipelines by way of the drain valve and the check valve and then returned by way of another drain valve to the desalting device. Water in the pipelines is thus always desalted and the cleaning step in the pipelines is no more required in the shut-down. (Kawakami, Y.)

  6. Power control device in nuclear reactor

    International Nuclear Information System (INIS)

    Koyama, Kazuaki.

    1981-01-01

    Purpose: To enable smooth power changes in power conditioning systems by calculating forecast values for the neutron flux distribution and power distribution and by controlling the driving speed of control rods so as to correspond the forecast values with aimed values. Constitution: Control rod position is detected by a position detector and sent to a control computer as the position information. At the same time, the neutron flux distribution information is obtained by the neutron monitors, the power distribution information is obtained by a reactor power computer and they are outputted to the control computer. The control computer calculates the forecast values for the neutron flux distribution and the reactor power distribution from the information, and compares them with the aimed values from a setter and then outputs control signals so as to correspond the forecast values with the aimed values. The control rods can be inserted in appropriate velocity by the control signals. (Horiuchi, T.)

  7. Method of controlling the reactor operation

    International Nuclear Information System (INIS)

    Ishiguro, Akira; Nakakura, Hiroyuki.

    1987-01-01

    Purpose: To moderate vibratory response due to delayed operation thereby obtain stable controlled response in the operation control for a PWR type reactor. Method: the reactor operation is controlled by the axial power distribution control by regulating the boron concentration in primary coolants with a boron density control system and controlling the average temperature for the primary coolants with the control rod control system. In this case, the control operation and the control response become instable due to transmission delay, etc. of aqueous boric acid injection to the primary coolant circuits to result in vibratory response. In the present invention, signals are prepared by adding the amount in proportion to the variation coefficient with time of xenone concentration obtained from the measured value for the reactor power added to the conventional axial power distribution parameter deviation and used as the input signals for the boron concentration control system. As a result, the instability due to the transmission delay of the aqueous boric acid injection is improved by the preceding control by the amount in proportion with the variation coefficient with time of the xenone concentration. An advantageous effect can be expected for the load following operation during day time according to the present invention. (Kamimura, M.)

  8. Design and construction of a preamplifier for research reactor control system using Russia’s neutron detectors

    International Nuclear Information System (INIS)

    Trinh Dinh Hai; Vo Van Tai; Le Van Diep; Nguyen Nhi Dien

    2016-01-01

    This paper presents the design and construction of a preamplifier device for Research Reactor Control System, using Russia’s Neutron Detectors of ionization and fission chambers. In this work, the preamplifier device which consists of a wide range Current to Frequency Converter block used with a compensation ionization chamber type KNK-3 to measure the thermal neutron flux in the range of 1x10"6 - 1x10"1"1 n/cm"2.s, a Pulse Preamplifier block used with a fission chamber type KNK-15 to measure the thermal neutron flux in the range of 1x10"0 - 1x10"6 n/cm"2.s, and a Power Supply block, was designed and tested in different conditions in the laboratory and at Dalat Nuclear Research Reactor (DNRR). Obtained results show that, the above blocks have almost design specifications as equivalent or better in comparison with the same function blocks of the DNRR Control System which were designed by the former Soviet Union. They also meet the utilization requirements as well as the experimental and training purposes. (author)

  9. Nuclear reactor safety system

    International Nuclear Information System (INIS)

    Sato, Takashi.

    1979-01-01

    Purpose: To allow sufficient removal of radioactive substance released in the reactor containment shell upon loss of coolants accidents thus to sufficiently decrease the exposure dose to human body. Constitution: A clean-up system is provided downstream of a heat exchanger and it is branched into a pipeway to be connected to a spray nozzle and further connected by way of a valve to a reactor container. After the end of sudden transient changes upon loss of coolants accidents, the pool water stored in the pressure suppression chamber is purified in the clean-up system and then sprayed in the dry-well by way of a spray nozzle. The sprayed water dissolves to remove water soluble radioactive substances floating in the dry-well and then returns to the pressure suppression chamber. Since radioactive substances in the dry-well can thus removed rapidly and effectively and the pool water can be reused, public hazard can also be decreased. (Horiuchi, T.)

  10. Reactor-core-reactivity control device

    International Nuclear Information System (INIS)

    Miura, Teruo; Sakuranaga, Tomonobu.

    1983-01-01

    Purpose: To improve the reactor safety upon failures of control rod drives by adapting a control rod not to drop out accidentally from the reactor core but be inserted into the reactor core. Constitution: The control rod is entered or extracted as usual from the bottom of the pressure vessel. A space is provided above the reactor core within the pressure vessel, in which the moving scope of the control rod is set between the space above the reactor core and the reactor core. That is, the control rod is situated above the reactor core upon extraction thereof and, if an accident occurs to the control rod drive mechanisms to detach the control rod and the driving rod, the control rod falls gravitationally into the reactor core to improve the reactor safety. In addition, since the speed limiter is no more required to the control rod, the driving force can be decreased to reduce the size of the rod drive mechanisms. (Ikeda, J.)

  11. Development and characterization of the control assembly system for the large 2400 MWth Generation IV gas-cooled fast reactor

    International Nuclear Information System (INIS)

    Girardin, G.; Rimpault, G.; Morin, F.; Bosq, J.C.; Coddington, P.; Mikityuk, K.; Chawla, R.

    2008-01-01

    The present paper is related to the design and neutronic characterization of the principal control assembly system for the reference large (2400 MWth) Generation IV gas-cooled fast reactor (GFR), which makes use of ceramic-ceramic (CERCER) plate-type fuel-elements with (U-Pu) carbide fuel contained within a SiC inert matrix. For the neutronic 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 assemblies as a starting point. More specifically, the core contains a total of 33 control (control system device: CSD) and safety (diverse safety device: DSD) assemblies implemented in three banks. In the design of the new control assembly system, particular attention was given to the heat generation within the assemblies, so that both neutronic and thermal-hydraulic constraints could be appropriately accounted for. The thermal-hydraulic calculations have been performed with the code COPERNIC, significant coolant mass flow rates being found necessary to maintain acceptable cladding temperatures of the absorber pins. Complementary to the design study, neutronic investigations have been performed to assess the impact of the control assemblies in the GFR core in greater detail (rod interactions, shift of the flux, peaking factors, etc.). Thus, considerable shadowing effects have been observed between the first bank and the safety bank, as also between individual assemblies within the first bank. Large anti-shadowing effects also occur, the most prominent being that between the two CSD banks, where the total assembly worth is almost doubled in comparison to the sum of the individual values. Additional investigations have been performed and, in this context, it has been found that computation of the first-order eigenvalue and the eigenvalue separation is a robust tool to anticipate control assembly interactions in a large fast-spectrum core. One interesting

  12. Qualification of process control system controbloc P20 developed for N4-type reactors

    International Nuclear Information System (INIS)

    Dauce, F.; De Tricornot, J.; Delmaire, P.H.; Delmas, G.; Marc, J.L.

    1989-01-01

    The essential mission of the CONTROBLOC P20 system in the control and instrumentation architecture of the N4 plant is to achieve automatic control and modulating functions and to prepare information for the control room or to receive orders from it. Its architecture has been designed to allow geographical decentralization of the interface cards, placing them as near as possible to the producer or consumer of the process data. The CONTROBLOC P20 system is a combination of functional subsystems (clusters) which cooperate each other according to deterministic mechanisms quaranteeing performance. Each cluster is thus a federation of blocks (subscribers) which process and exchange information according to a predetermined traffic arrangement. To qualify can be summed up as checking that the strict relations between information consumer and producer blocks are complied with for different specified environmental constraints. This particularly severe environment requires the acknowledgement of particular design precautions, especially because of new technologies implemented

  13. Recent development of nuclear power in Japan and instrumentation and control system and control room equipment for advanced light water reactors

    International Nuclear Information System (INIS)

    Wakayama, N.

    1992-01-01

    This paper was provided for the 13th IAEA/IWG-NPPCI Meeting and aims to introduce an outline of recent development of nuclear power in Japan and some topics in the field of nuclear power plant control and instrumentation. Forty units of nuclear power plants are in operation in Japan and five units of BWRs and six PWRs are under construction. Construction of prototype FBR Monju have almost completed an construction of High-Temperature Engineering Test Reactor, HTTR, started in March 1991. In parallel of those, extensive effort has been carried out to develop the third generation LWRs which are called Advanced BWR (ABWR) and Advanced PWR (APWR). Two Advanced BWRs are under safety review for construction. Instrumentation and control system of these Advanced LWRs adopts integrated digital I and C system, optical multiplexing signal transmission, fault tolerant control systems and software logic for reactor protection and safety systems and enhances plant control performance and provides human-friendly operation and maintenance environments. Main control room of these Advanced LWRs, comprised with large display panels and advanced console, has special futures such as one-man sit-down operation, human friendly man-machine interface, high level automation in operation and maintenance. (author). 7 refs, 9 figs, 1 tab

  14. Joint System Prognostics For Increased Efficiency And Risk Mitigation In Advanced Nuclear Reactor Instrumentation and Control

    Energy Technology Data Exchange (ETDEWEB)

    Donald D. Dudenhoeffer; Tuan Q. Tran; Ronald L. Boring; Bruce P. Hallbert

    2006-08-01

    The science of prognostics is analogous to a doctor who, based on a set of symptoms and patient tests, assesses a probable cause, the risk to the patient, and a course of action for recovery. While traditional prognostics research has focused on the aspect of hydraulic and mechanical systems and associated failures, this project will take a joint view in focusing not only on the digital I&C aspect of reliability and risk, but also on the risks associated with the human element. Model development will not only include an approximation of the control system physical degradation but also on human performance degradation. Thus the goal of the prognostic system is to evaluate control room operation; to identify and potentially take action when performance degradation reduces plant efficiency, reliability or safety.

  15. 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.)

  16. The control of reactor outages

    International Nuclear Information System (INIS)

    Bouget, Y.H.; Berteloot, J.M.

    1995-01-01

    The 1985-1992 period was marked by a continuous decay in French reactors operation. This situation has led the Committee for Outages Mastery to take steps for the improvement of nuclear power plants availability. The control of reactor outages requires an integrated vision of the safety, duration, dosimetry, costs and security aspects and a perfect management of contractors. The paper describes the methodology used for the management and the maintenance of the French PWR reactors stock. A detailed schedule of maintenance tasks with dose estimations is now required from each site to anticipate and optimize the duration of outages. Thanks to this action, a significant reduction of the maintenance costs is observed for the 1992-1995 period. (J.S.). 2 figs

  17. 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.)

  18. Implementation of a model reference adaptive control system using neural network to control a fast breeder reactor evaporator

    International Nuclear Information System (INIS)

    Ugolini, D.; Yoshikawa, S.; Endou, A.

    1994-01-01

    Artificial intelligence is foreseen as the base for new control systems aimed to replace traditional controllers and to assist and eventually advise plant operators. This paper discusses the development of an indirect model reference adaptive control (MRAC) system, using the artificial neural network (ANN) technique, and its implementation to control the outlet steam temperature of a sodium to water evaporator. The ANN technique is applied in the identification and in the control process of the indirect MRAC system. The emphasis is placed on demonstrating the efficacy of the indirect MRAC system in controlling the outlet steam temperature of the evaporator, and on showing the important function covered by the ANN technique. An important characteristic of this control system is that it relays only on some selected input variables and output variables of the evaporator model. These are the variables that can be actually measured or calculated in a real environment. The results obtained applying the indirect MRAC system to control the evaporator model are quite remarkable. The outlet temperature of the steam is almost perfectly kept close to its desired set point, when the evaporator is forced to depart from steady state conditions, either due to the variation of some input variables or due to the alteration of some of its internal parameters. The results also show the importance of the role played by the ANN technique in the overall control action. The connecting weights of the ANN nodes self adjust to follow the modifications which may occur in the characteristic of the evaporator model during a transient. The efficiency and the accuracy of the control action highly depends on the on-line identification process of the ANN, which is responsible for upgrading the connecting weights of the ANN nodes. (J.P.N.)

  19. Instrumentation and control strategies for an integral pressurized water reactor

    Directory of Open Access Journals (Sweden)

    Belle R. Upadhyaya

    2015-03-01

    Full Text Available Several vendors have recently been actively pursuing the development of integral pressurized water reactors (iPWRs that range in power levels from small to large reactors. Integral reactors have the features of minimum vessel penetrations, passive heat removal after reactor shutdown, and modular construction that allow fast plant integration and a secure fuel cycle. The features of an integral reactor limit the options for placing control and safety system instruments. The development of instrumentation and control (I&C strategies for a large 1,000 MWe iPWR is described. Reactor system modeling—which includes reactor core dynamics, primary heat exchanger, and the steam flashing drum—is an important part of I&C development and validation, and thereby consolidates the overall implementation for a large iPWR. The results of simulation models, control development, and instrumentation features illustrate the systematic approach that is applicable to integral light water reactors.

  20. Development of a software for the control of the quality management system of the TRIGA-Mark III reactor

    International Nuclear Information System (INIS)

    Herrera A, E.; Hernandez, L.V.; Hernandez, J.A.

    2006-01-01

    The quality has not only become one of the essential requirements of the product but rather at the present time it is a strategic factor key of which depends the bigger part of the organizations, not only to maintain their position in the market but also to assure their survival. The good organizations will have processes, procedures and standards to confront these challenges. The big organizations require of the certification of their administration systems, and once the organization has obtained this certification the following step it is to maintain it. The implementation and certification of an administration system requires of an appropriate operative organization that achieves continuous improvements in their operation. This is the case of the TRIGA Mark III reactor, which contains a computer program that upgrades, it controls and it programs activities to develop in the Installation, allowing one operative organization to the whole personnel of the same one. With the purpose of avoiding activities untimely. (Author)

  1. An autonomous control framework for advanced reactors

    Directory of Open Access Journals (Sweden)

    Richard T. Wood

    2017-08-01

    Full Text Available Several Generation IV nuclear reactor concepts have goals for optimizing investment recovery through phased introduction of multiple units on a common site with shared facilities and/or reconfigurable energy conversion systems. Additionally, small modular reactors are suitable for remote deployment to support highly localized microgrids in isolated, underdeveloped regions. The long-term economic viability of these advanced reactor plants depends on significant reductions in plant operations and maintenance costs. To accomplish these goals, intelligent control and diagnostic capabilities are needed to provide nearly autonomous operations with anticipatory maintenance. A nearly autonomous control system should enable automatic operation of a nuclear power plant while adapting to equipment faults and other upsets. It needs to have many intelligent capabilities, such as diagnosis, simulation, analysis, planning, reconfigurability, self-validation, and decision. These capabilities have been the subject of research for many years, but an autonomous control system for nuclear power generation remains as-yet an unrealized goal. This article describes a functional framework for intelligent, autonomous control that can facilitate the integration of control, diagnostic, and decision-making capabilities to satisfy the operational and performance goals of power plants based on multimodular advanced reactors.

  2. An autonomous control framework for advanced reactors

    International Nuclear Information System (INIS)

    Wood, Richard T.; Upadhyaya, Belle R.; Floyd, Dan C.

    2017-01-01

    Several Generation IV nuclear reactor concepts have goals for optimizing investment recovery through phased introduction of multiple units on a common site with shared facilities and/or reconfigurable energy conversion systems. Additionally, small modular reactors are suitable for remote deployment to support highly localized microgrids in isolated, underdeveloped regions. The long-term economic viability of these advanced reactor plants depends on significant reductions in plant operations and maintenance costs. To accomplish these goals, intelligent control and diagnostic capabilities are needed to provide nearly autonomous operations with anticipatory maintenance. A nearly autonomous control system should enable automatic operation of a nuclear power plant while adapting to equipment faults and other upsets. It needs to have many intelligent capabilities, such as diagnosis, simulation, analysis, planning, reconfigurability, self-validation, and decision. These capabilities have been the subject of research for many years, but an autonomous control system for nuclear power generation remains as-yet an unrealized goal. This article describes a functional framework for intelligent, autonomous control that can facilitate the integration of control, diagnostic, and decision-making capabilities to satisfy the operational and performance goals of power plants based on multimodular advanced reactors

  3. An autonomous control framework for advanced reactors

    Energy Technology Data Exchange (ETDEWEB)

    Wood, Richard T.; Upadhyaya, Belle R.; Floyd, Dan C. [Dept. of Nuclear Engineering, University of Tennessee, Knoxville (United States)

    2017-08-15

    Several Generation IV nuclear reactor concepts have goals for optimizing investment recovery through phased introduction of multiple units on a common site with shared facilities and/or reconfigurable energy conversion systems. Additionally, small modular reactors are suitable for remote deployment to support highly localized microgrids in isolated, underdeveloped regions. The long-term economic viability of these advanced reactor plants depends on significant reductions in plant operations and maintenance costs. To accomplish these goals, intelligent control and diagnostic capabilities are needed to provide nearly autonomous operations with anticipatory maintenance. A nearly autonomous control system should enable automatic operation of a nuclear power plant while adapting to equipment faults and other upsets. It needs to have many intelligent capabilities, such as diagnosis, simulation, analysis, planning, reconfigurability, self-validation, and decision. These capabilities have been the subject of research for many years, but an autonomous control system for nuclear power generation remains as-yet an unrealized goal. This article describes a functional framework for intelligent, autonomous control that can facilitate the integration of control, diagnostic, and decision-making capabilities to satisfy the operational and performance goals of power plants based on multimodular advanced reactors.

  4. Controlled thermonuclear fusion reactors

    International Nuclear Information System (INIS)

    Walstrom, P.L.

    1976-01-01

    Controlled production of energy by fusion of light nuclei has been the goal of a large portion of the physics community since the 1950's. In order for a fusion reaction to take place, the fuel must be heated to a temperature of 100 million degrees Celsius. At this temperature, matter can exist only in the form of an almost fully ionized plasma. In order for the reaction to produce net power, the product of the density and energy confinement time must exceed a minimum value of 10 20 sec m -3 , the so-called Lawson criterion. Basically, two approaches are being taken to meet this criterion: inertial confinement and magnetic confinement. Inertial confinement is the basis of the laser fusion approach; a fuel pellet is imploded by intense laser beams from all sides and ignites. Magnetic confinement devices, which exist in a variety of geometries, rely upon electromagnetic forces on the charged particles of the plasma to keep the hot plasma from expanding. Of these devices, the most encouraging results have been achieved with a class of devices known as tokamaks. Recent successes with these devices have given plasma physicists confidence that scientific feasibility will be demonstrated in the next generation of tokamaks; however, an even larger effort will be required to make fusion power commercially feasible. As a result, emphasis in the controlled thermonuclear research program is beginning to shift from plasma physics to a new branch of nuclear engineering which can be called fusion engineering, in which instrumentation and control engineers will play a major role. Among the new problem areas they will deal with are plasma diagnostics and superconducting coil instrumentation

  5. Development of process control capability through the Browns Ferry Integrated Computer System using Reactor Water Clanup System as an example. Final report

    International Nuclear Information System (INIS)

    Smith, J.; Mowrey, J.

    1995-12-01

    This report describes the design, development and testing of process controls for selected system operations in the Browns Ferry Nuclear Plant (BFNP) Reactor Water Cleanup System (RWCU) using a Computer Simulation Platform which simulates the RWCU System and the BFNP Integrated Computer System (ICS). This system was designed to demonstrate the feasibility of the soft control (video touch screen) of nuclear plant systems through an operator console. The BFNP Integrated Computer System, which has recently. been installed at BFNP Unit 2, was simulated to allow for operator control functions of the modeled RWCU system. The BFNP Unit 2 RWCU system was simulated using the RELAP5 Thermal/Hydraulic Simulation Model, which provided the steady-state and transient RWCU process variables and simulated the response of the system to control system inputs. Descriptions of the hardware and software developed are also included in this report. The testing and acceptance program and results are also detailed in this report. A discussion of potential installation of an actual RWCU process control system in BFNP Unit 2 is included. Finally, this report contains a section on industry issues associated with installation of process control systems in nuclear power plants

  6. Experimental development of power reactor intelligent control

    International Nuclear Information System (INIS)

    Edwards, R.M.; Garcia, H.E.; Lee, K.Y.

    1992-01-01

    The US nuclear utility industry initiated an ambitious program to modernize the control systems at a minimum of ten existing nuclear power plants by the year 2000. That program addresses urgent needs to replace obsolete instrumentation and analog controls with highly reliable state-of-the-art computer-based digital systems. Large increases in functionality that could theoretically be achieved in a distributed digital control system are not an initial priority in the industry program but could be logically considered in later phases. This paper discusses the initial development of an experimental sequence for developing, testing, and verifying intelligent fault-accommodating control for commercial nuclear power plant application. The sequence includes an ultra-safe university research reactor (TRIGA) and a passively safe experimental power plant (Experimental Breeder Reactor 2)

  7. Application of a fluidized bed reactor charged with aragonite for control of alkalinity, pH and carbon dioxide in marine recirculating aquaculture systems

    Science.gov (United States)

    Paul S Wills, PhD; Pfeiffer, Timothy; Baptiste, Richard; Watten, Barnaby J.

    2016-01-01

    Control of alkalinity, dissolved carbon dioxide (dCO2), and pH are critical in marine recirculating aquaculture systems (RAS) in order to maintain health and maximize growth. A small-scale prototype aragonite sand filled fluidized bed reactor was tested under varying conditions of alkalinity and dCO2 to develop and model the response of dCO2 across the reactor. A large-scale reactor was then incorporated into an operating marine recirculating aquaculture system to observe the reactor as the system moved toward equilibrium. The relationship between alkalinity dCO2, and pH across the reactor are described by multiple regression equations. The change in dCO2 across the small-scale reactor indicated a strong likelihood that an equilibrium alkalinity would be maintained by using a fluidized bed aragonite reactor. The large-scale reactor verified this observation and established equilibrium at an alkalinity of approximately 135 mg/L as CaCO3, dCO2 of 9 mg/L, and a pH of 7.0 within 4 days that was stable during a 14 day test period. The fluidized bed aragonite reactor has the potential to simplify alkalinity and pH control, and aid in dCO2 control in RAS design and operation. Aragonite sand, purchased in bulk, is less expensive than sodium bicarbonate and could reduce overall operating production costs.

  8. Support to the elaboration of the engineering of detail, configuration and programming of the control system of heat removal of the TRIGA Mark III reactor

    International Nuclear Information System (INIS)

    Diaz G, C. A.

    2016-01-01

    Nowadays, the peaceful and responsible use of nuclear energy in Mexico is of great importance and contributes to economic, social, scientist and technologic development in the country, highlighting the Instituto Nacional de Investigaciones Nucleares (ININ) and the Nuclear Power Plant of Laguna Verde as one of the most important dependences. Among the main facilities and laboratories of ININ is the Nuclear Research Reactor TRIGA Mark III, this is a pool type reactor with mobile core, cooled and moderated by light water and a flow of 1013 n/cm"2/sec. Due to the technological obsolescence is a growing problem that threatens the information, operation and/or efficacy of elements of control and safety systems of the reactor, these must be changed each time more frequently. In the modernization of reactor was used a Modicon M340 programmable logic control (PLC) and a Twido PLC for the control of heat removal system (Primary Cooling System (PCS) and Secondary Cooling System (SCS) respectively), this because the PLC has proven to be safe and effective devices, addition to reduce the wiring elements and increase the possibilities of performance and design of the digital control console. This document shows and describes the elements of heat removal system (PCS and SCS), and the signals and signal types that such items send or received by the PLC, likewise, is indicated the methodology used to develop the applications for the control of the Primary Cooling System and Secondary Cooling System, beginning with the PLC design, the development of PLC plans and the control logic, and finally, the simulation and debugging of applications on Unity Pro and Twido Suite. All this in compliance with the safety standards to nuclear research reactors (NS-R-4), the rules of industrial programming (IEC 61131-3), and the reactor operating limits postulated in the safety report and the software assurance system used in the ININ. (Author)

  9. Reactor feedwater system

    International Nuclear Information System (INIS)

    Hikabe, Katsumi.

    1978-01-01

    Purpose: In order to prevent thermal stresses of a core of PWR type reactor, described has been a method for feeding heated recirculating water to the core in the case of the reactor start-up or shut-down. Constitution: A recirculating water is degassed, cleaned up and heated in the steam condensers, and then feeds the water to the reactor, characterized in that heaters are provided in the bypasses of the turbine, so that heated water is constantly supplied to the reactor. (Nakamura, S.)

  10. Application of expert systems to heat exchanger control at the 100-megawatt high-flux isotope reactor

    International Nuclear Information System (INIS)

    Clapp, N.E. Jr.; Clark, F.H.; Mullens, J.A.; Otaduy, P.J.; Wehe, D.K.

    1985-01-01

    The High-Flux Isotope Reactor (HFIR) is a 100-MW pressurized water reactor at the Oak Ridge National Laboratory. It is used to produce isotopes and as a source of high neutron flux for research. Three heat exchangers are used to remove heat from the reactor to the cooling towers. A fourth heat exchanger is available as a spare in case one of the operating heat exchangers malfunctions. It is desirable to maintain the reactor at full power while replacing the failed heat exchanger with the spare. The existing procedures used by the operators form the initial knowledge base for design of an expert system to perform the switchover. To verify performance of the expert system, a dynamic simulation of the system was developed in the MACLISP programming language. 2 refs., 3 figs

  11. Substitution of the soluble boron reactivity control system of a pressurized water reactor by gadolinium burnable poisons

    International Nuclear Information System (INIS)

    Galperin, A.; Segev, M.; Radkowsky, A.

    1986-01-01

    The results are presented of a research project that is aimed at designing a gadolinium burnable poison (BP) system for complete reactivity control of a pressurized water reactor (PWR) core during the ''equilibrium'' cycle, resulting in the elimination of the soluble boron system, which represents a considerable saving in both capital and operating costs. A flat and strong negative moderator temperature coefficient is assured for a poison-free moderator. The design analysis of a core, heavily loaded with gadolinium BP rods, was based on a BGUCORE neutronic package and cluster model of a fuel assembly. The project objective was achieved by a novel lumped BP rod, designed as an annulus of gadolinium, clad by zirconium, and inserted into vacant guide thimbles of fresh-fuel assemblies. Specific combinations were found for the inner/outer radii of the poison ring, gadolinium densities, and number of rods per assembly, resulting in an almost flat criticality curve during the cycle. A reactivity swing of ≅1% ΔK can be easily controlled by an existing system of control rods. Comparison of the fuel cycle length of a gadolinium-controlled core with that of the reference, soluble, boron-controlled core indicated that there is no penalty due to residual poison at end of life. Unique guidelines for the fuel loading strategy were applied to find a practical fuel-shuffling scheme by which the design and operational constraints of a typical PWR core of current design were satisfied. Several problems should be solved for a practical implementation of the presented design relative to operational and safety requirements of the existing control rod system. Adequate movement of the regulating rods should be determined and shutdown margins of the safety rods should be ascertained. Final judgment of the feasibility of the concept may be made following the solution of these and other regulatory-related issues

  12. Monitor for reactor feedwater systems

    International Nuclear Information System (INIS)

    Takizawa, Yoji; Tomizawa, Teruaki

    1983-01-01

    Purpose: To improve the reliability of operator's procedures upon occurrence of the feedwater system abnormality in a BWR type reactor by presenting the operation with effective information to avoid such abnormality. Constitution: A feedwater temperature at the reactor inlet of a reactor feedwater system measured by a temperature detector and a predetermined value for the feedwater temperature at the reactor inlet determined depending on the reactor conditions are inputted to a start-up system. The start-up system outputs a start-up signal when the difference between the inputted values exceeds a predetermined value. Then, the start-up signal is inputted to a display device where information required for the operator is displayed in the device. Thus, the information required for the operator is rapidly provided upon abnormality of the feedwater system to thereby improve the reliability of the operator's procedures. (Moriyama, K.)

  13. Proceedings of workshop on reactor shutdown system

    International Nuclear Information System (INIS)

    1997-03-01

    India has gained considerable experience in design, development, construction and operation of research and power reactors during the last four decades. Reactor shutdown system (RSS) is the most important engineered safety system of any reactor. A lot of technological developments have taken place to improve the reactor shutdown systems, particularly with advancement in reliability analysis and instrumentation and control. If the reactor is not shutdown, the fuel may melt, releasing radioactivity and possibly reactivity addition as in the case of Fast Breeder Reactor (FBR). Apart from radiological safety consequences, large investment has to be written off. The function of the RSS is to stop fission chain reaction and prevent breach of fuel. The design of RSS is multidisciplinary. It requires reactor physics analysis, design of absorber rods, drive mechanisms, safety logic to order shutdown and instrumentation to detect unsafe conditions. High reliability is essential and this requires two independent shutdown systems. This book contains the proceedings of the workshop on reactor shutdown system and papers relevant to INIS are indexed separately

  14. Contributions to the research programs in nuclear and industrial electronics, domestic production of instrumentation, safety and control systems and equipment for nuclear reactors and auxiliary installations

    International Nuclear Information System (INIS)

    Talpariu, C; Talpariu, J.; Matei, C.

    2001-01-01

    Domestic production of component system and equipment for the control and safety of nuclear facilities was one of the priority objective of the Nuclear Research Institute Pitesti. The problems addressed were particularly related to design and production of analog and digital equipment for measurements, triggering and display of the values of process parameters as well as to regulating complex functions of this equipment. Associated to this effort were the research works concerning: - reliability and in-service life-time of the electronic components and equipment in the safety and control systems for nuclear processes; - radiation endurance of industrial electronic components; utilization of whirling currents in calandria tube testing; - expert systems and applications in nuclear reactor control and safety; design and testing methods of process real time software packages for safety in control critical systems for nuclear domain. There are presented characteristics of the following equipment: 1. amplifier for ionization chambers with triggering comparator circuits for the CANDU 600 reactor shut down system; 2. amplifier for ionization chambers without triggering comparator circuits for power regulating system; 3. safety and regulating computerized system for C9 and C5 cans; 4. acquisition system for dosimetric data in nuclear facilities; 5. program able digital comparator for the reactor shut down system; 6. stationary gamma areal monitors for CANDU 600 reactors and other nuclear facilities

  15. Method for controlling FBR type reactor

    International Nuclear Information System (INIS)

    Tamano, Toyomi; Iwashita, Tsuyoshi; Sakuragi, Masanori

    1991-01-01

    The present invention provides a controlling method for moderating thermal transient upon trip in an FBR type reactor. A flow channel for bypassing an intermediate heat exchanger is disposed in a secondary Na system. Then, bypassing flow rate is controlled so as to suppress fluctuations of temperature at a primary exit of the intermediate heat exchanger. Bypassing operation by using the bypassing flow channel is started at the same time with plant trip, to reduce the flow rate of secondary Na flown to the intermediate heat exchanger, so that the imbalance between the primary and the secondary Na flowrates is reduced. Accordingly, fluctuations of the temperature at the primary exit of the intermediate heat exchanger upon trip is suppressed. In view of the above, thermal transient applied to the reactor container upon plant trip can be moderated. As a result, the working life of the reactor can be extended, to improve plant integrity and safety. (I.S.)

  16. 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

  17. Modeling and Dynamic Simulation of the Adjust and Control System Mechanism for Reactor CAREM-25

    International Nuclear Information System (INIS)

    Larreteguy, A.E; Mazufri, C.M

    2000-01-01

    The adjust and control system mechanism, MSAC, is an advanced, and in some senses unique, hydromechanical device.The efforts in modeling this mechanism are aimed to: Get a deep understanding of the physical phenomena involved,Identify the set of parameters relevant to the dynamics of the system,Allow the numerical simulation of the system,Predict the behavior of the mechanism in conditions other than that obtainable within the range of operation of the experimental setup (CEM), and Help in defining the design of the CAPEM (loop for testing the mechanism under high pressure/high temperature conditions).Thanks to the close interaction between the mechanics, the experimenters, and the modelists that compose the MSAC task force, it has been possible to suggest improvements, not only in the design of the mechanism, but also in the design and the operation of the pulse generator (GDP) and the rest of the CEM.This effort has led to a design mature enough so as to be tested in a high-pressure loop

  18. Review of Operation and Maintenance Support Systems for Research Reactors

    International Nuclear Information System (INIS)

    Jin, Kyungho; Heo, Gyunyoung; Park, Jaekwan

    2014-01-01

    Operation support systems do not directly control the plant but it can aid decision making itself by obtaining and analyzing large amounts of data. Recently, the demand of research reactor is growing and the need for operation support systems is increasing, but it has not been applied for research reactors. This study analyzes operation and maintenance support systems of NPPs and suggests appropriate systems for research reactors based on analysis. In this paper, operation support systems for research reactors are suggested by comparing with those of power reactors. Currently, research reactors do not cover special systems in order to improve safety and operability in comparison with power reactors. Therefore we expect to improve worth to use by introducing appropriate systems for research reactors. In further research, we will develop an appropriate system such as applications or tools that can be applied to the research reactor

  19. Reactor safety protection system

    International Nuclear Information System (INIS)

    Nishi, Hiroshi; Yokoyama, Tsuguo.

    1989-01-01

    A plurality of neutron detectors are disposed around a reactor core and detection signals from optional two neutron detectors are inputted into a ratio calculation device. If the ratio between both of the neutron flux level signals exceeds a predetermined value, a reactor trip signal is generated from an alarm setting device. Further, detection signals from all of the neutron detection devices are inputted into an average calculation device and the reactor trip signal is generated also in a case where the average value exceeds a predetermined set value. That is, when the reactor core power is increased locally, the detection signal from the neutron detector nearer to the point of power increase is greater than the increase rate for the entire reactor core power, while the detection signal from the neutron detector remote from the point of power increase is smaller. Thus, the local power increase ratio in the FBR reactor core can be detected efficiently by calculating the ratio for the neutron flux level signals from two neutron detectors, thereby enabling to exactly recognize the local power increase rate in the reactor core. (N.H.)

  20. Development of an automatic reactor inspection system

    International Nuclear Information System (INIS)

    Kim, Jae Hee; Eom, Heung Seop; Lee, Jae Cheol; Choi, Yoo Raek; Moon, Soon Seung

    2002-02-01

    Using recent technologies on a mobile robot computer science, we developed an automatic inspection system for weld lines of the reactor vessel. The ultrasonic inspection of the reactor pressure vessel is currently performed by commercialized robot manipulators. Since, however, the conventional fixed type robot manipulator is very huge, heavy and expensive, it needs long inspection time and is hard to handle and maintain. In order to resolve these problems, we developed a new automatic inspection system using a small mobile robot crawling on the vertical wall of the reactor vessel. According to our conceptual design, we developed the reactor inspection system including an underwater inspection robot, a laser position control subsystem, an ultrasonic data acquisition/analysis subsystem and a main control subsystem. We successfully carried out underwater experiments on the reactor vessel mockup, and real reactor ready for Ulchine nuclear power plant unit 6 at Dusan Heavy Industry in Korea. After this project, we have a plan to commercialize our inspection system. Using this system, we can expect much reduction of the inspection time, performance enhancement, automatic management of inspection history, etc. In the economic point of view, we can also expect import substitution more than 4 million dollars. The established essential technologies for intelligent control and automation are expected to be synthetically applied to the automation of similar systems in nuclear power plants

  1. Spectral shift reactor control method

    International Nuclear Information System (INIS)

    Impink, A.J. Jr.

    1981-01-01

    A method of operating a nuclear reactor having a core and coolant displacer elements arranged in the core wherein is established a reator coolant temperature set point at which it is desired to operate said reactor and first reactor coolant temperature band limits are provided within which said set point is located and it is desired to operate said reactor charactrized in that said reactor coolant displacer elements are moved relative to the reactor core for adjusting the volume of reactor coolant in said core as said reactor coolant temperature approaches said first band limits thereby to maintain said reactor coolant temperature near said set point and within said first band limits

  2. Self-teaching neural network learns difficult reactor control problem

    International Nuclear Information System (INIS)

    Jouse, W.C.

    1989-01-01

    A self-teaching neural network used as an adaptive controller quickly learns to control an unstable reactor configuration. The network models the behavior of a human operator. It is trained by allowing it to operate the reactivity control impulsively. It is punished whenever either the power or fuel temperature stray outside technical limits. Using a simple paradigm, the network constructs an internal representation of the punishment and of the reactor system. The reactor is constrained to small power orbits

  3. Adaptive robust control of the EBR-II reactor

    International Nuclear Information System (INIS)

    Power, M.A.; Edwards, R.M.

    1996-01-01

    Simulation results are presented for an adaptive H ∞ controller, a fixed H ∞ controller, and a classical controller. The controllers are applied to a simulation of the Experimental Breeder Reactor II primary system. The controllers are tested for the best robustness and performance by step-changing the demanded reactor power and by varying the combined uncertainty in initial reactor power and control rod worth. The adaptive H ∞ controller shows the fastest settling time, fastest rise time and smallest peak overshoot when compared to the fixed H ∞ and classical controllers. This makes for a superior and more robust controller

  4. BWR type reactor system

    International Nuclear Information System (INIS)

    Morooka, Shin-ichi.

    1980-01-01

    Purpose: To reduce the internal structure in a reactor by rapidly and efficiently transferring heat generated in a reactor core out of the reactor and eliminating the danger of radiation exposure. Constitution: Steam generated in a pressure vessel is introduced into heat pipe group by inserting the heat pipe group into the steam dome of the pressure vessel. The introduced steam is condensed in the heat pipes to transfer the heat of the steam to the heat pipe group. The transferred heat is transmitted to a heat exchanger provided out of a containment vessel to generate steam to operate a turbine. Thus, it is not necessary to introduce the steam including radioactive substance externally and can remove only the heat so as to carry out the desired purpose. (Kamimura, M.)

  5. New technology for reactor protection system of CAREM reactor

    International Nuclear Information System (INIS)

    Dezzutti, J.C.; Verrastro, C.; Estryk, D.

    2009-01-01

    The use of FPGA in safety functions in a nuclear power plant, increase the reliability of software based systems, without loose any of the function required by the supervision and control systems. In this work the architecture of a Reactor Protection System is described, it use four independent measurement channels in 2 oo 4 configuration, each channel is based on diverse approach in 1 oo 2 configuration, the reliability of this system is near the same than the hardwired logic, with full performance like software based system. (author)

  6. Knowledge-based full-automatic control system for a nuclear ship reactor

    International Nuclear Information System (INIS)

    Shimazaki, J.; Nakazawa, T.; Yabuuchi, N.

    2000-01-01

    Plant operations aboard nuclear ships require quick judgements and actions due to changing marine conditions such as wind, waves and currents. Furthermore, additional human support is not available for nuclear ship operation at sea, so advanced automatic operations are necessary to reduce the number of operators required finally. Therefore, an advanced automatic operating system has been developed based on operational knowledge of nuclear ship 'Mutsu' plant. The advanced automatic operating system includes both the automatic operation system and the operator-support system which assists operators in completing actions during plant accidents, anomaly diagnosis and plant supervision. These system are largely being developed using artificial intelligent techniques such as neural network, fuzzy logic and knowledge-based expert. The automatic operation system is fundamentally based upon application of an operator's knowledge of both normal (start-up to rated power level) and abnormal (after scram) operations. Comparing plant behaviors from start-up to power level by the automatic operation with by 'Mutsu' manual operation, stable automatic operation was obtained almost same as manual operation within all operating limits. The abnormal automatic system was for hard work of manual operations after scram or LOCA accidents. An integrating system with the normal and the abnormal automatic systems are being developed for interacting smoothly both systems. (author)

  7. Monitoring and control of anaerobic reactors

    DEFF Research Database (Denmark)

    Pind, Peter Frode; Angelidaki, Irini; Ahring, Birgitte Kiær

    2003-01-01

    The current status in monitoring and control of anaerobic reactors is reviewed. The influence of reactor design and waste composition on the possible monitoring and control schemes is examined. After defining the overall control structure, and possible control objectives, the possible process...... control approaches that have been used are comprehensively described. These include simple and adaptive controllers, as well as more recent developments such as fuzzy controllers, knowledge-based controllers and controllers based on neural networks....

  8. Monitoring and control of anaerobic reactors

    DEFF Research Database (Denmark)

    Pind, Peter Frode; Angelidaki, Irini; Ahring, Birgitte Kiær

    2003-01-01

    The current status in monitoring and control of anaerobic reactors is reviewed. The influence of reactor design and waste composition on the possible monitoring and control schemes is examined. After defining the overall control structure, and possible control objectives, the possible process mea...... control approaches that have been used are comprehensively described. These include simple and adaptive controllers, as well as more recent developments such as fuzzy controllers, knowledge-based controllers and controllers based on neural networks....

  9. Reactor Systems Technology Division code development and configuration/quality control procedures

    International Nuclear Information System (INIS)

    Johnson, E.C.

    1985-06-01

    Procedures are prescribed for executing a code development task and implementing the resulting coding in an official version of a computer code. The responsibilities of the project manager, development staff members, and the Code Configuration/Quality Control Group are defined. Examples of forms, logs, computer job control language, and suggested outlines for reports associated with software production and implementation are included in Appendix A. 1 raf., 2 figs

  10. Reactor protection systems of 500 MWe PHWRs

    Energy Technology Data Exchange (ETDEWEB)

    Mallik, G; Kelkar, M G; Apte, Ravindra [C and I Group, Nuclear Power Corporation, Mumbai (India)

    1997-03-01

    The 500 MWe PHWR has two totally independent, diverse, fast acting shutdown system called Shutdown System 1 (SDS 1) and Shutdown System 2 (SDS 2). The trip generation circuitry of SDS 1 and SDS 2 are known as Reactor Protection System 1 (RPS 1) and Reactor Protection System 2 (RPS 2) respectively. Some of the features specific to 500 MWe reactors are Core Over Power Protection System (COPPS) based upon in core Self Powered Neutron Detector (SPND) signals, use of local two out of three coincidence logic and adoption of overlap testing for RPS 2, use of Fine Impulse Testing (FIT) in RPS 2, testing of the final control elements namely electro-magnetic clutch of individual Shutoff Rods (SRs) of SDS 1 and all the fast acting valves of SDS 2, etc. The two shutdown systems have totally separate sets of sensors and associated signal processing circuitry as well as physical arrangements. A separate computerised test and monitoring unit is used for each of the two shutdown systems. Use of Programmable Digital Comparator (PDC) unit exclusively for reactor protection systems, has been adopted. The capability of PDC unit is enhanced and communication links are provided for its integration in over all system. The paper describes the design features of reactor protection systems. (author). 3 refs., 7 figs., 3 tabs.

  11. 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

  12. Spectral shift reactor control method

    International Nuclear Information System (INIS)

    Impink, A.J.

    1982-01-01

    A method of operating a nuclear reactor having a core and coolant displacer elements arranged in the core where there is established a reactor coolant temperature set point at which it is desired to operate the reactor and first reactor coolant temperature band limits within which the set point is characterized. The reactor coolant displacer elements are moved relative to the reactor core for adjusting the volume of reactor coolant in the core as the reactor coolant temperature approaches the first band limits to maintain the reactor coolant temperature near the set point and within the first band limits. The reactivity charges associated with movement of respective coolant displacer element clusters is calculated and compared with a calculated derived reactivity charge in order to select the cluster to be moved. (author)

  13. State system experience with safeguarding power reactors

    International Nuclear Information System (INIS)

    Roehnsch, W.

    1982-01-01

    This session describes the development and operation of the State System of Accountancy and Control in the German Democratic Republic, and summarizes operating experience with safeguards at power reactor facilities. Overall organization and responsibilities, containment and surveillance measures, materials accounting, and inspection procedures will be outlined. Cooperation between the IAEA, State system, facility, and supplier authorities will also be addressed

  14. Laser fusion power reactor system (LFPRS)

    International Nuclear Information System (INIS)

    Kovacik, W.P.

    1977-01-01

    This report gives detailed information for each of the following areas: (1) reference concept description, (2) nuclear design, (3) structural design, (4) thermal and fluid systems design, (5) materials design and analysis, (6) reactor support systems and balance of plant, (7) instrumentation and control, (8) environment and safety, (9) economics assessment, and (10) development requirements

  15. Considerations on nuclear reactor passive safety systems

    International Nuclear Information System (INIS)

    2016-01-01

    After having indicated some passive safety systems present in electronuclear reactors (control bars, safety injection system accumulators, reactor cooling after stoppage, hydrogen recombination systems), this report recalls the main characteristics of passive safety systems, and discusses the main issues associated with the assessment of new passive systems (notably to face a sustained loss of electric supply systems or of cold water source) and research axis to be developed in this respect. More precisely, the report comments the classification of safety passive systems as it is proposed by the IAEA, outlines and comments specific aspects of these systems regarding their operation and performance. The next part discusses the safety approach, the control of performance of safety passive systems, issues related to their reliability, and the expected contribution of R and D (for example: understanding of physical phenomena which have an influence of these systems, capacities of simulation of these phenomena, needs of experimentations to validate simulation codes)

  16. Power supply trip control for nuclear reactor

    International Nuclear Information System (INIS)

    Hager, R.E.; Gutman, Jerzy.

    1987-01-01

    A control system for a trip coil in a switchgear mechanism controls the supply of electrical power to a process control device and ensures de-energization of the trip coil shortly after the trip coil is energized. The trip coil is energized not by an independent dc source as in prior art, but from rectified power from a step down transformer supplied from the switchgear output side. The transformer feeds a rectifier which is connected to the trip coil via a trip activation device. The output of the rectifier can be monitored using an optical converter to determine the ability of the control system to activate the trip coil and the condition of the power supplied to the process control device. The control device may be a rod positioner in a pressurised water nuclear reactor. (author)

  17. Modular reactor head shielding system

    International Nuclear Information System (INIS)

    Jacobson, E. B.

    1985-01-01

    An improved modular reactor head shielding system is provided that includes a frame which is removably assembled on a reactor head such that no structural or mechanical alteration of the head is required. The shielding system also includes hanging assemblies to mount flexible shielding pads on trolleys which can be moved along the frame. The assemblies allow individual pivoting movement of the pads. The pivoting movement along with the movement allowed by the trolleys provides ease of access to any point on the reactor head. The assemblies also facilitate safe and efficient mounting of the pads directly to and from storage containers such that workers have additional shielding throughout virtually the entire installation and removal process. The flexible shielding pads are designed to interleave with one another when assembled around the reactor head for substantially improved containment of radiation leakage

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-01

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

  20. Electrical supervisory control and data acquisition system for Power Reactor Fuel Reprocessing Facility (PREFRE-II) at Tarapur

    International Nuclear Information System (INIS)

    Singh, V.K.; Kaushik, S.; Haneef, K.K.M.

    2014-01-01

    Power Reactor Fuel Reprocessing Facility (PREFRE II) is a radio-chemical plant located at Tarapur. The electric power supply to PREFRE-II plant has been provided from a 33 KV Sub-station located in PREFRE-II complex to meet the O and M requirement of plant. The major equipment of the substation includes HT switch board, dry type transformer, LT switch board, Diesel Generator (DG) set and Un-interrupted Power Supply (UPS) System. The power supply to the plant is provided at 415 V voltage level with the help of Power Distribution Boards (PDBs) and Motor Control Centres (MCCs) located at different floors of the plant. The parameters of the electrical equipment of substation and plant are recorded in each shift in the log-book, manually by the operator. To maintain the log-book, the operator needs to go closer to the individual equipment in radioactive area of the plant to read the parameters. The log-book method of monitoring is a cumbersome method and does not include fault event records, trends and diagnostic behavior of the equipment. Electrical Supervisory Control and Data Acquisition (E-SCADA) system has been increasingly used in Nuclear Power Plants (NPPs) for control and monitoring of electrical parameters of plant and switchyard equipment. E-SCADA system has been designed, developed and implemented for PREFRE-II plant for remote monitoring of parameters of electrical equipment. E-SCADA system provides monitoring of electrical equipment and renders complete guidance and information with the help of Graphical User Interface (GUI) to the operator to take necessary action during normal and abnormal conditions. E-SCADA with intelligent communication features helps in reducing cabling from field equipment which enhances the fire safety of plant. As a result engineering, operation and maintenance efforts for monitoring are reduced considerably in terms of requirement of skilled man power and accurate data logging of electrical parameters. This system also helps to

  1. REACTOR: an expert system for diagnosis and treatment of nuclear reactor accidents

    International Nuclear Information System (INIS)

    Nelson, W.R.

    1982-01-01

    REACTOR is an expert system under development at EG and G Idaho, Inc., that will assist operators in the diagnosis and treatment of nuclear reactor accidents. This paper covers the background of the nuclear industry and why expert system technology may prove valuable in the reactor control room. Some of the basic features of the REACTOR system are discussed, and future plans for validation and evaluation of REACTOR are presented. The concept of using both event-oriented and function-oriented strategies for accident diagnosis is discussed. The response tree concept for representing expert knowledge is also introduced

  2. The BWR [Boiling Water Reactor] Emergency Operating Procedures Tracking System (EOPTS): Evaluation by control-room operating crews

    International Nuclear Information System (INIS)

    Spurgin, A.J.; Orvis, D.D.; Spurgin, J.P.; Luna, C.J.

    1990-05-01

    This report presents the results of a project sponsored by the Electric Power Research Institute (EPRI) and Taiwan Power Company (TPC) and conducted by APG and TPC to perform evaluation of the Emergency Operating Procedures Tracking System (EOPTS). The EOPTS is an expert system employing artificial intelligence techniques developed by EPRI for Boiling Water Reactor (BWR) plants based on emergency operating procedures (EOPs). EOPTS is a computerized decision aid used to assist plant operators in efficient and reliable use of EOPs. The main objective of this project was to evaluate the EOPTS and determine how an operator aid of this type could noticeably improve the response time and the reliability of control room crews to multi-failure scenarios. A secondary objective was to collect data on how crew performance was affected. Experiments results indicate that the EOPTS measurably improves crew performance over crews using the EOP flow charts. Time-comparison measurements indicate that crews using the EOPTS perform required actions more quickly than do those using the flowcharts. The results indicate that crews using the EOPTS are not only faster and more consistent in their actions but make fewer errors. In addition, they have a higher likelihood of recovering from the errors that they do make. Use of the EOPTS in the control room should result in faster termination and mitigation of accidents and reduced risk of power plant operations. Recommendations are made towards possible applications of the EOPTS to operator training and evaluation, and for the applicability of the evaluation methodology developed for this project to the evaluation of similar operator aides. 17 refs., 14 figs., 14 tabs

  3. Centralized digital computer control of a research nuclear reactor

    International Nuclear Information System (INIS)

    Crawford, K.C.

    1987-01-01

    A hardware and software design for the centralized control of a research nuclear reactor by a digital computer are presented, as well as an investigation of automatic-feedback control. Current reactor-control philosophies including redundancy, inherent safety in failure, and conservative-yet-operational scram initiation were used as the bases of the design. The control philosophies were applied to the power-monitoring system, the fuel-temperature monitoring system, the area-radiation monitoring system, and the overall system interaction. Unlike the single-function analog computers currently used to control research and commercial reactors, this system will be driven by a multifunction digital computer. Specifically, the system will perform control-rod movements to conform with operator requests, automatically log the required physical parameters during reactor operation, perform the required system tests, and monitor facility safety and security. Reactor power control is based on signals received from ion chambers located near the reactor core. Absorber-rod movements are made to control the rate of power increase or decrease during power changes and to control the power level during steady-state operation. Additionally, the system incorporates a rudimentary level of artificial intelligence

  4. Control unit of a nuclear reactor

    International Nuclear Information System (INIS)

    Desfontaines, Guy; Le Helloco, Michel.

    1981-01-01

    Control unit comprising multiple leak-tight vessels, in communication with the inside of the reactor vessel, extending this vessel above its cover, in the vertical direction and each one enclosing a mechanism for moving a cluster of material absorbing the neutrons in the reactor core, actuated by a motor. This control unit is of reduced height above the vessel cover and provides efficient protection of the leak tight containments and the mechanisms in the event of earthquakes, easy removal and refitting of the vessel cover, good ventilation of the power devices of the mechanisms without the use of a complex ventilation system, efficient thermal insulation of the leak-tight containments assembly, as well as easy access to the motors and mechanism located in the leak-tight containment for carrying out any maintenance and repairs that might be required [fr

  5. Safety system for reactor container

    International Nuclear Information System (INIS)

    Shimizu, Miwako; Seki, Osamu; Mano, Takio.

    1995-01-01

    A slanted structure is formed below a reactor core where there is a possibility that molten reactor core materials are dropped, and above a water level of a pool which is formed by coolants flown from a reactor recycling system and accumulated on the inner bottom of the reactor container, to prevent molten fuels from dropping at once in the form of a large amount of lump. The molten materials are provisionally received on the structure, gradually formed into small pieces and then dropped. Further, the molten materials are dropped and received provisionally on a group of coolant-flowing pipelines below the structure, to lower the temperature of the molten materials, and then the reactor core molten materials are gradually formed into small pieces and dropped into the pool water. Since they are not dropped directly into the pool water but dropped gradually into the pool water as small droplets, occurrence of steam explosion can be reduced. The occurrence of steam explosion due to dropped molten reactor core material and pool water is suppressed, and the molten materials are kept in the pool water, thereby enabling to maintain the integrity of the reactor container more effectively. (N.H.)

  6. Principle of human system interface (HSI) design for new reactor console of PUSPATI TRIGA Reactor (RTP)

    International Nuclear Information System (INIS)

    Zareen Khan Abdul Jalil Khan; Ridzuan Abdul Mutalib; Mohd Idris Taib; Mohd Khairulezwan Abdul Manan; Nurfarhana Ayuni Joha; Mohd Sabri Minhat; Izhar Abu Hussin

    2013-01-01

    Full-text: This paper will describe the principle of human system interface design for new reactor console in control room at TRIGA reactor facility. In order to support these human system interface challenges in digital reactor console. Software-based instrumentation and control (I and C) system for new reactor console could lead to new human machine integration. The proposed of Human System Interface (HSI) which included the large display panels which shows reactor status, compact and computer-based workstations for monitoring, control and protection function. The proposed Human System Interface (HIS) has been evaluated using various human factor engineering. It can be concluded that the Human System Interface (HIS) is designed as to address the safety related computer controlled system. (author)

  7. Development of a software for the control of the quality management system of the TRIGA-Mark III reactor; Desarrollo de un software para el control del sistema de gestion de calidad del reactor TRIGA Mark III

    Energy Technology Data Exchange (ETDEWEB)

    Herrera A, E. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico); Hernandez, L.V.; Hernandez, J.A. [UAEM, Depto. de Ingenieria en Computacion, 50000 Toluca, estado de Mexico (Mexico)]. e-mail: eha@nuclear.inin.mx

    2006-07-01

    The quality has not only become one of the essential requirements of the product but rather at the presenme it is a strategic factor key of which depends the bigger part of the organizations, not only to maintain their position in the market but also to assure their survival. The good organizations will have processes, procedures and standards to confront these challenges. The big organizations require of the certification of their administration systems, and once the organization has obtained this certification the following step it is to maintain it. The implementation and certification of an administration system requires of an appropriate operative organization that achieves continuous improvements in their operation. This is the case of the TRIGA Mark III reactor, which contains a computer program that upgrades, it controls and it programs activities to develop in the Installation, allowing one operative organization to the whole personnel of the same one. With the purpose of avoiding activities untimely. (Author)

  8. 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.

  9. 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'

  10. Reactor protection system. Revision 1

    International Nuclear Information System (INIS)

    Fairbrother, D.B.; Vincent, D.R.; Lesniak, L.M.

    1975-04-01

    The reactor protection system-II (RPS-II) designed for use on Babcock and Wilcox 145- and 205-fuel assembly pressurized water reactors is described. In this system, relays in the trip logic have been replaced by solid state devices. A calculating module for the low DNBR, pump status, and offset trip functions has replaced the overpower trip (based on flow and imbalance), the power/RC pump trip, and the variable low pressure trip. Included is a description of the changes from the present Oconee-type reactor protection system (RPS-I), a functional and hardware description of the calculating module, and a discussion of the qualification program conducted to ensure that the degree of protection provided by RPS-II is not less than that provided by previously licensed systems supplied by B and W. (U.S.)

  11. Large scale replacement of fuel channels in the Pickering CANDU reactor using a man-in-the-loop remote control system

    International Nuclear Information System (INIS)

    Stratton, D.

    1991-01-01

    Spar Aerospace Limited of Toronto is presently under contract to Ontario Hydro to design a Remote Manipulation and Control System (RMCS) to be used during the large scale replacement of the fuel channels in the Pickering A Nuclear Generating Station. The system is designed to support the replacement of all 390 fuel channels in each of the four reactors at the Pickering A station in a safe manner that minimizes worker radiation exposure and unit outage time

  12. Coolant cleanup system for a nuclear reactor

    International Nuclear Information System (INIS)

    Shiina, Atsushi; Usui, Naoshi; Yamamoto, Michiyoshi; Osumi, Katsumi.

    1983-01-01

    Purpose: To maintain the electric conductivity of reactor water lower and to minimize the heat loss in the cleanup system by providing a low temperature cleanup system and a high temperature cleanup system together. Constitution: A low temperature cleanup system using ion exchange resins as filter aids and a high temperature cleanup system using inorganic ion exchange materials as filter aids are provided in combination. A part of the reactor water in a reactor pressure vessel is passed through a conductivity meter, one portion of which flows into the high temperature cleanup system having no heat exchanger and filled with inorganic ion exchange materials by way of a first flow rate control valve and the other portion of which flows into the low temperature cleanup system having heat exchangers and filled with the ion exchange materials by way of a second control valve. The first control valve is adjusted so as to flow, for example, about more than 15% of the feedwater flow rate to the high temperature cleanup system and the second control valve is adjusted with its valve opening degree depending on the indication of the conductivity meter so as to flow about 2 - 7 % of the feedwater flow rate into the low temperature cleanup system, to thereby control the electric conductivity to between 0.055 - 0.3 μS/cm. (Moriyama, K.)

  13. Power conditioning system for a nuclear reactor

    International Nuclear Information System (INIS)

    Higashigawa, Yuichi; Joge, Toshio.

    1981-01-01

    Purpose: To provide a power conditioning system for a BWR type reactor which has a function to be automatically operated within a range that the relationship between the heat power of the reactor and the electric power of an electric generator does not lose the safety of fuel by eliminating the unnecessary fluctuation of the power of the reactor. Constitution: A load request error signal fed from a conventional turbine control system to recirculation flow regulator is eliminated, and a reactor power conditioning system is newly provided, to which an electric generator power signal, a reactor average power area monitor signal and a load request signal are inputted. Thus, the load request signal is compared directly with the electric power of the electric generator, the recirculation flow rate is controlled by the compared result, and whether the correlation between the heat power of the reqctor and the electric power of the generator satisfies the correlation determined to prove the safety of fuel or not is checked. If this correlation is satisfied, the recirculation flow rate is merely automatically controlled. (Yoshino, Y.)

  14. Cooling system upon reactor isolation

    International Nuclear Information System (INIS)

    Yamamoto, Kohei; Oda, Shingo; Miura, Satoshi

    1992-01-01

    A water level indicator for detecting the upper limit value for a range of using a suppression pool and a thermometer for detecting the temperature of water at the cooling water inlet of an auxiliary device are disposed. When a detection signal is intaken and the water level in the suppression pool reach the upper limit value for the range of use, a secondary flow rate control value is opened and a primary flow rate control valve is closed. When the temperature of the water at the cooling water inlet of the auxiliary device reaches the upper limit value, the primary and the secondary flow rate control valves are opened. During a stand-by state, the first flow rate control valve is set open and the secondary flow rate control valve is set closed respectively. After reactor isolation, if a reactor water low level signal is received, an RCIC pump is actuated and cooling water is sent automatically under pressure from a condensate storage tank to the reactor and the auxiliary device requiring coolants by way of the primary flow rate control valve. Rated flow rate is ensured in the reactor and cooling water of an appropriate temperature can be supplied to the auxiliary device. (N.H.)

  15. Direct digital temperature control of the A-1 nuclear reactor

    International Nuclear Information System (INIS)

    Karpeta, C.

    1975-01-01

    The application is described of one of the modern control methods for designing an experimental digital temperature control system for heavy water moderated gas cooled reactors. The synthesis of the optimal stochastic regulator for reactor control in the area of the rated steady state was carried out using the method of dynamic programming and the Kalman filter technique. The analysis of the feedback circuit was conducted using control simulation on a universal digital computer. Results and experience are summed up. (author)

  16. 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)

  17. Neutron flux measuring system for nuclear reactor

    International Nuclear Information System (INIS)

    Aoki, Kazuo.

    1977-01-01

    Purpose: To avoid the generation of an undesired scram signal due to abrupt changes in the neutron level given to the detectors disposed near the boundary between the moderator and the atmosphere. Constitution: In a nuclear reactor adapted to conduct power control by the change of the level in the moderator such as heavy water, the outputs from the neutron detectors disposed vertically are averaged and the nuclear reactor is scramed corresponding to the averaged value. In this system, moderator level detectors are additionally provided to the nuclear reactor and their outputs, moderator level signal, are sent to a power averaging device where the output signals of the neutron detectors are judged if they are delivered from neutrons in the moderator or not depending on the magnitude of the level signal and the outputs of the detectors out of the moderator are substantially excluded. The reactor interlock signal from the device is utilized as a scram signal. (Seki, T.)

  18. Reactor fueling system

    International Nuclear Information System (INIS)

    Hattori, Noriaki; Hirano, Haruyoshi.

    1983-01-01

    Purpose: To optimally position a fuel catcher by mounting a television camera to a fuel catching portion and judging video images by the use of a computer or the like. Constitution: A television camera is mounted to the lower end of a fuel catching mechanism for handling nuclear fuels and a fuel assembly disposed within a reactor core or a fuel storage pool is observed directly from above to judge the position for the fuel assembly by means of video signals. Then, the relative deviation between the actual position of the fuel catcher and that set in a memory device is determined and the positional correction is carried out automatically so as to reduce the determined deviation to zero. This enables to catch the fuel assembly without failure and improves the efficiency for the fuel exchange operation. (Moriyama, K.)

  19. Power control device for nuclear reactors

    International Nuclear Information System (INIS)

    Kagawa, Tatsuo

    1984-01-01

    Purpose: To eliminate for requirement of control rods and movable portions, as well as ensure the safety and reliability of the operation. Constitution: A plurality of control tubes are disposed within a reactor core instead of control rods. Tubes are connected from below the reactor core to the control tubes for supplying liquid poisons such as aqueous boric acid to the inside of the control tubes. Further, tubes are connected to the upper portion of the control tubes for guiding the liquid poisons from the reactor core to the outside. The tubes for supplying and discharging the liquid poisons are introduced externally through the flange disposed at the upper portion of a pressure vessel. At the outside of the pressure vessel, are disposed a liquid poison tank, a pressurizing source, a pressure control valve, a liquid level meter and the like. The control for the reactor power is conducted by controlling the level of the liquid poisons in the control tubes. (Ikeda, J.)

  20. Power reactor information system (PRIS)

    International Nuclear Information System (INIS)

    1989-06-01

    Since the very beginning of commercial operation of nuclear power plants, the nuclear power industry worldwide has accumulated more than 5000 reactor years of experience. The IAEA has been collecting Operating Experience data for Nuclear Power Plants since 1970 which were computerized in 1980. The Agency has undertaken to make Power Reactor Information System (PRIS) available on-line to its Member States. The aim of this publication is to provide the users of PRIS from their terminals with description of data base and communication systems and to show the methods of accessing the data

  1. Breeder control fusion reactor. Topical interview

    Energy Technology Data Exchange (ETDEWEB)

    Schlueter, A [Max-Planck-Institut fuer Plasmaphysik, Garching/Muenchen (Germany, F.R.)

    1977-09-01

    The energy sources of the future are extremely controversial. The consumption of fossil fuel shall decrease during the next decades, because exhaustion of the resources, pollution, increase of CO/sub 2/ in the atmosphere and other reasons. But at present the question it not yet settled which alternative energy system should replace the fossil fuel. First of all nuclear energy in the form of fission reactions seems to come into operation to a larger extent. The next step may be the controlled thermonuclear fusion reaction. Furthermore, a comparison between fusion and fission is given which shows that fusion would bring about less risks than the breeders. An advantage of the fusion reactor would be the fact that the fuel cycle is closed. Unfortunately, the physical questions are not as yet satisfactorily clarified so that one cannot be sure whether a fusion reactor can really be built.

  2. Solvent refined coal reactor quench system

    Science.gov (United States)

    Thorogood, Robert M.

    1983-01-01

    There is described an improved SRC reactor quench system using a condensed product which is recycled to the reactor and provides cooling by evaporation. In the process, the second and subsequent reactors of a series of reactors are cooled by the addition of a light oil fraction which provides cooling by evaporation in the reactor. The vaporized quench liquid is recondensed from the reactor outlet vapor stream.

  3. Report from the Light Water Reactor Sustainability Workshop on Advanced Instrumentation, Information, and Control Systems and Human-System Interface Technologies

    International Nuclear Information System (INIS)

    Hallbert, Bruce P.; Persensky, J.J.; Smidts, Carol; Aldemir, Tunc; Naser, Joseph

    2009-01-01

    The Light Water Reactor Sustainability (LWRS) Program is a research and development (R and D) program sponsored by the U.S. Department of Energy (DOE). The program is operated in close collaboration with industry R and D programs to provide the technical foundations for licensing and managing the long-term, safe, and economical operation of Nuclear Power Plants that are currently in operation. The LWRS Program focus is on longer-term and higher-risk/reward research that contributes to the national policy objectives of energy and environmental security. Advanced instruments and control (I and C) technologies are needed to support the safe and reliable production of power from nuclear energy systems during sustained periods of operation up to and beyond their expected licensed lifetime. This requires that new capabilities to achieve process control be developed and eventually implemented in existing nuclear assets. It also requires that approaches be developed and proven to achieve sustainability of I and C systems throughout the period of extended operation. The strategic objective of the LWRS Program Advanced Instrumentation, Information, and Control Systems Technology R and D pathway is to establish a technical basis for new technologies needed to achieve safety and reliability of operating nuclear assets and to implement new technologies in nuclear energy systems. This will be achieved by carrying out a program of R and D to develop scientific knowledge in the areas of: (1) Sensors, diagnostics, and prognostics to support characterization and prediction of the effects of aging and degradation phenomena effects on critical systems, structures, and components (SSCs); (2) Online monitoring of SSCs and active components, generation of information, and methods to analyze and employ online monitoring information; (3) New methods for visualization, integration, and information use to enhance state awareness and leverage expertise to achieve safer, more readily available

  4. Report from the Light Water Reactor Sustainability Workshop on Advanced Instrumentation, Information, and Control Systems and Human-System Interface Technologies

    Energy Technology Data Exchange (ETDEWEB)

    Bruce P. Hallbert; J. J. Persensky; Carol Smidts; Tunc Aldemir; Joseph Naser

    2009-08-01

    The Light Water Reactor Sustainability (LWRS) Program is a research and development (R&D) program sponsored by the U.S. Department of Energy (DOE). The program is operated in close collaboration with industry R&D programs to provide the technical foundations for licensing and managing the long-term, safe, and economical operation of Nuclear Power Plants that are currently in operation. The LWRS Program focus is on longer-term and higher-risk/reward research that contributes to the national policy objectives of energy and environmental security. Advanced instruments and control (I&C) technologies are needed to support the safe and reliable production of power from nuclear energy systems during sustained periods of operation up to and beyond their expected licensed lifetime. This requires that new capabilities to achieve process control be developed and eventually implemented in existing nuclear assets. It also requires that approaches be developed and proven to achieve sustainability of I&C systems throughout the period of extended operation. The strategic objective of the LWRS Program Advanced Instrumentation, Information, and Control Systems Technology R&D pathway is to establish a technical basis for new technologies needed to achieve safety and reliability of operating nuclear assets and to implement new technologies in nuclear energy systems. This will be achieved by carrying out a program of R&D to develop scientific knowledge in the areas of: • Sensors, diagnostics, and prognostics to support characterization and prediction of the effects of aging and degradation phenomena effects on critical systems, structures, and components (SSCs) • Online monitoring of SSCs and active components, generation of information, and methods to analyze and employ online monitoring information • New methods for visualization, integration, and information use to enhance state awareness and leverage expertise to achieve safer, more readily available electricity generation

  5. Management system requirements for small reactors

    Energy Technology Data Exchange (ETDEWEB)

    Jones, K.A., E-mail: kenneth.jones@cnsc-ccsn.gc.ca [Canadian Nuclear Safety Commission, Ottawa, Ontario (Canada)

    2013-07-01

    This abstract identifies the management system requirements for the life cycle of small reactors from initial conception through completion of decommissioning. For small reactors, the requirements for management systems remain the same as those for 'large' reactors regardless of the licensee' business model and objectives. The CSA N-Series of standards provides an interlinked set of requirements for the management of nuclear facilities. CSA N286 provides overall direction to management to develop and implement sound management practices and controls, while other CSA nuclear standards provide technical requirements and guidance that support the management system. CSA N286 is based on a set of principles. The principles are then supported by generic requirements that are applicable to the life cycle of nuclear facilities. CNSC regulatory documents provide further technical requirements and guidance. (author)

  6. A digital controller for the Omega West Reactor

    International Nuclear Information System (INIS)

    Minor, M.M.; Kaufman, M.D.; Smith, T.W.

    1992-05-01

    A new nuclear reactor control system for the Omega West Reactor (OWR) has been designed to replace the aging and hard to maintain controller presently installed. The controller uses single board computers, digital and analog input and output modules, and stepping motor indexers installed on a standard bus (VME bus). The eight poison control rod drive motors are replaced with stepping motors. The control algorithm for the OWR was not changed in order to expedite approval for installation. This report presents the results of the development of the new control system. Included in the report are copies of some of the software that drives the new controller

  7. Material accountancy and control practice at a research reactor facility

    International Nuclear Information System (INIS)

    Bouchard, J.; Maurel, J.J.; Tromeur, Y.

    1982-01-01

    This session surveys the regulations, organization, and accountancy practice that compose the French State System of Accountancy and Control. Practical examples are discussed showing how inventories are verified at a critical assembly facility and at a materials testing reactor

  8. Protection of semiconductor converters for controlled bypass reactors

    International Nuclear Information System (INIS)

    Dolgopolov, A. G.; Akhmetzhanov, N. G.; Karmanov, V. F.

    2010-01-01

    Possible ways of protecting thyristor converters in systems for magnetizing 110 - 500 kV controlled bypass reactors during switching and automatic reclosing are examined based on experience with the development of equipment, line tests, and mathematical modelling.

  9. Instrumentation and control improvements at Experimental Breeder Reactor II

    International Nuclear Information System (INIS)

    Christensen, L.J.; Planchon, H.P.

    1993-01-01

    The purpose of this paper is to describe instrumentation and control (I ampersand C) system improvements at Experimental Breeder Reactor 11 (EBR-11). The improvements are focused on three objectives; to keep the reactor and balance of plant (BOP) I ampersand C systems at a high level of reliability, to provide diagnostic systems that can provide accurate information needed for analysis of fuel performance, and to provide systems that will be prototypic of I ampersand C systems of the next generation of liquid metal reactor (LMR) plants

  10. Feed water control device in a reactor

    International Nuclear Information System (INIS)

    Okutani, Tetsuro.

    1984-01-01

    Purpose: To prevent substantial fluctuations of the water level in a nuclear reactor and always keep a constant standard level under any operation condition. Constitution: When the causes for fluctuating the reactor water level is resulted, a certain amount of correction signal is added to a level deviation signal for the difference between the reactor standard level and the actual reactor water level to control the flow rate of the feed water pump depending on the addition signal. If reactor scram should occur, for instance, a level correction signal changing stepwise depending on a scram signal is outputted and added to the level deviation signal. As the result, the flow rate of feed water sent into the reactor just after the scram is increased, whereby the lowering in the reactor water level upon scram can be decreased as compared with the case where no such level compensation signal is inputted. (Kamimura, M.)

  11. Application of the dose limitation system to the control of carbon-14 releases from heavy-water-moderated reactors

    International Nuclear Information System (INIS)

    Beninson, D.; Gonzalez, A.J.

    1982-01-01

    Heavy-water-moderated reactors produce substantially more carbon-14 than light-water reactors. Applying the principles of the systems of dose limitation, the paper presents the rationale used for establishing the release limit for effluents containing this nuclide and for the decisions made regarding the effluent treatment in the third nuclear power station in Argentina. Production of carbon-14 in PHWR and the release routes are analysed in the light of the different effluent treatment possibilities. An optimization assessment is presented, taking into account effluent treatment and waste management costs, and the collective effective dose commitment due to the releases. The contribution of present carbon-14 releases to future individual doses is also analysed in the light of an upper bound for the contribution, representing a fraction of the individual dose limits. The paper presents the resulting requirements for the effluent treatment regarding carbon-14 and the corresponding regulatory aspects used in Argentina. (author)

  12. Dynamics and control of molten-salt breeder reactor

    Directory of Open Access Journals (Sweden)

    Vikram Singh

    2017-08-01

    Full Text Available Preliminary results of the dynamic analysis of a two-fluid molten-salt breeder reactor (MSBR system are presented. Based on an earlier work on the preliminary dynamic model of the concept, the model presented here is nonlinear and has been revised to accurately reflect the design exemplified in ORNL-4528. A brief overview of the model followed by results from simulations performed to validate the model is presented. Simulations illustrate stable behavior of the reactor dynamics and temperature feedback effects to reactivity excursions. Stable and smooth changes at various nodal temperatures are also observed. Control strategies for molten-salt reactor operation are discussed, followed by an illustration of the open-loop load-following capability of the molten-salt breeder reactor system. It is observed that the molten-salt breeder reactor system exhibits “self-regulating” behavior, minimizing the need for external controller action for load-following maneuvers.

  13. Dynamics and control of molten-salt breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sing, Vikram; Lish, Matthew R.; Chvala, Ondrej; Upadhyaya, Belle R. [Dept. of Nuclear Engineering, University of Tennessee, Knoxville (United States)

    2017-08-15

    Preliminary results of the dynamic analysis of a two-fluid molten-salt breeder reactor (MSBR) system are presented. Based on an earlier work on the preliminary dynamic model of the concept, the model presented here is nonlinear and has been revised to accurately reflect the design exemplified in ORNL-4528. A brief overview of the model followed by results from simulations performed to validate the model is presented. Simulations illustrate stable behavior of the reactor dynamics and temperature feedback effects to reactivity excursions. Stable and smooth changes at various nodal temperatures are also observed. Control strategies for molten-salt reactor operation are discussed, followed by an illustration of the open-loop load-following capability of the molten-salt breeder reactor system. It is observed that the molten-salt breeder reactor system exhibits “self-regulating” behavior, minimizing the need for external controller action for load-following maneuvers.

  14. Controllability studies for an advanced CANDU boiling light water reactor

    International Nuclear Information System (INIS)

    Lepp, R.M.; Hinds, H.W.

    1976-12-01

    Bulk controllability studies carried out as part of a conceptual design study of a 1200 MWe CANDU boiling-light-water reactor fuelled with U 235 - or Pu-enriched uranium oxide are outlined. The concept, the various models developed for its simulation on a hybrid computer and the perturbations used to test system controllability, are described. The results show that this concept will have better bulk controllability than similar CANDU-BLW reactors fuelled with natural uranium. (author)

  15. Reactor power control method and device

    International Nuclear Information System (INIS)

    Fushimi, Atsushi; Ishii, Yoshihiko; Miyamoto, Yoshiyuki; Ishii, Kazuhiko; Kiyoharu, Norihiko; Aizawa, Yuko.

    1997-01-01

    The present invention provides a method and a device suitable to rise the temperature and increase the pressure of the reactor to an aimed pressure in accordance with an aimed value for a reactor water temperature changing rate in the course of rising temperature and increasing pressure of the reactor upon start up of a BWR type power plant. Namely, neutron fluxes in the reactor and the temperature of reactor water are detected respectively. The maximum value among the detected values for the neutron fluxes is detected. The reactor water temperature changing rate is calculated based on the detected values of the reactor water temperature, from which the maximum value of the reactor water temperature changing rate is detected. An aimed value for the neutron flux is calculated in accordance with both detected maximum values and the aimed value of the reactor water temperature changing rate. The position of control rods is adjusted in accordance with the aimed value for the calculated neutron flux. Then, an aimed value for the neutron flux for realizing the aimed value for the reactor water temperature changing rate can be obtained accurately with no influence of the sensitivity of the detected values of the neutron fluxes and the time delay of the reactor water temperature changing rate. (I.S.)

  16. Annexes to the lecture on reactor protection system including engineered features actuation system

    International Nuclear Information System (INIS)

    Palmaers, W.

    1982-01-01

    The present paper deals with the fundamentals for a reactor protection system and discusses the following topics: - System lay-out - Analog measured data acquisition - Analog measured data processing - Limit value generation and logical gating - Procesing of the reactor protection actuation signals - Decoupling of the reactor protection system - Mechanical lay-out - Monitoring system and - Emergency control station. (orig./RW)

  17. Review of the book: Yurkevich, G.P. Energetic reactor control systems. Editing by N.S. Khlopkin. Moscow, 2001, 344 p

    International Nuclear Information System (INIS)

    Kondrat'ev, V.V.

    2002-01-01

    Review of the book: Yurkevich, G.P. Energetic reactor control systems. Editing by N.S. Khlopkin. Moscow, 2001, is presented. Advantages of the book, specifically, easy of presentation of sophistical physical processes, new, practical and valid theoretical ways of decision of the problems are discussed. New and original decisions proposed by the author of the book are discussed. The book is beneficial for wide round of specialists [ru

  18. Coolant clean up system in nuclear reactor

    International Nuclear Information System (INIS)

    Tajima, Fumio; Iwami, Hiroshi.

    1981-01-01

    Purpose: To decrease the amount of main steams and improve the plant heat efficiency by the use of condensated water as coolants for not-regenerative heat exchangers in a coolant clean up system of a nuclear reactor. Constitution: In a coolant clean up system of a nuclear reactor, a portion of condensates is transferred to the shell of a non-regenerative heat exchanger by way of a condensate pump for non-regenerative heat exchanger through a branched pipeway provided to the outlet of a condensate desalter for using the condensates as the coolants for the shell of the heat exchanger and the condensates are then returned to the inlet of a feedwater heater after the heat exchange. The branched flow rate of the condensates is controlled by the flow rate control valve mounted in the pipeway. Condensates passed through the heat exchanger and the condensates not passed through the heat exchanger are mixed and heated in a heater and then fed to the nuclear reactor. In a case where no feedwater is necessary to the nuclear reactor such as upon shutdown of the reactor, the condensates are returned by way of feedwater bypass pipeway to the condensator. By the use of the condensates as the coolants for the heat exchanger, the main steam loss can be decreased and the thermal load for the auxiliary coolant facility can be reduced. (Kawakami, Y.)

  19. Power Reactor Information System (PRIS)

    International Nuclear Information System (INIS)

    Spiegelberg, R.

    1992-01-01

    The IAEA has been collecting Operating Experience data for Nuclear Power Plants of the IAEA Member States since 1970. In order to facilitate an analysis of nuclear power plant performance as well as to produce relevant publications, all previously collected data supplied from the questionnaires were computerized in 1980 and the Power Reactor Information System was implemented. PRIS currently contains production records for the years up to and including 1990 and about 98% of the reactors-years operating experience in the world is contained in PRIS. (orig.)

  20. Modern control technology for improved nuclear reactor performance

    International Nuclear Information System (INIS)

    Oakes, L.C.

    1986-01-01

    One of the main complaints leveled at reactor control systems by utility spokesmen is complexity. One only has to look inside a power reactor control room to appreciate this viewpoint. The high reliability and versatility of modern microprocessors makes possible distributed control systems with only performance data and abnormal conditions being relayed to the control room. In a sense, this emulates the human-body control system where routine repetitive actions are handled in an involuntary manner. The significance of expert systems to the nuclear reactor control and safety systems is their ability to capture human and other expertise and make it available, upon demand, and under almost all circumstances. Thus, human problem-solving skills acquired by the learning process over a long period of time can be captured and employed with the reliability inherent in computers. This is especially important in nuclear plants when human operators are burdened by stress and emotional factors that have a dramatic effect on performance level

  1. A programming language study and an implementation of a simulation system with formal derivation. Application to nuclear reactors, control systems and elecronic networks

    International Nuclear Information System (INIS)

    Nakhle, Michel.

    1979-06-01

    Physical systems simulation requires a lot of information about the controlled process, and the mathematical approach must be appropriate. On the other hand, the parameters describing most systems components are nonlinear and time dependent. Moreover the differential equations describing them are 'stiff' equations of high order. The scope of the study is the description of the NEPTUNIX language and the differential equations. Most of the algorithms used, and the programs implementing these algorithms, are dealt with. Examples of nuclear reactors and mechanical processes simulation are investigated. NEPTUNIX handles a given mathematical description of a continuous system such as: f (x, x(.), t) = 0. Even more, symbolic derivation is performed automatically in order to compute the jacobian associated with the system, requisite for the numerical integration. So, for large systems the manual method for computer the jacobian and the classical method of differentiation are avoided, the former being tiresome and consuming of human time and the latter being costly in run time. The jacobian evaluated in this way is dealt with, by the approach of sparse matrices. Every element of this matrix is assigned a type attribute to improve time execution. Moreover, this is done only once, for a physical system which is described by a mathematical model which topology is invariant. The results of this process are sayed on a suitable device ready for performing repeated simulations [fr

  2. Study of reactor parameters of on critical systems, Phase I: Safety report for RB zero power reactor

    International Nuclear Information System (INIS)

    Raisic, N.

    1962-09-01

    In addition to the safety analysis for the zero power RB reactor, this report contains a general description of the reactor, reactor components, auxiliary equipment and the reactor building. Reactor Rb has been reconstructed during 1961-1962 and supplied with new safety-control system as well as with a complete dosimetry instrumentation. Since RB reactor was constructed without shielding special attention is devoted to safety and protection of the staff performing experiments. Due to changed circumstances in the Institute ( start-up of the RA 7 MW power reactor) the role of the RB reactor was redefined

  3. Method and device for controlling reactor power

    International Nuclear Information System (INIS)

    Oohashi, Masahisa; Masuda, Hiroyuki.

    1982-01-01

    Purpose: To enable load following-up operation of a reactor adapted to perform power conditioning by the control of the liquid poison density in the core and by the control rods. Constitution: In a case where the reactor power is repeatedly changed in a reactor having a liquid poison density control device and control rods, the time period for the power control is divided depending on the magnitude of the change with time in the reactivity and the optimum values are set for the injection and removal amount of the liquid poison within the divided period. Then, most parts of the control required for the power change are alloted to the liquid poison that gives no effect on the power distribution while minimizing the movement of the control rods, whereby the power change in the reactor as in the case of the load following-up operation can be practiced with ease. (Kawakami, Y.)

  4. Status of control assembly materials in Indian water reactors

    International Nuclear Information System (INIS)

    Date, V.G.; Kulkarni, P.G.

    2000-01-01

    India's present operating water cooled power reactors comprise boiling water reactors of Tarapur Atomic Power Station (TAPS) and pressurized heavy water reactors (PHWRs) at Kota (RAPS), Kalpakkam (MAPS), Narora (NAPS) and Kakrapara (KAPS). Boiling water reactors of TAPS use boron carbide control blades for control of power as well as for shut down (scram). PHWRs use boron steel and cobalt absorber rods for power control and Cd sandwiched shut off rods (primary shut down system) and liquid poison rods (secondary shut down system) for shut down. In TAPS, Gadolinium rods (burnable poison rods) are also incorporated in fuel assembly for flux flattening. Boron carbide control blades and Gadolinium rods for TAPS, cobalt absorber rods and shut down assemblies for PHWRs are fabricated indigenously. Considerable development work was carried out for evolving material specifications, component and assembly drawings, and fabrication processes. Details of various control and shut off assemblies being fabricated currently are highlighted in the paper. (author)

  5. Reactor shutdown back-up system

    International Nuclear Information System (INIS)

    Hirao, Seizo; Sakashita, Motoaki.

    1982-01-01

    Purpose: To prevent back flow of poison upon injection to a moderator recycling pipeway. Constitution: In a nuclear reactor comprising a moderator recycling system for recycling and cooling moderator through a control rod guide pipe and a rapid poison injection system for rapidly injecting a poison solution at high density into the moderator by way of the same control rod guide pipe as a reactor shutdown back-up system, a mechanism is provided for preventing the back flow of a poison solution at high density into the moderator recycling system upon rapid injection of poison. An orifice provided in the joining pipeway to the control rod guide pipe on the side of the moderator recycling system is utilized as the back flow preventing device for the poison solution and the diameter for the orifice is determined so as to provide a constant ratio between the pressure loss in the control rod guide pipe and the pressure loss in the moderator recycling system pipe line upon usual reactor operation. (Kawakami, Y.)

  6. Transients in reactors for power systems compensation

    Science.gov (United States)

    Abdul Hamid, Haziah

    This thesis describes new models and investigations into switching transient phenomena related to the shunt reactors and the Mechanically Switched Capacitor with Damping Network (MSCDN) operations used for reactive power control in the transmission system. Shunt reactors and MSCDN are similar in that they have reactors. A shunt reactor is connected parallel to the compensated lines to absorb the leading current, whereas the MSCDN is a version of a capacitor bank designed as a C-type filter for use in the harmonic-rich environment. In this work, models have been developed and transient overvoltages due to shunt reactor deenergisation were estimated analytically using MathCad, a mathematical program. Computer simulations used the ATP/EMTP program to reproduce both single-phase and three-phase shunt reactor switching at 275 kV operational substations. The effect of the reactor switching on the circuit breaker grading capacitor was also examined by considering various switching conditions.. The main original achievement of this thesis is the clarification of failure mechanisms occurring in the air-core filter reactor due to MSCDN switching operations. The simulation of the MSCDN energisation was conducted using the ATP/EMTP program in the presence of surge arresters. The outcome of this simulation shows that extremely fast transients were established across the air-core filter reactor. This identified transient event has led to the development of a detailed air-core reactor model, which accounts for the inter-turn RLC parameters as well as the stray capacitances-to-ground. These parameters are incorporated into the transient simulation circuit, from which the current and voltage distribution across the winding were derived using electric field and equivalent circuit modelling. Analysis of the results has revealed that there are substantial dielectric stresses imposed on the winding insulation that can be attributed to a combination of three factors. (i) First, the

  7. Power distribution monitoring and control in the RBMK type reactors

    International Nuclear Information System (INIS)

    Emel'yanov, I.Ya.; Postnikov, V.V.; Volod'ko, Yu.I.

    1980-01-01

    Considered are the structures of monitoring and control systems for the RBMK-1000 reactor including three main systems with high independence: the control and safety system (CSS); the system for physical control of energy distribution (SPCED) as well as the Scala system for centralized control (SCC). Main functions and peculiarities of each system are discussed. Main attention is paid to new structural solutions and new equipment components used in these systems. Described are the RBMK operation software and routine of energy distribution control in it. It is noted that the set of reactor control and monitoring systems has a hierarchical structure, the first level of which includes analog systems (CSS and SPCED) normalizing and transmitting detector signals to the systems of the second level based on computers and realizing computer data processing, data representation to the operator, automatic (through CSS) control for energy distribution, diagnostics of equipment condition and local safety with provision for existing reserves with respect to crisis and thermal loading of fuel assemblies. The third level includes a power computer carrying out complex physical and optimization calculations and providing interconnections with the external computer of power system. A typical feature of the complex is the provision of local automatic safety of the reactor from erroneous withdrawal of any control rod. The complex is designed for complete automatization of energy distribution control in reactor in steady and transient operation conditions

  8. On-line computer control of a nuclear reactor using optimal control and state estimation methods

    International Nuclear Information System (INIS)

    Tye, C.

    1980-01-01

    This paper describes the experimental implementation of a nuclear reactor control system using combined optimal state feedback based on the Quadratic Regulator and state estimation using Kalman filtering techniques. The results obtained from the experiments indicate that a reactor control loop designed using this approach has improved stability margins, greater speed of response and noise filtering properties compared with a conventional reactor control loop. 11 refs

  9. Tokamak reactor systems studies

    International Nuclear Information System (INIS)

    1992-01-01

    A summary of work completed on the ARIES project during this report period is given. The main areas of effort were: neutronics, shield optimization and design, safety, systems, startup and shutdown, and ripple loss

  10. Modelling and control design for SHARON/Anammox reactor sequence

    DEFF Research Database (Denmark)

    Valverde Perez, Borja; Mauricio Iglesias, Miguel; Sin, Gürkan

    2012-01-01

    metabolism against fast chemical reaction and mass transfer. Likewise, the analysis of the dynamics contributed to establish qualitatively the requirements for control of the reactors, both for regulation and for optimal operation. Work in progress on quantitatively analysing different control structure......With the perspective of investigating a suitable control design for autotrophic nitrogen removal, this work presents a complete model of the SHARON/Anammox reactor sequence. The dynamics of the reactor were explored pointing out the different scales of the rates in the system: slow microbial...

  11. Nuclear reactor engineering: Reactor systems engineering. Fourth edition, Volume Two

    International Nuclear Information System (INIS)

    Glasstone, S.; Sesonske, A.

    1994-01-01

    This new edition of this classic reference combines broad yet in-depth coverage of nuclear engineering principles with practical descriptions of their application in the design and operation of nuclear power plants. Extensively updated, the fourth edition includes new materials on reactor safety and risk analysis, regulation, fuel management, waste management and operational aspects of nuclear power. This volume contains the following: the systems concept, design decisions, and information tools; energy transport; reactor fuel management and energy cost considerations; environmental effects of nuclear power and waste management; nuclear reactor safety and regulation; power reactor systems; plant operations; and advanced plants and the future

  12. MODELLING AND CONTROL OF CONTINUOUS STIRRED TANK REACTOR WITH PID CONTROLLER

    Directory of Open Access Journals (Sweden)

    Artur Wodołażski

    2016-09-01

    Full Text Available This paper presents a model of dynamics control for continuous stirred tank reactor (CSTR in methanol synthesis in a three-phase system. The reactor simulation was carried out for steady and transient state. Efficiency ratio to achieve maximum performance of the product per reactor unit volume was calculated. Reactor dynamics simulation in closed loop allowed to received data for tuning PID controller (proportional-integral-derivative. The results of the regulation process allow to receive data for optimum reactor production capacity, along with local hot spots eliminations or temperature runaway.

  13. 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.

  14. Knowledge-based systems and interactive graphics for reactor control using the Automated Reasoning Tool(ART) system

    International Nuclear Information System (INIS)

    Ragheb, M.; Clayton, B.; Davies, P.

    1987-01-01

    The use of Knowledge-Based systems and advanced graphic concepts are described using the Automated Reasoning Tool (ART) for a model nuclear plant system. Through the sue of asynchronous graphic input/output, the user is allowed to communicate through a graphical display to a Production-Rule Analysis System modelling the plant while its rules are actively being fired. The user changes the status of system components by pointing at them on the system configuration display with a mouse cursor and clicking one of the buttons on the mouse. The Production-Rule Analysis System accepts the new input and immediately displays its diagnosis of the system state and any associated recommendations as to the appropriate course of action. This approach offers a distinct advantage over typing the components statuses in response to queries by a conventional Production-Rule Analysis system. Moreover, two effective ways of communication between man and machine are combined

  15. Dynamic simulation platform to verify the performance of the reactor regulating system for a research reactor

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2015-07-01

    Digital instrumentation and controls system technique is being introduced in new constructed research reactor or life extension of older research reactor. Digital systems are easy to change and optimize but the validated process for them is required. Also, to reduce project risk or cost, we have to make it sure that configuration and control functions is right before the commissioning phase on research reactor. For this purpose, simulators have been widely used in developing control systems in automotive and aerospace industries. In these literatures, however, very few of these can be found regarding test on the control system of research reactor with simulator. Therefore, this paper proposes a simulation platform to verify the performance of RRS (Reactor Regulating System) for research reactor. This simulation platform consists of the reactor simulation model and the interface module. This simulation platform is applied to I and C upgrade project of TRIGA reactor, and many problems of RRS configuration were found and solved. And it proved that the dynamic performance testing based on simulator enables significant time saving and improves economics and quality for RRS in the system test phase. (authors)

  16. Fusion reactor control study. Volume 3. Tandem mirror reactors. Final report

    International Nuclear Information System (INIS)

    Chang, F.R.; DeCanio, F.; Fisher, J.L.; Madden, P.A.

    1982-03-01

    A study of the control requirements of the Tandem Mirror Reactor concept is reported. The study describes the development of a control simulator that is based upon a spatially averaged physics code of the reactor concept. The simulator portrays the evolution of the plasma through the complete reactor operating cycle; it includes models of the control and measurement system, thus allowing the exploration of various strategies for reactor control. Startup, shutdown, and control during the quasi-steady-state power producing phase were explored. Configurations are described which use a variety of control effectors including modulation of the refueling rate, beam current, and electron cyclotron resonance heating. Multivariable design techniques were used to design the control laws and compensators for the feedback controllers and presume the practical measurement of only a subset of the plasma and machine variables. Performance of the various controllers is explored using the nonlinear control simulator. Derivative control strategies using new or developed sensors and effectors appropriate to a power reactor environment are postulated, based upon the results of the control configurations tested. Research and development requirements for these controls are delineated

  17. Reactor protection and shut-down system

    International Nuclear Information System (INIS)

    Klar

    1980-01-01

    The reactor protection system being a part of the reactor safety system. The requirements on the reactor protection system are: high safety with regard to signal processing, high availability, self-reporting of faults etc. The functional sections of the reactor protection system are the analog section, the logic section and the generating of output signals. Description of the operation characteristics and of the extension of function. (orig.)

  18. Reactor feedwater pump control device

    International Nuclear Information System (INIS)

    Nishiyama, Hiroyuki.

    1990-01-01

    An amount of feedwater necessary for ensuring reactor inventory after scram is ensured automatically based on the reactor output before scram of a BWR type reactor. That is, if scram should occur, a feedwater flow rate just before the scram is stored by reactor output signals. Further, the amount of feedwater required after the scram is determined based on the output of the memory. The reactor power after the scram based on a feedwater flow rate and a main steam flow rate is inputted to an integrator, to calculate and output the amount of the feedwater flow rate (1) injected after the scram for the inventory. A coast down flowrate (2) in a case of pump trip is forecast by the output signals. Automatic trip is outputted to all turbine driving feedwater pumps when the sum of (1) and (2) exceeds a necessary and sufficient amount of feedwater required for ensuring inventory. For motor driving feedwater pumps, only a portion, for example, one of the pumps is automatically started while other pumps are stopped their operation, only in this case, to prevent excess water feeding. (I.S.)

  19. 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

  20. Permeation of hydrogen at low pressures through stainless steel and implications for tritium control in fusion reactor systems

    International Nuclear Information System (INIS)

    Axtmann, R.C.; Johnson, E.F.; Kuehler, C.W.

    1976-01-01

    New experimental data on the permeation of hydrogen through stainless steel indicate that at driving pressures below 10 -2 torr, the permeation rate is linearly dependent on the driving pressure. A possible consequence is that the permeation rates of hydrogenic species in fusion reactor systems might be much lower than those reported in contemporary conceptual design studies which assume that the rates are dependent on the square root of the driving pressure. The important implications of these low permeation rates are: (1) tritium losses to the environment may be more dependent on ordinary leaks from equipment than on permeation to the steam cycle; (2) recovery of tritium from breeding blankets via permeation windows may be impracticable; and (3) recovery of tritium from breeding blankets not dependent on permeation windows may be simplified by the possibility of operating at much higher average tritium concentrations in the blanket and cooling systems

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  2. 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.)

  3. Design, construction and implementation of two redundant circuits of the actuation logic of the protection system of the new control console of TRIGA Mark III reactor of ININ

    International Nuclear Information System (INIS)

    Celestino M, E.

    2016-01-01

    The Instituto Nacional de Investigaciones Nucleares (ININ) in Mexico has a nuclear reactor type TRIGA Mark III, which was put into operation in 1968. The reactor is used for staff training, radioisotope production, and for research projects of different areas. Over time and due to advances constantly has the electronics industry, maintenance of electronic systems is complicated because basically sometimes components that are no longer manufactured or no longer exist in the market, making it necessary to create projects required modernization. This is the case of the TRIGA reactor of ININ, so the Department of Automation and Instrumentation ININ is undertaking a new project to update the reactor control console. Systems that make up a nuclear reactor protection system (Ps) is relevant, since it is responsible for generating the necessary steps to shut down the reactor to an event of uncertainty which could affect the operators or the installation own actions. As part of the renovation project, this study design is presented to update the Logic of Action (La) of the Ps, whose final design must meet the requirements or specifications set by users and or regulations applicable to nuclear research reactors. One of the requirements established for the proposed new design La, is that it must be implemented with components and devices manufactured with latest technologies, and readily available on the market. The design which is operating currently uses TTL logic whose components are no longer available in the market, so for the new design you decide to use programmable circuits, and specifically, the CPLDs called (by the acronym Complex Programmable Logic Device). These CPLDs are electronic devices that solve complex logic equations and meeting the requirements of functionality and modernity for the new design of the La. In this work the criteria used for the selection of the CPLDs considering the availability and ease of software and hardware to use, and the design and

  4. MATLAB/SIMULINK model of CANDU reactor for control studies

    International Nuclear Information System (INIS)

    Javidnia, H.; Jiang, J.

    2006-01-01

    In this paper a MATLAB/SIMULINK model is developed for a CANDU type reactor. The data for the reactor are taken from an Indian PHWR, which is very similar to CANDU in its design. Among the different feedback mechanisms in the core of the reactor, only xenon has been considered which plays an important role in spatial oscillations. The model is verified under closed loop scenarios with simple PI controller. The results of the simulation show that this model can be used for controller design and simulation of the reactor systems. Adding models of the other components of a CANDU reactor would ultimately result in a complete model of CANDU plant in MATLAB/SIMULINK. (author)

  5. Adaptive nonlinear control for a research reactor

    International Nuclear Information System (INIS)

    Benitez R, J.S.

    1994-01-01

    Linearization by feedback of states is based on the idea of transform the nonlinear dynamic equation of a system in a linear form. This linear behavior can be achieve well in a complete way (input state) or in partial way (input output). This can be applied to systems of single or multiple inputs, and can be used to solve problems of stabilization and tracking of references trajectories. Comparing this method with conventional ones, linearization by feedback of states is exact in certain region of the space of state, instead of linear approximations of the equations in a certain point of the operation. In the presence of parametric uncertainties in the model of the system, the introduction of adaptive schemes provide a type toughness to the control system by nonlinear feedback, which gives as result the eventual cancellation of the nonlinear terms in the dynamic relationship between the output and the input of the auxiliary control. In the same way, it has been presented the design of a nonlinearizing control for the non lineal model of a TRIGA Mark III type reactor, with the aim of tracking a predetermined power profile. The asymptotic tracking of such profile is, at the present moment, in the stage of verification by computerized simulation the relative easiness in the design of auxiliary variable of control, as well as the decoupling action of the output variable, make very attractive the utilization of the method herein presented. (Author)

  6. Reactor alarm system development and application issues

    Energy Technology Data Exchange (ETDEWEB)

    Drexler, J E; Oicese, G O [INVAP S.E. (Argentina)

    1997-09-01

    The new hardware and software technologies, and the need in research reactors for assistance systems in operation and maintenance, have given an appropriate background to develop a computer based system named ``Reactor Alarm System`` (RAS). RAS is a software package, user oriented, with emphasis on production, experiments and maintenance goals. It is designed to run on distributed systems conformed with microcomputers under QNX operating system. RAS main features are: (a) Alarm Panel Display; (b) Alarm Page; (c) Alarm Masking and Inhibition; (d) Alarms Color and Attributes; (e) Condition Classification; and (f) Arrangement Presentation. RAS design allows it to be installed as a part of a computer based Supervision and Control System in new installations or retrofit existing reactor instrumentation systems. The analysis of human factors during development stage and successive user feedback from different applications, brought out several RAS improvements: (a) Multiple-copy alarm summaries; (b) Improved alarm handling; (c) Extended dictionary; and (d) Enhanced hardware availability. It has proved successful in providing new capabilities for operators, and also has shown the continuous increase of user-demands, reflecting the expectations placed today on computer-based systems. (author). 6 figs, 1 tabs.

  7. Reactor alarm system development and application issues

    International Nuclear Information System (INIS)

    Drexler, J.E.; Oicese, G.O.

    1997-01-01

    The new hardware and software technologies, and the need in research reactors for assistance systems in operation and maintenance, have given an appropriate background to develop a computer based system named ''Reactor Alarm System'' (RAS). RAS is a software package, user oriented, with emphasis on production, experiments and maintenance goals. It is designed to run on distributed systems conformed with microcomputers under QNX operating system. RAS main features are: a) Alarm Panel Display; b) Alarm Page; c) Alarm Masking and Inhibition; d) Alarms Color and Attributes; e) Condition Classification; and f) Arrangement Presentation. RAS design allows it to be installed as a part of a computer based Supervision and Control System in new installations or retrofit existing reactor instrumentation systems. The analysis of human factors during development stage and successive user feedback from different applications, brought out several RAS improvements: a) Multiple-copy alarm summaries; b) Improved alarm handling; c) Extended dictionary; and d) Enhanced hardware availability. It has proved successful in providing new capabilities for operators, and also has shown the continuous increase of user-demands, reflecting the expectations placed today on computer-based systems. (author). 6 figs, 1 tabs

  8. Dosimetry system of the RB reactor

    International Nuclear Information System (INIS)

    Lolic, B.; Vukadin, D.

    1962-01-01

    Although RB reactor is operated at very low power levels, safety and dosimetry systems have high importance. This paper shows detailed dosimetry system with fundamental typical components. Estimated radiation doses dependent on reactor power are given at some characteristic points in the rooms nearby reactor

  9. Control aid for xenon vibration in reactor

    International Nuclear Information System (INIS)

    Kanekawa, Takashi.

    1990-01-01

    In the present invention, the control operation for suppressing xenon vibrations in a reactor is aided for saving forecasting analysis and operator's skills. That is, parameters to be controlled for the suppression of xenon vibrations are power distribution, iodine distribution and xenon distribution. But what can be observed by operaters by the conventional fast overtone method is only the output distribution. In the present invention, the output level of the reactor core is always observed. Then, mathematical processings are conducted for the iodine distribution, the xenon distribution and the power distribution in the reactor core based on the histeresis of the parameters obtained by the measurement using physical constants and reactor design data. The xenon vibration control is aided by displaying the change with time of the distortion in axial direction. Accordingly, operators can always recognize the axial distortion of the power distribution, the iodine distribution and the xenon distribution. (I.S.)

  10. Nuclear reactor vessel decontamination systems

    International Nuclear Information System (INIS)

    McGuire, P. J.

    1985-01-01

    There is disclosed in the present application, a decontamination system for reactor vessels. The system is operatable without entry by personnel into the contaminated vessel before the decontamination operation is carried out and comprises an assembly which is introduced into the vertical cylindrical vessel of the typical boiling water reactor through the open top. The assembly includes a circular track which is centered by guideways permanently installed in the reactor vessel and the track guides opposed pairs of nozzles through which water under very high pressure is directed at the wall for progressively cutting and sweeping a tenacious radioactive coating as the nozzles are driven around the track in close proximity to the vessel wall. The whole assembly is hoisted to a level above the top of the vessel by a crane, outboard slides on the assembly brought into engagement with the permanent guideways and the assembly progressively lowered in the vessel as the decontamination operation progresses. The assembly also includes a low pressure nozzle which forms a spray umbrella above the high pressure nozzles to contain radioactive particles dislodged during the decontamination

  11. Reactor technology: power conversion systems and reactor operation and maintenance

    International Nuclear Information System (INIS)

    Powell, J.R.

    1977-01-01

    The use of advanced fuels permits the use of coolants (organic, high pressure helium) that result in power conversion systems with good thermal efficiency and relatively low cost. Water coolant would significantly reduce thermal efficiency, while lithium and salt coolants, which have been proposed for DT reactors, will have comparable power conversion efficiencies, but will probably be significantly more expensive. Helium cooled blankets with direct gas turbine power conversion cycles can also be used with DT reactors, but activation problems will be more severe, and the portion of blanket power in the metallic structure will probably not be available for the direct cycle, because of temperature limitations. A very important potential advantage of advanced fuel reactors over DT fusion reactors is the possibility of easier blanket maintenance and reduced down time for replacement. If unexpected leaks occur, in most cases the leaking circuit can be shut off and a redundant cooling curcuit will take over the thermal load. With the D-He 3 reactor, it appears practical to do this while the reactor is operating, as long as the leak is small enough not to shut down the reactor. Redundancy for Cat-D reactors has not been explored in detail, but appears feasible in principle. The idea of mobile units operating in the reactor chamber for service and maintenance of radioactive elements is explored

  12. Emergency system for nuclear reactors

    International Nuclear Information System (INIS)

    1976-01-01

    The invention concerns a circuit called 'of emergency help' intended to remove, in a safe and quick manner, the residual thermal power on the safety vessel of a fast neutron reactor cooled by a liquid metal flow, in the event of a failure occurring inside the main reactor vessel or on it. This system includes a network of spray nozzle tubes, distributed around and near the external surface of the safety vessel, to project on to the surface of the vessel a mist of a liquid having high latent vaporisation heat. The steam produced on contact with the safety vessel is collected in the space provided between the safety vessel and the external protection vessel by at least one collector pipe for dischaging this steam outside the vessel. Under a preferred design mode of the invention the liquid is water the use of which turns out to be particularly advantageous in practice owing to its favourable physical properties and its low cost [fr

  13. Fuel transfer system for a nuclear reactor

    International Nuclear Information System (INIS)

    Katz, L.R.; Marshall, J.R.; Desmarchais, W.E.

    1977-01-01

    Disclosed is a fuel transfer system for moving nuclear reactor fuel assemblies from a new fuel storage pit to a containment area containing the nuclear reactor, and for transferring spent fuel assemblies under water from the reactor to a spent fuel storage area. The system includes an underwater track which extends through a wall dividing the fuel building from the reactor containment and a car on the track serves as the vehicle for moving fuel assemblies between these two areas. The car is driven by a motor and linkage extending from an operating deck to a chain belt drive on the car. A housing pivotally mounted at its center on the car is hydraulically actuated to vertically receive a fuel assembly which then is rotated to a horizontal position to permit movement through the wall between the containment and fuel building areas. Return to the vertical position provides for fuel assembly removal and the reverse process is repeated when transferring an assembly in the opposite direction. Limit switches used in controlling operation of the system are designed to be replaced from the operating deck when necessary by tools designed for this purpose. 5 claims, 8 figures

  14. Support to the elaboration of the engineering of detail, configuration and programming of the control system of heat removal of the TRIGA Mark III reactor; Apoyo a la elaboracion de la ingenieria de detalle, configuracion y programacion del sistema de control de remocion de calor del reactor Triga Mark III

    Energy Technology Data Exchange (ETDEWEB)

    Diaz G, C. A.

    2016-07-01

    Nowadays, the peaceful and responsible use of nuclear energy in Mexico is of great importance and contributes to economic, social, scientist and technologic development in the country, highlighting the Instituto Nacional de Investigaciones Nucleares (ININ) and the Nuclear Power Plant of Laguna Verde as one of the most important dependences. Among the main facilities and laboratories of ININ is the Nuclear Research Reactor TRIGA Mark III, this is a pool type reactor with mobile core, cooled and moderated by light water and a flow of 1013 n/cm{sup 2}/sec. Due to the technological obsolescence is a growing problem that threatens the information, operation and/or efficacy of elements of control and safety systems of the reactor, these must be changed each time more frequently. In the modernization of reactor was used a Modicon M340 programmable logic control (PLC) and a Twido PLC for the control of heat removal system (Primary Cooling System (PCS) and Secondary Cooling System (SCS) respectively), this because the PLC has proven to be safe and effective devices, addition to reduce the wiring elements and increase the possibilities of performance and design of the digital control console. This document shows and describes the elements of heat removal system (PCS and SCS), and the signals and signal types that such items send or received by the PLC, likewise, is indicated the methodology used to develop the applications for the control of the Primary Cooling System and Secondary Cooling System, beginning with the PLC design, the development of PLC plans and the control logic, and finally, the simulation and debugging of applications on Unity Pro and Twido Suite. All this in compliance with the safety standards to nuclear research reactors (NS-R-4), the rules of industrial programming (IEC 61131-3), and the reactor operating limits postulated in the safety report and the software assurance system used in the ININ. (Author)

  15. Coordinate control of integral reactor based on single neuron PID controller

    International Nuclear Information System (INIS)

    Liu Yan; Xia Hong

    2014-01-01

    As one of the main type of reactors in the future, the development of the integral reactor has attracted worldwide attention. On the basis of understanding the background of the integral reactor, the author will be familiar with and master the power control of reactor and the feedwater flow control of steam generator, and the speed control of turbine (turbine speed control is associated with the turbine load control). According to the expectative program 'reactor power following turbine load' of the reactor, it will make coordinate control of the three and come to a overall control scheme. The author will use the supervisory learning algorithm of Hebb for single neuron PID controller with self-adaptation to study the coordinate control of integral reactor. Compared with conventional PI or PID controller, to a certain extent, it solves the problems that traditional PID controller is not easy to tune real-time parameters and lack of effective control for a number of complex processes and slow-varying parameter systems. It improves the security, reliability, stability and flexibility of control process and achieves effective control of the system. (authors)

  16. Design analysis and microprocessor based control of a nuclear reactor

    International Nuclear Information System (INIS)

    Sabbakh, N.J.

    1988-01-01

    The object of this thesis is to design and test a microprocessor based controller, to a simulated nuclear reactor system. The mathematical model that describes the dynamics of a typical nuclear reactor of one group of delayed neutrons approximations with temperature feedback was chosen. A digital computer program has been developed for the design and analysis of a simulated model based on the concept of state-variable feedback in order to meet a desired system response with maximum overshoot of 3.4% and setting time of 4 sec. The state variable feedback coefficients are designed for the continuous system, then an approximation is used to obtain in the state variable feedback vector for the discrete system. System control was implemented utilizing Direct Digital Control (DDC) of a nuclear reactor simulated model through a control algorithm that was performed by means of a microprocessor based system. The controller performance was satisfactorily tested by exciting the reactor system with a transient reactivity disturbance and by a step change in power demand. Direct digital control, when implemented on a microprocessor adds versatility, flexibility in system design with the added advantage of possible use of optimal control algorithms. 6 tabs.; 30 figs.; 46 refs.; 6 apps

  17. Nuclear reactor system for ABWR

    International Nuclear Information System (INIS)

    Miyano, Hiroshi; Kitagawa, Koji

    1997-01-01

    Various tests and measurements were performed during the pre-operational test run of Unit No. 6 of The Tokyo Electric Power Co., Inc.'s Kashiwazaki-Kariwa Nuclear Power Station, the first advanced boiling water reactor (ABWR) unit in the world, and the design and performance adequacy of the ABWR were confirmed. The realization of the ABWR in Japan took about 20 years. It was decided that technologies for the reactor internal pump (RIP) and the fine-motion control rod drive (FMCRD), which had been applied in Europe, would be incorporated in the ABWR aiming at simplification of its structure and operation. These main components were evaluated, modified and verified in consideration of the unique Japanese environment, such as seismic conditions, through a joint study program with Japanese utilities as well as an improvement and standardization program in cooperation with the government. In addition to incorporating RIP and FMCRD technologies, the ABWR also has improved features in terms of the design of the reactor pressure vessel and internals, as well as automated servicing equipment for the RIP, FMCRD, and primary containment vessel. (author)

  18. 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

  19. Development of telerobotic systems for reactor decommissioning, (3)

    International Nuclear Information System (INIS)

    Usui, Hozumi; Fujii, Yoshio; Shinohara, Yoshikuni

    1991-01-01

    This paper describes the telerobotic system for reactor decommissioning in the scope of engineering demonstration of dismantling radioactive reactor internals of an experimental boiling water power reactor JPDR. The total system consists of a telerobotic manipulator system equipped with a multi-functional amphibious slave manipulator with a load capacity of 25 daN, a chain-driven transport system, and a computer-assisted monitoring and control system. Preceding to the application of the telerobotic system to actual dismantling operation, a mockup test was performed of dismantling the simulated reactor internals of actual-size by the method of underwater plasma arc cutting in order to study the performance of the telerobotic system in a realistic environment. The system was then successfully applied to dismantling the actual reactor internals according to the JPDR decommissioning program. (author)

  20. Application of H∞ control theory to power control of a nonlinear reactor model

    International Nuclear Information System (INIS)

    Suzuki, Katsuo; Shimazaki, Junya; Shinohara, Yoshikuni

    1993-01-01

    The H∞ control theory is applied to the compensator design of a nonlinear nuclear reactor model, and the results are compared with standard linear quadratic Gaussian (LQG) control. The reactor model is assumed to be provided with a control rod drive system having the compensation of rod position feedback. The nonlinearity of the reactor model exerts a great influence on the stability of the control system, and hence, it is desirable for a power control system of a nuclear reactor to achieve robust stability and to improve the sensitivity of the feedback control system. A computer simulation based on a power control system synthesized by LQG control was performed revealing that the control system has some stationary offset and less stability. Therefore, here, attention is given to the development of a methodology for robust control that can withstand exogenous disturbances and nonlinearity in view of system parameter changes. The developed methodology adopts H∞ control theory in the feedback system and shows interesting features of robustness. The results of the computer simulation indicate that the feedback control system constructed by the developed H∞ compensator possesses sufficient robustness of control on the stability and disturbance attenuation, which are essential for the safe operation of a nuclear reactor

  1. Reactor power reduction system and method

    International Nuclear Information System (INIS)

    Bruno, S.J.; Dunn, S.A.; Raber, M.

    1978-01-01

    A method of operating a nuclear power reactor is disclosed which enables an accelerated power reduction of the reactor without completely shutting the reactor down. The method includes monitoring the incidents which, upon their occurrence, would require an accelerated power reduction in order to maintain the reactor in a safe operation mode; calculating the power reduction required on the occurrence of such an incident; determining a control rod insertion sequence for the normal operation of the reactor, said sequence being chosen to optimize reactor power capability; selecting the number of control rods necessary to respond to the accelerated power reduction demand, said selection being made according to a priority determined by said control rod insertion sequence; and inserting said selected control rods into the reactor core. 11 claims, 13 figures

  2. Control system design and validation platform development for small pressurized water reactors (SPWR) by coupling an engineering simulator and MATLAB/Simulink

    International Nuclear Information System (INIS)

    Sun, Peiwei; Zhao, Huanhuan; Liao, Longtao; Zhang, Jianmin; Su, Guanghui

    2017-01-01

    Highlights: • An SPWR control system design and validation platform is developed. • The platform is developed by coupling MATLAB/Simulink and an engineering simulator. • SPWR is modeled using Relap5 and preliminary control system is designed. • The platform is verified through numerical simulation over two typical load patterns. - Abstract: Significant progress has been made in the development of the small pressurized water reactors (SPWR). Unique characteristics of the SPWR deliver challenges to its control system design. In order to facilitate the control system design process and enhance its efficiency, it is important and necessary to establish a control system design and validation platform. Using shared memory technology, an engineering simulator coupled with MATLAB/Simulink is employed to achieve this objective. Shared memory is an efficient method to exchange data within programs. Dynamic data exchange and simulation time synchronization methods are particularly treated. To verify the platform, an SPWR with its control system is modeled using the platform and the simulator. Thermal-hydraulic modeling of the SPWR is carried out using Relap5, and its nodalization is introduced. The objectives of the control strategy are to maintain the average coolant temperature linearly varying with the reactor power and steam pressure constant. A preliminary SPWR control system is designed with proportional-integral-derivative (PID) controllers, and is implemented in MATLAB/Simulink associated with the engineering simulator. Subsequently, in order to evaluate the performance of the established simulation platform, transients of abrupt load changes and wide range load changes are simulated and simulation results are verified against those obtained from the engineering simulator alone. It is demonstrated that simulation results of both platforms are consistent with each other, which proves that the coupling of engineering simulator and MATLAB/Simulink is successful

  3. Recirculation system for nuclear reactors

    International Nuclear Information System (INIS)

    Braun, H. E.; Dollard, W. J.; Tower, S. N.

    1980-01-01

    A recirculation system for use in pressurized water nuclear reactors to increase the output temperature of the reactor coolant, thereby achieving a significant improvement in plant efficiency without exceeding current core design limits. A portion of the hot outlet coolant is recirculated to the inlets of the peripheral fuel assemblies which operate at relatively low power levels. The outlet temperature from these peripheral fuel assemblies is increased to a temperature above that of the average core outlet. The recirculation system uses external pumps and introduces the hot recirculation coolant to the free space between the core barrel and the core baffle, where it flows downward and inward to the inlets of the peripheral fuel assemblies. In the unlikely event of a loss of coolant accident, the recirculation system flow path through the free space and to the inlets of the fuel assemblies is utilized for the injection of emergency coolant to the lower vessel and core. During emergency coolant injection, the emergency coolant is prevented from bypassing the core through the recirculation system by check valves inserted into the recirculation system piping

  4. Inherent reactor power controller for a metal-fueled ALMR

    International Nuclear Information System (INIS)

    Wood, R.T.; Wilson, T.L. Jr.

    1990-01-01

    Inherent power control for metal-fueled ALMR designs involves using reactivity thermal feedback effects to control reactor power. This paper describes how, using classical control design techniques, a control system for normal load following maneuvers was deigned for a pool-type ALMR. This design provides active control of power removal in the balance of plant, direct control of selected primary and intermediate loop temperatures, and passive control of reactor power. The inherent stability of the strong, fast reactivity feedback effects bring heat production in the core into balance with the heat removal system temperatures, which are controlled to meet power demand. A simulation of the control system successfully responded to a 10% step change in power demand by changing power at an acceptable rate without causing large temperature fluctuations or exceeding thermal limits

  5. Safety Evaluation of Kartini Reactor Based on Instrumentation System Design

    International Nuclear Information System (INIS)

    Tjipta Suhaemi; Djen Djen Dj; Itjeu K; Johnny S; Setyono

    2003-01-01

    The safety of Kartini reactor has been evaluated based on instrumentation system aspect. The Kartini reactor is designed by BATAN. Design power of the reactor is 250 kW, but it is currently operated at 100 kW. Instrumentation and control system function is to monitor and control the reactor operation. Instrumentation and control system consists of safety system, start-up and automatic power control, and process information system. The linear power channel and logarithmic power channel are used for measuring power. There are 3 types of control rod for controlling the power, i.e. safety rod, shim rod, and regulating rod. The trip and interlock system are used for safety. There are instrumentation equipment used for measuring radiation exposure, flow rate, temperature and conductivity of fluid The system of Kartini reactor has been developed by introducing a process information system, start-up system, and automatic power control. It is concluded that the instrumentation of Kartini reactor has followed the requirement and standard of IAEA. (author)

  6. 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.)

  7. Computerized reactor pressure vessel materials information system

    International Nuclear Information System (INIS)

    Strosnider, J.; Monserrate, C.; Kenworthy, L.D.; Tether, C.D.

    1980-10-01

    A computerized information system for storage and retrieval of reactor pressure vessel materials data was established, as part of Task Action Plan A-11, Reactor Vessel Materials Toughness. Data stored in the system are necessary for evaluating the resistance of reactor pressure vessels to flaw-induced fracture. This report includes (1) a description of the information system; (2) guidance on accessing the system; and (3) a user's manual for the system

  8. Use of university research reactors to teach control engineering

    International Nuclear Information System (INIS)

    Bernard, J.A.

    1991-01-01

    University research reactors (URRs) have provided generations of students with the opportunity to receive instruction and do hands-on work in reactor dynamics, neutron scattering, health physics, and neutron activation analysis. Given that many URRs are currently converting to programmable control systems, the opportunity now exists to provide a similar learning experience to those studying systems control engineering. That possibility is examined here with emphasis on the need for the inclusion of experiment in control engineering curricula, the type of activities that could be performed, and safety considerations

  9. Limit regulation system for pressurized water nuclear reactors

    International Nuclear Information System (INIS)

    Aleite, W.; Bock, H.W.

    1976-01-01

    Described is a limit regulation system for a pressurized water nuclear reactor in combination with a steam generating system connected to a turbine, the nuclear reactor having control rods as well as an operational regulation system and a protective system, which includes reactor power limiting means operatively associated with the control rods for positioning the same and having response values between operating ranges of the operational regulation system, on the one hand, and response values of the protective system, on the other hand, and a live steam-minimal pressure regulation system cooperating with the reactor power limiting means and operatively connected to a steam inlet valve to the turbine for controlling the same

  10. Breeding description for fast reactors and symbiotic reactor systems

    International Nuclear Information System (INIS)

    Hanan, N.A.

    1979-01-01

    A mathematical model was developed to provide a breeding description for fast reactors and symbiotic reactor systems by means of figures of merit type quantities. The model was used to investigate the effect of several parameters and different fuel usage strategies on the figures of merit which provide the breeding description. The integrated fuel cycle model for a single-reactor is reviewed. The excess discharge is automatically used to fuel identical reactors. The resulting model describes the accumulation of fuel in a system of identical reactors. Finite burnup and out-of-pile delays and losses are treated in the model. The model is then extended from fast breeder park to symbiotic reactor systems. The asymptotic behavior of the fuel accumulation is analyzed. The asymptotic growth rate appears as the largest eigenvalue in the solution of the characteristic equations of the time dependent differential balance equations for the system. The eigenvector corresponding to the growth rate is the core equilibrium composition. The analogy of the long-term fuel cycle equations, in the framework of this model, and the neutron balance equations is explored. An eigenvalue problem adjoint to the one generated by the characteristic equations of the system is defined. The eigenvector corresponding to the largest eigenvalue, i.e. to the growth rate, represents the ''isotopic breeding worths.'' Analogously to the neutron adjoint flux it is shown that the isotopic breeding worths represent the importance of an isotope for breeding, i.e. for the growth rate of a system

  11. Reactor component inventory system at FFTF

    International Nuclear Information System (INIS)

    Ordonez, C.R.; Redekopp, R.D.; Reed, E.A.

    1985-02-01

    A reliable inventory control system was developed at the Fast Flux Test Facility (FFTF) to keep track of the occupancy of 900 refueling facility locations, to compile historical data on the movement of each reactor assembly, and to simulate assembly moves. The simulate capability is valuable because it allows verification of documents before they are issued for use in the plant, and eliminates the possibility of planning illegal or impossible moves. The system is installed on a UNIVAC 1100 computer and is maintained using a data base management system by Sperry Univac called MAPPER

  12. Automatic power control for a pressurized water reactor

    International Nuclear Information System (INIS)

    Hah, Yung Joon

    1994-02-01

    During a normal operation of a pressurized water reactor (PWR), the reactivity is controlled by control rods, boron, and the average temperature of the primary coolant. Especially in load follow operation, the reactivity change is induced by changes in power level and effects of xenon concentration. The control of the core power distribution is concerned, mainly, with the axial power distribution which depends on insertion and withdrawal of the control rods resulting in additional reactivity compensation. The utilization of part strength control element assemblies (PSCEAs) is quite appropriate for a control of the power distribution in the case of Yonggwang Nuclear Unit 3 (YGN Unit 3). However, control of the PSCEAs is not automatic, and changes in the boron concentration by dilution/boration are done manually. Thus, manual control of the PSCEAs and the boron concentration require the operator's experience and knowledge for a successful load follow operation. In this thesis, the new concepts have been proposed to adapt for an automatic power control in a PWR. One of the new concepts is the mode K control, another is a fuzzy power control. The system in mode K control implements a heavy-worth bank dedicated to axial shape control, independent of the existing regulating banks. The heavy bank provides a monotonic relationship between its motion and the axial power shape change, which allows automatic control of the axial power distribution. And the mode K enables precise regulation, by using double closed-loop control of the reactor coolant temperature and the axial power difference. Automatic reactor power control permits the nuclear power plant to accommodate the load follow operations, including frequency control, to respond to the grid requirements. The mode K reactor control concepts were tested using simulation responses of a Korean standardized 1000-MWe PWR which is a reference plant for the YGN Unit 3. The simulation results illustrate that the mode K would be

  13. Functional systems of a pressurized water reactor

    International Nuclear Information System (INIS)

    Heinzel, V.

    1982-01-01

    The main topics, discussed in the present paper, are: - Principle design of the reactor coolant system - reactor pressure vessel with internals - containment design - residual heat removal and emergency cooling systems - nuclear component cooling systems - emergency feed water systems - plant electric power supply system. (orig./RW)

  14. Nuclear reactor internals construction and failed fuel rod detection system

    International Nuclear Information System (INIS)

    Frisch, E.; Andrews, H.N.

    1976-01-01

    A system is provided for determining during operation of a nuclear reactor having fluid pressure operated control rod mechanisms the exact location of a fuel assembly with a defective fuel rod. The construction of the reactor internals is simplified in a manner to facilitate the testing for defective fuel rods and the reduce the cost of producing the upper internals of the reactor. 13 claims, 10 drawing figures

  15. 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)

  16. Reactor noise analysis applications in NPP I and C systems

    Energy Technology Data Exchange (ETDEWEB)

    Gloeckler, O. [International Atomic Energy Agency, Wagramer Strosse 5, A-1400 Vienna, Austria Ontario Power Generation, 230 Westney Road South, Ajax, Ont. L1S 7R3 (Canada)

    2006-07-01

    Reactor noise analysis techniques are used in many NPPs on a routine basis as 'inspection tools' to get information on the dynamics of reactor processes and their instrumentation in a passive, non-intrusive way. The paper discusses some of the tasks and requirements an NPP has to take to implement and to use the full advantages of reactor noise analysis techniques. Typical signal noise analysis applications developed for the monitoring of the reactor shutdown system and control system instrumentation of the Candu units of Ontario Power Generation and Bruce Power are also presented. (authors)

  17. Model Based Cyber Security Analysis for Research Reactor Protection System

    International Nuclear Information System (INIS)

    Sho, Jinsoo; Rahman, Khalil Ur; Heo, Gyunyoung; Son, Hanseong

    2013-01-01

    The study on the qualitative risk due to cyber-attacks into research reactors was performed using bayesian Network (BN). This was motivated to solve the issues of cyber security raised due to digitalization of instrumentation and control (I and C) system. As a demonstrative example, we chose the reactor protection system (RPS) of research reactors. Two scenarios of cyber-attacks on RPS were analyzed to develop mitigation measures against vulnerabilities. The one is the 'insertion of reactor trip' and the other is the 'scram halt'. The six mitigation measures are developed for five vulnerability for these scenarios by getting the risk information from BN

  18. Environment control system

    International Nuclear Information System (INIS)

    Sammarone, D.G.

    1978-01-01

    Disclosed is a system for controlling the environment of an enclosed area in nuclear reactor installations. The system permits the changing of the environment from nitrogen to air, or from air to nitrogen, without the release of any radioactivity or process gas to the outside atmosphere

  19. Reactivity requirements and safety systems for heavy water reactors

    International Nuclear Information System (INIS)

    Kati, S.L.; Rustagi, R.S.

    1977-01-01

    The natural uranium fuelled pressurised heavy water reactors are currently being installed in India. In the design of nuclear reactors, adequate attention has to be given to the safety systems. In recent years, several design modifications having bearing on safety, in the reactor processes, protective and containment systems have been made. These have resulted either from new trends in safety and reliability standards or as a result of feed-back from operating reactors of this type. The significant areas of modifications that have been introduced in the design of Indian PHWR's are: sophisticated theoretical modelling of reactor accidents, reactivity control, two independent fast acting systems, full double containment and improved post-accident depressurisation and building clean-up. This paper brings out the evolution of design of safety systems for heavy water reactors. A short review of safety systems which have been used in different heavy water reactors, of varying sizes, has been made. In particular, the safety systems selected for the latest 235 MWe twin reactor unit station in Narora, in Northern India, have been discussed in detail. Research and Development efforts made in this connection are discussed. The experience of design and operation of the systems in Rajasthan and Kalpakkam reactors has also been outlined

  20. Multilayer robust control for safety enhancement of reactor operations

    International Nuclear Information System (INIS)

    Edwards, R.M.; Lee, K.Y.; Ray, A.

    1991-01-01

    A novel concept of reactor power and temperature control has been recently reported in which a conventional output feedback controller is embedded within a state feedback setting. The embedded output feedback controller at the inner layer largely compensates for plant modeling uncertainties and external disturbances, and the outer layer generates an optimal control signal via feedback of the estimated plant states. A major advantage of this embedded architecture is the robustness of the control system relative to parametric and nonparametric uncertainties and thus the opportunity for designing fault-accommodating control algorithms to improve reactor operations and plant safety. The paper illustrates the architecture of the state-feedback-assisted classical (SFAC) control, which utilizes an embedded output feedback controller designed via classical techniques. It demonstrates the difference between the performance of conventional state feedback control and SFAC by examining the sensitivity of the dominant eigenvalues of the individual closed-loop systems

  1. Compliance of the Savannah River Plant P-Reactor cooling system with environmental regulations. Demonstrations in accordance with Sections 316(a) and (b) of the Federal Water Pollution Control Act of 1972

    International Nuclear Information System (INIS)

    Wilde, E.W.

    1985-12-01

    This document presents demonstrations under Sections 316(a) and (b) of the Federal Water Pollution Control Act of 1972 for the P-Reactor cooling system at the Savannah River Plant (SRP). The demonstrations were mandated when the National Pollution Discharge Elimination System (NPDES) permit for the SRP was renewed and the compliance point for meeting South Carolina Class B water quality criteria in the P-Reactor cooling system was moved from below Par Pond to the reactor cooling water outfall, No. P-109. Extensive operating, environmental, and biological data, covering most of the current P-Reactor cooling system history from 1958 to the present are discussed. No significant adverse effects were attributed to the thermal effluent discharged to Par Pond or the pumping of cooling water from Par Pond to P Reactor. It was conluded that Par Pond, the principal reservoir in the cooling system for P Reactor, contains balanced indigenous biological communities that meet all criteria commonly used in defining such communities. Par Pond compares favorably with all types of reservoirs in South Carolina and with cooling lakes and reservoirs throughout the southeast in terms of balanced communities of phytoplankton, macrophytes, zooplankton, macroinvertebrates, fish, and other vertebrate wildlife. The report provides the basis for negotiations between the South Carolina Department of Health and Environmental Control (SCDHEC) and the Department of Energy - Savannah River (DOE-SR) to identify a mixing zone which would relocate the present compliance point for Class B water quality criteria for the P-Reactor cooling system

  2. Modernization of control instrumentation and security of reactor IAN - R1

    International Nuclear Information System (INIS)

    Gonzalez, J. M.

    1993-01-01

    The program to modernize IAN-R1 research reactor control and safety instrumentation has been carried out considering two main aspects: updating safety philosophy requirements and acquiring the newest reactor control instrumentation controlled by computer, following the present criteria internationally recognized, for safety and reliable reactor operations and the latest developments of nuclear electronic technology. The new IAN-R1 reactor instrumentation consist of two wide range neutron monitoring channels, commanded by microprocessor a data acquisition system and reactor control, (controlled by computers). The reactor control desk is providing through two displays; all safety and control signals to the reactor operators; furthermore some signals like reactor power, safety and period signals are also showed on digital bar graphics, which are hard wired directly from the neutron monitoring channels

  3. Validation of reactor core protection system

    International Nuclear Information System (INIS)

    Lee, Sang-Hoon; Bae, Jong-Sik; Baeg, Seung-Yeob; Cho, Chang-Ho; Kim, Chang-Ho; Kim, Sung-Ho; Kim, Hang-Bae; In, Wang-Kee; Park, Young-Ho

    2008-01-01

    Reactor COre Protection System (RCOPS), an advanced core protection calculator system, is a digitized one which provides core protection function based on two reactor core operation parameters, Departure from Nucleate Boiling Ratio (DNBR) and Local Power Density (LPD). It generates a reactor trip signal when the core condition exceeds the DNBR or LPD design limit. It consists of four independent channels adapted a two-out-of-four trip logic. System configuration, hardware platform and an improved algorithm of the newly designed core protection calculator system are described in this paper. One channel of RCOPS was implemented as a single channel facility for this R and D project where we performed final integration software testing. To implement custom function blocks, pSET is used. Software test is performed by two methods. The first method is a 'Software Module Test' and the second method is a 'Software Unit Test'. New features include improvement of core thermal margin through a revised on-line DNBR algorithm, resolution of the latching problem of control element assembly signal and addition of the pre-trip alarm generation. The change of the on-line DNBR calculation algorithm is considered to improve the DNBR net margin by 2.5%-3.3%. (author)

  4. Nuclear reactor coolant and cover gas system

    International Nuclear Information System (INIS)

    George, J.A.; Redding, A.H.; Tower, S.N.

    1976-01-01

    A core cooling system is disclosed for a nuclear reactor of the type utilizing a liquid coolant with a cover gas above free surfaces of the coolant. The disclosed system provides for a large inventory of reactor coolant and a balanced low pressure cover gas arrangement. A flow restricting device disposed within a reactor vessel achieves a pressure of the cover gas in the reactor vessel lower than the pressure of the reactor coolant in the vessel. The low gas pressure is maintained over all free surfaces of the coolant in the cooling system including a coolant reservoir tank. Reactor coolant stored in the reservoir tank allows for the large reactor coolant inventory provided by the invention

  5. Cost benefit analysis of reactor safety systems

    International Nuclear Information System (INIS)

    Maurer, H.A.

    1984-01-01

    Cost/benefit analysis of reactor safety systems is a possibility appropriate to deal with reactor safety. The Commission of the European Communities supported a study on the cost-benefit or cost effectiveness of safety systems installed in modern PWR nuclear power plants. The following systems and their cooperation in emergency cases were in particular investigated in this study: the containment system (double containment), the leakage exhaust and control system, the annulus release exhaust system and the containment spray system. The benefit of a safety system is defined according to its contribution to the reduction of the radiological consequences for the environment after a LOCA. The analysis is so far performed in two different steps: the emergency core cooling system is considered to function properly, failure of the emergency core cooling system is assumed (with the possible consequence of core melt-down) and the results may demonstrate the evidence that striving for cost-effectiveness can produce a safer end result than the philosophy of safety at any cost. (orig.)

  6. Power controlling method for BWR type reactors

    International Nuclear Information System (INIS)

    Yoshida, Kenji.

    1983-01-01

    Purpose: To enable reactor operation exactly following after an aimed curve in the high power resuming and maintaining period without failures in cladding tubes. Method: Upon recovery of the reactor power to a high power level after changing the reactor power from the high power to the low power level, control rod is operated under such conditions that the linear power density after operation of the control rod does not exceed the PC envelope in the low power period, and the core flow rate is coordinated to the control rod operation. The linear power density can be suppressed within an allowable linear power density by the above operation during high power resuming and maintaining period and, as the result, PCI failures can be prevented. (Kamimura, M.)

  7. Primary cooling system for BWR type reactor

    International Nuclear Information System (INIS)

    Ibe, Eishi; Takahashi, Masanori; Aoki, Yasuko

    1993-01-01

    The present invention effectively uses information from a plurality of sensors in order to suppress corrosion circumstance of a nuclear reactor. That is, a predetermined general water quality factor at a predetermined position is determined as a standard index. A concentration of a water quality improver is controlled such that the index is within an aimed range. For this purpose, the entire sensor groups disposed in a primary coolant system of a nuclear reactor are divided into a plural systems of sensor groups each disposed on every different positions. Then, a predetermined sensor group (standard sensor group) is connected to a computing device and a data base so that it is always monitored for calculating and estimating the standard index. Only oxidative ingredient in water at the measuring point is noted, and a concentration distribution which agrees with an actually measured value of oxidative ingredients is extracted from data base and used as a correct concentration distribution. With such procedures, reactor water quality can be estimated accurately while compensating erroneous factors of individual sensors. Even when a new sensor is used, it is not necessary to greatly change control logic. (I.S.)

  8. Computation system for nuclear reactor core analysis

    International Nuclear Information System (INIS)

    Vondy, D.R.; Fowler, T.B.; Cunningham, G.W.; Petrie, L.M.

    1977-04-01

    This report documents a system which contains computer codes as modules developed to evaluate nuclear reactor core performance. The diffusion theory approximation to neutron transport may be applied with the VENTURE code treating up to three dimensions. The effect of exposure may be determined with the BURNER code, allowing depletion calculations to be made. The features and requirements of the system are discussed and aspects common to the computational modules, but the latter are documented elsewhere. User input data requirements, data file management, control, and the modules which perform general functions are described. Continuing development and implementation effort is enhancing the analysis capability available locally and to other installations from remote terminals

  9. An expert system for pressurized water reactor load maneuvers

    International Nuclear Information System (INIS)

    Chaung Lin; Jungping Chen; Yihjiunn Lin; Lianshin Lin

    1993-01-01

    Restartup after reactor shutdown and load-follow operations are the important tasks in operating pressurized water reactors. Generally, the most efficient method is to apply constant axial offset control (CAOC) strategy during load maneuvers. An expert system using CAOC strategy, fuzzy reasoning, a two-node core model, and operational constraints has been developed. The computation time is so short that this system, which leads to an approximate closed-loop control, could be useful for on-site calculation

  10. On some control problems of dynamic of reactor

    Science.gov (United States)

    Baskakov, A. V.; Volkov, N. P.

    2017-12-01

    The paper analyzes controllability of the transient processes in some problems of nuclear reactor dynamics. In this case, the mathematical model of nuclear reactor dynamics is described by a system of integro-differential equations consisting of the non-stationary anisotropic multi-velocity kinetic equation of neutron transport and the balance equation of delayed neutrons. The paper defines the formulation of the linear problem on control of transient processes in nuclear reactors with application of spatially distributed actions on internal neutron sources, and the formulation of the nonlinear problems on control of transient processes with application of spatially distributed actions on the neutron absorption coefficient and the neutron scattering indicatrix. The required control actions depend on the spatial and velocity coordinates. The theorems on existence and uniqueness of these control actions are proved in the paper. To do this, the control problems mentioned above are reduced to equivalent systems of integral equations. Existence and uniqueness of the solution for this system of integral equations is proved by the method of successive approximations, which makes it possible to construct an iterative scheme for numerical analyses of transient processes in a given nuclear reactor with application of the developed mathematical model. Sufficient conditions for controllability of transient processes are also obtained. In conclusion, a connection is made between the control problems and the observation problems, which, by to the given information, allow us to reconstruct either the function of internal neutron sources, or the neutron absorption coefficient, or the neutron scattering indicatrix....

  11. Nuclear reactor kinetics and plant control

    CERN Document Server

    Oka, Yoshiaki

    2013-01-01

    Understanding time-dependent behaviors of nuclear reactors and the methods of their control is essential to the operation and safety of nuclear power plants. This book provides graduate students, researchers, and engineers in nuclear engineering comprehensive information on both the fundamental theory of nuclear reactor kinetics and control and the state-of-the-art practice in actual plants, as well as the idea of how to bridge the two. The first part focuses on understanding fundamental nuclear kinetics. It introduces delayed neutrons, fission chain reactions, point kinetics theory, reactivit

  12. Control of operational transients in power reactors - Methodology

    International Nuclear Information System (INIS)

    Vukovic, D.

    1983-01-01

    By introducing the nuclear power stations in the electric power system, questions of their possibilities to satisfy system's demand arise. Control of operational transients (temperature and Xe 135 ) in power reactors by determining the optimal control rod strategy is given. Ti optimize the Xe 135 transients, the Pantryagin theorem of optimal processes is applied. For solving three dimensional, two-group diffusion equations the heterogeneous Feinberg-Galanin method with axial flux harmonics is adopted. An application of this formalism to three-dimensional, finite cylindrical pressurised water reactor radially reflected is presented. (author)

  13. Adaptive control method for core power control in TRIGA Mark II reactor

    Science.gov (United States)

    Sabri Minhat, Mohd; Selamat, Hazlina; Subha, Nurul Adilla Mohd

    2018-01-01

    The 1MWth Reactor TRIGA PUSPATI (RTP) Mark II type has undergone more than 35 years of operation. The existing core power control uses feedback control algorithm (FCA). It is challenging to keep the core power stable at the desired value within acceptable error bands to meet the safety demand of RTP due to the sensitivity of nuclear research reactor operation. Currently, the system is not satisfied with power tracking performance and can be improved. Therefore, a new design core power control is very important to improve the current performance in tracking and regulate reactor power by control the movement of control rods. In this paper, the adaptive controller and focus on Model Reference Adaptive Control (MRAC) and Self-Tuning Control (STC) were applied to the control of the core power. The model for core power control was based on mathematical models of the reactor core, adaptive controller model, and control rods selection programming. The mathematical models of the reactor core were based on point kinetics model, thermal hydraulic models, and reactivity models. The adaptive control model was presented using Lyapunov method to ensure stable close loop system and STC Generalised Minimum Variance (GMV) Controller was not necessary to know the exact plant transfer function in designing the core power control. The performance between proposed adaptive control and FCA will be compared via computer simulation and analysed the simulation results manifest the effectiveness and the good performance of the proposed control method for core power control.

  14. Human factors evaluation of the engineering test reactor control room

    International Nuclear Information System (INIS)

    Banks, W.W.; Boone, M.P.

    1981-03-01

    The Reactor and Process Control Rooms at the Engineering Test Reactor were evaluated by a team of human factors engineers using available human factors design criteria. During the evaluation, ETR, equipment and facilities were compared with MIL-STD-1472-B, Human Engineering design Criteria for Military Systems. The focus of recommendations centered on: (a) displays and controls; placing displays and controls in functional groups; (b) establishing a consistent color coding (in compliance with a standard if possible); (c) systematizing annunciator alarms and reducing their number; (d) organizing equipment in functional groups; and (e) modifying labeling and lines of demarcation

  15. Reactor shutdown system of prototype fast breeder reactor

    International Nuclear Information System (INIS)

    Govindarajan, S.; Singh, Om Pal; Kasinathan, N.; Paramasivan Pillai, C.; Arul, A.J.; Chetal, S.C.

    2002-01-01

    Full text: The shutdown system of PFBR is designed to assure a very high reliability by employing well known principles of redundancy, diversity and independence. The failure probability of the shutdown system limited to -6 / ry. Salient features of the shutdown system are: Two independent shutdown systems, each of them able to accommodate an additional single failure and made up of a trip system and an associated absorber rod group. Diversity between trip systems, rods and mechanisms. Initiation of SCRAM by two diverse physical parameters of the two shutdown systems for design events leading potentially to unacceptable conditions is the core. The first group of nine rods called control and safety rods (CSR) is used for both shutdown as well as power regulation. The second group consisting of three rods known as diverse safety rods (DSR) is used only for shutdown. Diversity between the two groups is ensured by varying the operating conditions of the electromagnets and the configurations of the mobile parts. The reactivity worth of the absorber rods have been chosen such that each group of rods would ensure cold shutdown on SCRAM even when the most reactive rod of the group fails to drop. Together the two groups ensure a shutdown margin of 5000 pcm. The speed and individual rod worth of the CSR is chosen from operational and safety considerations during reactor start up and raising of power. Required drop time of rods during SCRAM depends on the incident considered. For a severe reactivity incident of 3 $/s this has to be limited to 1s and is ensured by limiting electromagnet response time and facilitating drop by gravity. Design safety limits for core components have been determined and SCRAM parameters have been identified by plant dynamic analysis to restrict the temperatures of core components within the limits. The SCRAM parameters are distributed between the two systems appropriately. Fault tree analysis of the system has been carried out to determine the

  16. Artificial Intelligent Control for a Novel Advanced Microwave Biodiesel Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wali, W A; Hassan, K H; Cullen, J D; Al-Shamma' a, A I; Shaw, A; Wylie, S R, E-mail: w.wali@2009.ljmu.ac.uk [Built Environment and Sustainable Technologies Institute (BEST), School of the Built Environment, Faculty of Technology and Environment Liverpool John Moores University, Byrom Street, Liverpool L3 3AF (United Kingdom)

    2011-08-17

    Biodiesel, an alternative diesel fuel made from a renewable source, is produced by the transesterification of vegetable oil or fat with methanol or ethanol. In order to control and monitor the progress of this chemical reaction with complex and highly nonlinear dynamics, the controller must be able to overcome the challenges due to the difficulty in obtaining a mathematical model, as there are many uncertain factors and disturbances during the actual operation of biodiesel reactors. Classical controllers show significant difficulties when trying to control the system automatically. In this paper we propose a comparison of artificial intelligent controllers, Fuzzy logic and Adaptive Neuro-Fuzzy Inference System(ANFIS) for real time control of a novel advanced biodiesel microwave reactor for biodiesel production from waste cooking oil. Fuzzy logic can incorporate expert human judgment to define the system variables and their relationships which cannot be defined by mathematical relationships. The Neuro-fuzzy system consists of components of a fuzzy system except that computations at each stage are performed by a layer of hidden neurons and the neural network's learning capability is provided to enhance the system knowledge. The controllers are used to automatically and continuously adjust the applied power supplied to the microwave reactor under different perturbations. A Labview based software tool will be presented that is used for measurement and control of the full system, with real time monitoring.

  17. Artificial Intelligent Control for a Novel Advanced Microwave Biodiesel Reactor

    International Nuclear Information System (INIS)

    Wali, W A; Hassan, K H; Cullen, J D; Al-Shamma'a, A I; Shaw, A; Wylie, S R

    2011-01-01

    Biodiesel, an alternative diesel fuel made from a renewable source, is produced by the transesterification of vegetable oil or fat with methanol or ethanol. In order to control and monitor the progress of this chemical reaction with complex and highly nonlinear dynamics, the controller must be able to overcome the challenges due to the difficulty in obtaining a mathematical model, as there are many uncertain factors and disturbances during the actual operation of biodiesel reactors. Classical controllers show significant difficulties when trying to control the system automatically. In this paper we propose a comparison of artificial intelligent controllers, Fuzzy logic and Adaptive Neuro-Fuzzy Inference System(ANFIS) for real time control of a novel advanced biodiesel microwave reactor for biodiesel production from waste cooking oil. Fuzzy logic can incorporate expert human judgment to define the system variables and their relationships which cannot be defined by mathematical relationships. The Neuro-fuzzy system consists of components of a fuzzy system except that computations at each stage are performed by a layer of hidden neurons and the neural network's learning capability is provided to enhance the system knowledge. The controllers are used to automatically and continuously adjust the applied power supplied to the microwave reactor under different perturbations. A Labview based software tool will be presented that is used for measurement and control of the full system, with real time monitoring.

  18. A stereoscopic television system for reactor inspection

    International Nuclear Information System (INIS)

    Friend, D.B.; Jones, A.

    1980-03-01

    A stereoscopic television system suitable for reactor inspection has been developed. Right and left eye views, obtained from two conventional black and white cameras, are displayed by the anaglyph technique and observers wear appropriately coloured viewing spectacles. All camera functions, such as zoom, focus and toe-in are remotely controlled. A laboratory experiment is described which demonstrates the increase in spatial awareness afforded by the use of stereo television and illustrates its potential in the supervision of remote handling tasks. Typical depth resolutions of 3mm at 1m and 10mm at 2m have been achieved with the reactor instrument. Trials undertaken during routine inspection at Oldbury Power Station in June 1978 are described. They demonstrate that stereoscopic television can indeed improve the convenience of remote handling and that the added display realism is beneficial in visual inspection. (author)

  19. Passive systems for light water reactors

    International Nuclear Information System (INIS)

    Adinolfi, R.; Noviello, L.

    1990-01-01

    The paper reviews the most original concepts that have been considered in Italy for the back-fitting of the nuclear power plants in order to reduce the probability and the importance of the release to the environment in case of a core melt. With reference either to BWR or PWR, passive concepts have been considered for back-fitting in the following areas: pump seals damage prevention and ECCS passive operation; reactor passive depressurization; molten reactor core passive cooling; metal containment passive water cooling through a water tank located at high level; containment isolation improvement through a sealing system; containment leaks control and limitation of environmental release. In addition some considerations will be made on the protection against external events introduced from the beginning on the PUN design either on building and equipment lay-out either on structure design. (author). 5 figs

  20. Monitoring system for accuracy and reliability characteristics of standard temperature measurements in WWER-440 reactors

    International Nuclear Information System (INIS)

    Stanc, S.; Repa, M.

    2001-01-01

    Description of a monitoring system for accuracy and reliability characteristics of standard temperature measurements in WWER-440 reactors and benefits obtained from its use are shown in the presentation. As standard reactor temperature measurement, coolant temperature measurement at fuel assembly outlets and in loops, entered into the In-Reactor Control System , are considered. Such systems have been implemented at two V-230 reactors and are under implementation at other four V-213 reactors. (Authors)