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Sample records for reactor diagnostic system

  1. Expert system for fast reactor diagnostic

    International Nuclear Information System (INIS)

    Parcy, J.P.

    1982-09-01

    A general description of expert systems is given. The operation of a fast reactor is reviewed. The expert system to the diagnosis of breakdowns limited to the reactor core. The structure of the system is described: specification of the diagnostics; structure of the data bank and evaluation of the rules; specification of the prediagnostics and evaluation; explanation of the diagnostics; time evolution of the system; comparison with other expert systems. Applications to some cases of faults are finally presented [fr

  2. Development of Vibration Diagnostic System in Research Reactors

    International Nuclear Information System (INIS)

    EL-Kafas, A. A.

    1999-01-01

    Early failure detection and diagnosis system are an important group with increasing interest with the operating support system. Already existing system to monitor integrity of primary system components are vibration and acoustic monitoring system (2,3). The development of vibration diagnostic system for MARIA reactor (30 MW)-the second research reactor in Poland -was made. The new system is applied for the Egypt research reactor (ETRR-1). This paper represents the result obtained during the operation of this activity that carried out at MARIA and ETRR-1 reactors

  3. Online monitoring and diagnostic system on RA-6 nuclear reactor

    International Nuclear Information System (INIS)

    Garcia Peyrano, O. A.; Marticorena, M.; Koch, R. G.; Martinez, J. S; Berruti, G. E.; Nunez, W. M.; Gonzales, L. A.; Tarquini, L. D.; Sotelo, J. P

    2009-01-01

    This paper presents the Online Automatic Monitoring and Diagnostic System for mechanical components, installed on RA-6 Nuclear Reactor (San Carlos de Bariloche, Argentina). This system has been designed, installed and set-up by the Vibrations and Mechatronics Laboratory (Centro Atomico Bariloche, Comision Nacional de Energia Atomica) and Sitrack.com Argentina SA. This system provides an online mechanical diagnostic of the main reactor components, allowing incipient failures to be early detected and identified, avoiding unscheduled shut-downs and reducing maintenance times. The diagnostic is accomplished by an online analysis of the vibratory signature of the mechanical components, obtained by vibrations sensors on the main pump and the decay tank. The mechanical diagnostic and the main operational parameters are displayed on the reactor control room and published on the internet. [es

  4. Diagnostic system for primary circuits of pressurized-water reactors

    International Nuclear Information System (INIS)

    Liska, J.; Majer, J.

    1983-01-01

    The diagnostic system monitors the reactor, the main circulating pipe, the main circulating pump, the main shut-off valve, the steam generator and the pressurizer. Diagnostic signals are obtained from the sensors designed for operation measurements and from sensors for special diagnostic purposes. The following operations are carried out: detection of dangerous dynamic stress of components, detection of damage to functional surfaces of components, detection of occurrence and propagation of defects in component materials, detection of loose particles and foreign bodies, detection of coolant leakage, detection of coolant boiling in the core and detection of impermissible non-homogeneities of fields of physical quantities in the core. The diagnostic system comprises: monitoring, classification of properly investigated effects, periodical tracing and long-term tracing. The operational diagnostics system developed by the SKODA Concern consists of a vibration monitoring system, a spectral analysis system and a central evaluation system. (M.D.)

  5. Reactor protection system with automatic self-testing and diagnostic

    International Nuclear Information System (INIS)

    Gaubatz, D.C.

    1996-01-01

    A reactor protection system is disclosed having four divisions, with quad redundant sensors for each scram parameter providing input to four independent microprocessor-based electronic chassis. Each electronic chassis acquires the scram parameter data from its own sensor, digitizes the information, and then transmits the sensor reading to the other three electronic chassis via optical fibers. To increase system availability and reduce false scrams, the reactor protection system employs two levels of voting on a need for reactor scram. The electronic chassis perform software divisional data processing, vote 2/3 with spare based upon information from all four sensors, and send the divisional scram signals to the hardware logic panel, which performs a 2/4 division vote on whether or not to initiate a reactor scram. Each chassis makes a divisional scram decision based on data from all sensors. Automatic detection and discrimination against failed sensors allows the reactor protection system to automatically enter a known state when sensor failures occur. Cross communication of sensor readings allows comparison of four theoretically ''identical'' values. This permits identification of sensor errors such as drift or malfunction. A diagnostic request for service is issued for errant sensor data. Automated self test and diagnostic monitoring, sensor input through output relay logic, virtually eliminate the need for manual surveillance testing. This provides an ability for each division to cross-check all divisions and to sense failures of the hardware logic. 16 figs

  6. Reactor accident diagnostic expert system: DISKET

    International Nuclear Information System (INIS)

    Yoshida, Kazuo; Yokobayashi, Masao

    1989-11-01

    A reactor accident diagnostic system DISKET has been developed to identify the cause and the type of an abnormal transient of a nuclear power plant. The system is based on the knowledge engineering and consists of an inference engine IERIAS and a knowledge base. The main features of DISKET are the following: Time-varying characteristics of transient can be treated and knowledge base can be divided into several knowledge units to handle a lot of rules effectively. This report has been provided for the convenience of DISKET's users and consists of three parts. The first part is the description of the whole system, the details of the knowledge base of DISKET are described in the second part, and how to use the DISKET system is explained in the third part. (author)

  7. Conceptual design of neutron diagnostic systems for fusion experimental reactor

    International Nuclear Information System (INIS)

    Iguchi, T.; Kaneko, J.; Nakazawa, M.

    1994-01-01

    Neutron measurement in fusion experimental reactors is very important for burning plasma diagnostics and control, monitoring of irradiation effects on device components, neutron source characterization for in-situ engineering tests, etc. A conceptual design of neutron diagnostic systems for an ITER-like fusion experimental reactor has been made, which consists of a neutron yield monitor, a neutron emission profile monitor and a 14-MeV spectrometer. Each of them is based on a unique idea to meet the required performances for full power conditions assumed at ITER operation. Micro-fission chambers of 235 U (and 238 U) placed at several poloidal angles near the first wall are adopted as a promising neutron yield monitor. A collimated long counter system using a 235 U fission chamber and graphite neutron moderators is also proposed to improve the calibration accuracy of absolute neutron yield determination

  8. In-operation diagnostic system for reactor coolant pump

    International Nuclear Information System (INIS)

    Sugiyama, Mitsunobu; Hasegawa, Ichiro; Kitahara, Hiromichi; Shimamura, Kazuo; Yasuda, Chiaki; Ikeda, Yasuhiro; Kida, Yasuo.

    1996-01-01

    A reactor coolant pump (RCP) is one of the most important rotating machines in the primary loop nuclear power plants. To improve the reliability and of nuclear power plants, a new diagnostic system that enables early detection of RCP faults has been developed. This system is based on continuous monitoring of vibration and other process data. Vibration is an important indicator of mechanical faults providing information on physical phenomena such as changes in dynamic characteristics and excitation forces changes that signal failure or incipient failure. This new system features comparative vibration analysis and simulation to anticipate equipment failure. (author)

  9. Regression models of reactor diagnostic signals

    International Nuclear Information System (INIS)

    Vavrin, J.

    1989-01-01

    The application is described of an autoregression model as the simplest regression model of diagnostic signals in experimental analysis of diagnostic systems, in in-service monitoring of normal and anomalous conditions and their diagnostics. The method of diagnostics is described using a regression type diagnostic data base and regression spectral diagnostics. The diagnostics is described of neutron noise signals from anomalous modes in the experimental fuel assembly of a reactor. (author)

  10. Towards diagnostics for a fusion reactor

    International Nuclear Information System (INIS)

    Costley, A. E.

    2009-01-01

    The requirements for measurements on modern tokamak fusion plasmas are outlined, and the techniques and systems used to make the measurements, usually referred to as 'diagnostics', are introduced. The basics of three particular diagnostics - magnetics, neutron systems and a laser based optical system - are outlined as examples of modern diagnostic systems, and the implementation of these diagnostics on a current tokamak (JET) are described. The next major step in magnetic confinement fusion is the construction and operation of the International Thermonuclear Experimental Reactor (ITER), which is a joint project of China, Europe, Japan, India, Korea, the Russian Federation, and the United States. Construction has begun in Cadarache, France. It is expected that ITER will operate at the 500 MW level. Because of the harsh environment in the vacuum vessel where many diagnostic components are located, the development of diagnostics for ITER is a major challenge - arguably the most difficult challenge ever undertaken in the field of diagnostics. The main elements in the diagnostic step are outlined using the three chosen techniques as examples. Finally, the step beyond ITER to a demonstration reactor, DEMO, that is expected to produce several GWs of fusion power is considered and the impact on diagnostics outlined. It is shown that the applicability and development steps needed for the individual diagnostics techniques will differ. The challenges for DEMO diagnostics are substantial and a dedicated effort should be made to find and develop new techniques, and especially techniques appropriate to the DEMO environment. It is argued that the limitations and difficulties in diagnostics should be a consideration in the optimization and designs of DEMO. (author)

  11. evelopment of a boiling water reactor fault diagnostic system with a signed directed graph method

    International Nuclear Information System (INIS)

    Chen, M.; Yu, C.C.; Liou, C.T.; Liao, L.Y.

    1990-01-01

    The fault diagnostic system for a nuclear power reactor is expected to be a useful decision support system for the operators during transients and accident conditions. A considerable research effort has been devoted to the development of automated fault diagnostic systems. One major approach, which has been widely used in chemical engineering, is to identify the possible causes of process disturbance using a logic-oriented method called signed directed graph (SDG). A knowledge based system was developed with the rules derived from the SDG representation. The SDG for the Chinshan nuclear power plant, which is a typical boiling water reactor, is established. The personal consultant system is used as the expert system development tool in this paper

  12. Upgrading the reactor noise diagnostic systems at the Paks NPP

    International Nuclear Information System (INIS)

    Czibok, T.; Dezsoe, Z.; Kiss, K.; Krinizs, K.; Lipcsei, S.

    2002-01-01

    The paper reports on the actual step in upgrading process of the reactor noise diagnostic systems at Paks NPP. This step has mainly a technical character. Renewal of facilities for signal conditioning and for data acquisition is going on. Autonomous systems at each of the four reactor units will be able to acquire a set of data series which can be arbitrarily chosen from the whole set of several hundred in-core neutron and other signals. The autonomous systems can be remotely controlled by a central computer through the local network. Modularity and extensibility are important features of the new systems: the size of the set of available signals can be extended and new modules for more advanced evaluations can be installed later. Present plans for system hardware upgrading are outlined, together with some technical details of measurement control, data acquisition moduls and network communication.(abstract)

  13. Constitution and application of reactor make-up system's fault diagnostic Bayesian networks

    International Nuclear Information System (INIS)

    Liang Jie; Cai Qi; Chu Zhuli; Wang Haiping

    2013-01-01

    A fault diagnostic Bayesian network of reactor make-up system was constituted. The system's structure characters, operation rules and experts' experience were combined and an initial net was built. As the fault date sets were learned with the particle swarm optimization based Bayesian network structure, the structure of diagnostic net was completed and used to inference case. The built net can analyze diagnostic probability of every node in the net and afford assistant decision to fault diagnosis. (authors)

  14. observer-based diagnostics and monitoring of vibrations in nuclear reactor core cooling system

    International Nuclear Information System (INIS)

    Siry, S.A K.

    2007-01-01

    analysis and diagnostics of vibration in industrial systems play a significant rule to prevent severe severe damages . drive shaft vibration is a complicated phenomenon composed of two independent forms of vibrations, translational and torsional. translational vibration measurements in case of the reactor core cooling system are introduced. the system under study consists of the three phase induction motor, flywheel, centrifugal pump, and two coupling between motor-flywheel, and flywheel-pump. this system structure is considered to be one where the blades are pegged into the discs fitting into the shafts. a non-linear model to simulate vibration in the reactor core cooling system will be introduced. simulation results of an operating reactor core cooling system using the actual parameters will be presented to validate the accuracy and reliability of the proposed analytical method the accuracy in analyzing the results depends on the system model. the shortcomings of the conventional model will be avoided through the use of that accurate nonlinear model which improve the simulation of the reactor core cooling system

  15. Artificial intelligence applications to nuclear reactor diagnostics

    International Nuclear Information System (INIS)

    Lee, J.C.; Hassberger, J.A.; Wehe, D.K.

    1987-01-01

    The authors research into applications of artificial intelligence to nuclear reactor diagnostics involves three main areas. In the first area, the authors combine reactor simulation models and expert systems to diagnose the state of the plant. The second area examines ways in which the rule or knowledge base of an intelligent controller can be generated systematically from either fault trees or acquired plant data. Third, efforts are described to develop the capabilities to validate these techniques in a realistic reactor setting. The techniques are applicable to all reactor types, including fast reactors

  16. IERIAS: inference engine for reactor accident diagnostic system using knowledge engineering technique

    International Nuclear Information System (INIS)

    Yokobayashi, Masao; Yoshida, Kazuo; Kohsaka, Atsuo; Yamamoto, Minoru.

    1984-11-01

    This report describes an inference engine IERIAS which has been devoloped for a diagnostic system to identify the cause and type of an abnormal transient of a reactor plant. This system using knowledge engineering technique consists of a knowledge base and an inference engine. The inference engine IERIAS is designed so as to treat time-varying data of a plant. The major features of IERIAS are ; (1) histroy of transients can be treated, (2) knowledge base can be divided into some knowledge units, (3) program language UTILISP is used which is suitable for symbolic data manipulation. Inference was made using IERIAS with a knowledge base which was created from simulated results of various transients by a PWR plant simulator. The results showed a good applicability of IERIAS for reactor diagnosis. (author)

  17. Application of noise analysis methods in nuclear reactor diagnostics

    International Nuclear Information System (INIS)

    Dach, K.

    1985-01-01

    By statistical evaluation of the fluctuation component of signals from selected detectors, noise diagnostics detects conditions of equipment which might later result in failure. The objective of early diagnostics is to detect the failed integrity of primary circuit components, failed detectors or anomalies of the thermohydraulic process. The commonest method of experimental data analysis is spectral analysis in the frequency range 0 to 50 Hz. Recently, expert diagnostic systems have been built based on artificial intelligence systems. Czechoslovakia participates in the experimental research of noise diagnostics in the context of the development of diagnostic assemblies for WWER-440 reactors. (M.D.)

  18. Expert system applications in support of system diagnostics and prognostics at EBR-II

    International Nuclear Information System (INIS)

    Lehto, W.K.; Gross, K.C.

    1989-01-01

    Expert systems have been developed to aid in the monitoring and diagnostics of the Experimental Breeder Reactor-II (EBR-II) at the Idaho National Engineering Laboratory (INEL) in Idaho Falls, Idaho. Systems have been developed for failed fuel surveillance and diagnostics and reactor coolant pump monitoring and diagnostics. A third project is being done jointly by ANL-W and EG ampersand G Idaho to develop a transient analysis system to enhance overall plant diagnostic and prognostic capability. The failed fuel surveillance and diagnosis system monitors, processes, and interprets information from nine key plant sensors. It displays to the reactor operator diagnostic information needed to make proper decisions regarding technical specification conformance during reactor operation with failed fuel. 8 refs., 9 figs., 2 tabs

  19. A diagnostic expert system for a boiling water reactor using a dynamic model

    International Nuclear Information System (INIS)

    Sonoda, Y.; Kanemoto, S.; Imaruoka, H.

    1990-01-01

    A diagnostic expert system for abnormal disturbances in a BWR (Boiling Water Reactor) plant has been developed. The peculiar feature of this system is a diagnostic method which combines artificial intelligence technique with numerical analysis technique. The system has three diagnostic functions, 1) identification of anomaly position (device or sensor), 2) identification of anomaly mode and 3) identification of anomaly cause. Function 1) is implemented as follows. First, a hypothesis about anomaly propagation paths is built up by qualitative reasoning, using knowledge of causal relations among observed signals. Next, the abnormal device or sensor is found by applying model reference method and fuzzy set theory to test the hypothesis, using knowledge of plant structure and function, heuristic strategy of diagnosis and module type dynamic simulator. This simulator is composed of basic transfer function modules. The simulation model for the testing region is built up automatically, according to the requirement from the diagnostic task. Function 2) means identification of dynamic characteristics for an anomaly. It is realized by tuning model parameters so as to reproduce the abnormal signal behavior using the non-linear programing method. Function 3) derives probable anomaly causes from heuristic rules between anomaly mode and cause. A basic plant dynamic model was built up and adjusted to dynamic characteristics for one BWR plant (1100MWe). In order to verify the diagnostic functions of this system, data for several abnormal events was compiled by modifying this model. The diagnostic functions were proved useful, through the simulated abnormal data

  20. Reactor noise diagnostics based on multivariate autoregressive modeling: Application to LOFT [Loss-of-Fluid-Test] reactor process noise

    International Nuclear Information System (INIS)

    Gloeckler, O.; Upadhyaya, B.R.

    1987-01-01

    Multivariate noise analysis of power reactor operating signals is useful for plant diagnostics, for isolating process and sensor anomalies, and for automated plant monitoring. In order to develop a reliable procedure, the previously established techniques for empirical modeling of fluctuation signals in power reactors have been improved. Application of the complete algorithm to operational data from the Loss-of-Fluid-Test (LOFT) Reactor showed that earlier conjectures (based on physical modeling) regarding the perturbation sources in a Pressurized Water Reactor (PWR) affecting coolant temperature and neutron power fluctuations can be systematically explained. This advanced methodology has important implication regarding plant diagnostics, and system or sensor anomaly isolation. 6 refs., 24 figs

  1. Data handling at EBR-II [Experimental Breeder Reactor II] for advanced diagnostics and control work

    International Nuclear Information System (INIS)

    Lindsay, R.W.; Schorzman, L.W.

    1988-01-01

    Improved control and diagnostics systems are being developed for nuclear and other applications. The Experimental Breeder Reactor II (EBR-II) Division of Argonne National Laboratory has embarked on a project to upgrade the EBR-II control and data handling systems. The nature of the work at EBR-II requires that reactor plant data be readily available for experimenters, and that the plant control systems be flexible to accommodate testing and development needs. In addition, operational concerns require that improved operator interfaces and computerized diagnostics be included in the reactor plant control system. The EBR-II systems have been upgraded to incorporate new data handling computers, new digital plant process controllers, and new displays and diagnostics are being developed and tested for permanent use. In addition, improved engineering surveillance will be possible with the new systems

  2. Clustering methods and visualization algorithms to aid nuclear reactor operative diagnostics

    International Nuclear Information System (INIS)

    Pepelyshev, Yu.N.; Dzwinel, W.

    1990-01-01

    The software system developed plays the role of the aid to an operator for nuclear reactor diagnostics. The noise analysis of the reactor parameters such as power, temperature and coolant flow rate constitutes the basis of the system. Combination of data acquisition, data preprocessing, clustering and cluster visualization algorithms with heuristic techniques of results analysis, determine the way of its implementation. Two regimes are available. The first one - extended - is recommended for a long term investigations and the second - suppressed for the aid to the reactor operation monitoring. The system has been tested and developed at the JINR IBR-2 pulsed reactor. 13 refs.; 4 figs.; 2 tabs

  3. Vacuum system design and tritium inventory for the charge exchange diagnostic on the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Medley, S.S.

    1986-01-01

    The application of charge exchange analyzers for the measurement of ion temperature in fusion plasma experiments requires a direct connection between the diagnostic and plasma-discharge vacuum chambers. Differential pumping of the gas load from the diagnostic stripping cell operated at > or approx. = 10 -3 Torr is required to maintain the analyzer chamber at a pressure of -6 Torr. The migration of gases between the diagnostic and plasma vacuum chambers must be minimized. In particular, introduction of the analyzer stripping cell gas into the plasma chamber having a base pressure of -8 Torr must be suppressed. The charge exchange diagnostic for the Tokamak Fusion Test Reactor (TFTR) is comprised of two analyzer systems designed to contain a total of 18 independent mass/energy analyzers and one diagnostic neutral beam rated at 80 keV, 15 A. The associated arrays of multiple, interconnected vacuum systems were analyzed using the Vacuum System Transient Simulator (Vsts) computer program which models the transient transport of multigas species through complex networks of ducts, valves, traps, vacuum pumps, and other related vacuum system components. In addition to providing improved design performance at reduced costs, the analysis yields estimates for the exchange of tritium from the torus to the diagnostic components and of the diagnostic working gases to the torus

  4. Development of reactor accident diagnostic system DISKET using knowledge engineering technique

    International Nuclear Information System (INIS)

    Yokobayashi, Masao; Yoshida, Kazuo; Kohsaka, Atsuo; Yamamoto, Minoru.

    1986-01-01

    An accident diagnostic system DISKET has been developed to identify the cause and the type of an abnormal transient of a nuclear power plant. The system is based on the knowledge engineering (KE) and consists of an inference engine IERIAS and a knowledge base. The main features of DISKET are the following : (1) Time-varying characteristics of transients can be treated. (2) Knowledge base can be divided into several knowledge units to handle a lot of rules effectively. (3) Programming language UTILISP, which is a dialect of LISP, is used to manipulate symbolic data effectively. For the verification of DISKET, performance tests have been conducted for several types of accidents. The knowledge base used in the tests was generated from the data of various types of transients produced by a PWR plant simulator. The results of verification studies showed a good applicability of DISKET to reactor accident diagnosis. (author)

  5. 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

  6. Reactor coolant pump monitoring and diagnostic system

    International Nuclear Information System (INIS)

    Singer, R.M.; Gross, K.C.; Walsh, M.; Humenik, K.E.

    1990-01-01

    In order to reliably and safely operate a nuclear power plant, it is necessary to continuously monitor the performance of numerous subsystems to confirm that the plant state is within its prescribed limits. An important function of a properly designed monitoring system is the detection of incipient faults in all subsystems (with the avoidance of false alarms) coupled with an information system that provides the operators with fault diagnosis, prognosis of fault progression and recommended (either automatic or prescriptive) corrective action. In this paper, such a system is described that has been applied to reactor coolant pumps. This system includes a sensitive pattern-recognition technique based upon the sequential probability ratio test (SPRT) that detects incipient faults from validated signals, an expert system embodying knowledge bases on pump and sensor performance, extensive hypertext files containing operating and emergency procedures as well as pump and sensor information and a graphical interface providing the operator with easily perceived information on the location and character of the fault as well as recommended corrective action. This system is in the prototype stage and is currently being validated utilizing data from a liquid-metal cooled fast reactor (EBR-II). 3 refs., 4 figs

  7. Problems in calculating reactor model (primary circuit) for nuclear power plant diagnostics

    International Nuclear Information System (INIS)

    Markov, P.

    1986-01-01

    Some results are presented of the calculation of eigen-vibrations of the system of WWER-440 nuclear reactor vessels in a vacuum and in a liquid. Computer code BOSOR 4 has been written for calculating forced vibrations of shells with axial symmetry and of a simplified system of reactor vessels. A description is given of this code, which is based on the so-called energy method of finite differences. Briefly discussed is the feasibility of applying the results of the latest computation techniques in the diagnostics of the major components of a nuclear reactor. (Z.M.)

  8. Implementation of an integrated on-line process surveillance and diagnostic system at the Halden reactor project: MOAS

    International Nuclear Information System (INIS)

    Kim, I.S.; Grini, R.-E.; Nilsen, S.

    2001-01-01

    MOAS is an integrated on-line process surveillance and diagnostic system that uses several different models for knowledge acquisition and diagnostic reasoning, such as goal-tree success-tree model, process monitor trees, and sensor failure diagnosis trees. Within these models, the knowledge of the process and its operation, including deep knowledge, like mass balance or controller algorithm, is incorporated. During an extensive review, made as part of the integrated diagnosis system project of the Halden reactor project, MOAS (Maryland Operator Advisory System) was identified as one of the most thorough systems developed thus far. MOAS encompasses diverse functional aspects that are required for an effective process disturbance management: (1) intelligent process monitoring and alarming, (2) on-line sensor data validation and sensor failure diagnosis, (3) on-line hardware (besides sensors) failure diagnosis, and (4) real-time corrective measure synthesis. The MOAS methodology was used for the NORS (Nokia Research Simulator) process at the Halden man-machine laboratory HAMMLAB of the OECD Halden reactor project. The performance tests of MOAS, implemented in G2 real-time expert system shell, show that MOAS successfully carries out its intended functions, i.e. quickly recognizing an occurring disturbance, correctly diagnosing its cause, and presenting advice on its control to the operator. The lessons learned and insights gained during the implementation and performance tests also are discussed

  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. Nuclear reactor pump diagnostics via noise analysis/artificial neural networks

    International Nuclear Information System (INIS)

    Keyvan, S.; Rabelo, L.C.

    1991-01-01

    A feasibility study is performed on the utilization of artificial neural networks as a tool for reactor diagnostics. Reactor pump signals utilized in a wear-out monitoring system developed for early detection of degradation of pump shaft are analyzed as a semi-benchmark test to study the feasibility of neural networks for pattern recognition. The Adaptive Resonance Theory (ART 2) paradigm of artificial neural networks is applied in this study. The signals are collected signals as well as generated signals simulating the wear progress. The wear-out monitoring system applies noise analysis techniques, and is capable of distinguishing between these signals and providing a measure of the progress of the degradation. This paper presents the results of the analysis of these data via the ART 2 paradigm

  11. Dynamics model for real time diagnostics of Triga RC-1 system

    International Nuclear Information System (INIS)

    Gadomski, A.M.; Nanni, V.; Meo, G.

    1988-01-01

    This paper presents dynamics model of TRIGA RC-1 reactor system. The model is dedicated to the real-time early fault detection during a reactor operation in one week exploitation cycle. The algorithms are specially suited for real-time, long time and also accelerated simulations with assumed diagnostic oriented accuracy. The approximations, modular structure, numerical methods and validation are discussed. The elaborated model will be build in the TRIGA Supervisor System and TRIGA Diagnostic Simulator

  12. A study on the fault diagnostic techniques for reactor internal structures using neutron noise analysis

    International Nuclear Information System (INIS)

    Kim, Tae Ryong; Jeong, Seong Ho; Park, Jin Ho; Park, Jin Suk

    1994-08-01

    The unfavorable phenomena, such as flow induced vibration and aging process in reactor internals, cause degradation of structural integrity and may result in loosing some mechanical binding components which might impact other equipments and components or cause flow blockage. Since these malfunctions and potential failures change reactor noise signal, it is necessary to analyze reactor noise signal for early fault diagnosis in the point of few of safety and plant economics. The objectives of this study are to establish fault diagnostic and TS(thermal shield), and to develop a data acquisition and signal processing software system. In the first year of this study, an analysis technique for the reactor internal vibration using the reactor noise was proposed. With the technique proposed and the reactor noise signals (ex-core neutron and acceleration), the dynamic characteristics of Ulchin-1 reactor internals were obtained, and compared with those of Tricastin-1 which is the prototype of Ulchin-1. In the second year, a PC-based expert system for reactor internals fault diagnosis is developed, which included data acquisition, signal processing, feature extraction function, and represented diagnostic knowledge by the IF-THEN rule. To know the effect of the faults, the reactor internals of Ulchin-1 is modeled using FEM and simulated with an artificial defect given in the hold-down spring. Trend in the dynamic characteristics of reactor internals is also observed during one fuel cycle to know the effect of boron concentration. 100 figs, 7 tabs, 18 refs. (Author)

  13. On-line diagnostics for a real time system

    International Nuclear Information System (INIS)

    Sreenivasan, P.

    1976-01-01

    The purpose of an on-line diagnostics is to infuse the ability of self diagnosing in an online computer to enhance its dependability in a real time system. Such a diagnostics evolved for the CDPS of the Fast Breeder Test Reactor at Kalpakkam is reported. The two phases of the diagnostics, i.e., the malfunction detection and post detection action are described in some detail. (A.K.)

  14. Dynamics model for real time diagnostics of TRIGA RC-1 system

    International Nuclear Information System (INIS)

    Gadomski, A.M.; Nanni, V.; Meo, G.B.

    1986-01-01

    This paper presents dynamics model of TRIGA RC-1 reactor system. The model is dedicated to the real-time early fault detection during a reactor operation in one week exploitation cycle. The algorithms are specially suited for real-time, long time and also accelerated simulations with assumed diagnostic oriented accuracy. The approximations, modular structure, numerical methods and validation are discussed. The elaborated model will be build in the TRIGA Supervisory System and TRIGA Diagnostic Simulator. (author)

  15. Nuclear power plant diagnostic system

    International Nuclear Information System (INIS)

    Prokop, K.; Volavy, J.

    1982-01-01

    Basic information is presented on diagnostic systems used at nuclear power plants with PWR reactors. They include systems used at the Novovoronezh nuclear power plant in the USSR, at the Nord power plant in the GDR, the system developed at the Hungarian VEIKI institute, the system used at the V-1 nuclear power plant at Jaslovske Bohunice in Czechoslovakia and systems of the Rockwell International company used in US nuclear power plants. These diagnostic systems are basically founded on monitoring vibrations and noise, loose parts, pressure pulsations, neutron noise, coolant leaks and acoustic emissions. The Rockwell International system represents a complex unit whose advantage is the on-line evaluation of signals which gives certain instructions for the given situation directly to the operator. The other described systems process signals using similar methods. Digitized signals only serve off-line computer analyses. (Z.M.)

  16. Report on the diagnostics for control of the fusion DEMO reactors

    International Nuclear Information System (INIS)

    2014-05-01

    The range of diagnostics that can be used in DEMO will be severely restricted compared to that used in the current experiments or to be used in ITER. Therefore, a study is planned on the technical feasibility of sensors and diagnostics on the basis of specific tokamak and helical DEMO designs, with the involvement of a wide range of specialists covering reactor design, diagnostics, neutronics, reactor structure, remote maintenance, plasma physics, plasma and machine control, and computer simulation. Topics included typical characteristic times of target plasma behavior, diagnostics tools with their resolution and lifetime, response time of actuators, and plasmas. Through these studies, possible candidates for DEMO diagnostics were identified. The outcome of two years of activities is summarized in this report with a recommendation to the government of Japan. (J.P.N.)

  17. Development of a reactor-coolant-pump monitoring and diagnostic system. Semi-annual progress report, December 1981-May 1982

    International Nuclear Information System (INIS)

    Morris, D.J.; Gabler, H.C.

    1982-10-01

    Reactor coolant (RC) pump seal failures have resulted in excessive leakage of primary coolant into reactor containment buildings. In some cases, high levels of airborne activity and surface contamination following these failures have necessitated extensive cleanup efforts and personnel radiation exposure. Unpredictable pump seal performance has also caused forced outages and frequent maintenance. The quality of operating data has been insufficient to allow proper evaluation of theoretical RC pump seal failure mechanisms. The RC pump monitoring and diagnostic system being developed and installed at Toledo Edison's Davis-Besse Nuclear Power Station will examine the relationship between seal failures and three other variables. This report describes system software and hardware development, testing, and installation work performed during the period of December 1981 through May 1982. Also described herein is a parallel effort being conducted by a B and W/Byron Jackson/Utility group to improve pump seal performance

  18. Systems for noise diagnostics of WWER nuclear power plants

    International Nuclear Information System (INIS)

    Por, G.

    1996-01-01

    The aim of this paper is to give a short overview of the noise diagnostics system developed by Hungarian firms which are in operation in WWER type NPP Units. Giving a list of systems developed for noise diagnostics of WWER reactors we present their main characteristics, their goal and some of their achievements. The second part deals with the problem of acceptance of noise system by NPP and regulations. (author). 24 refs

  19. Systems for noise diagnostics of WWER nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Por, G [Technical Univ. of Budapest, Budapest (Hungary)

    1997-12-31

    The aim of this paper is to give a short overview of the noise diagnostics system developed by Hungarian firms which are in operation in WWER type NPP Units. Giving a list of systems developed for noise diagnostics of WWER reactors we present their main characteristics, their goal and some of their achievements. The second part deals with the problem of acceptance of noise system by NPP and regulations. (author). 24 refs.

  20. Preliminary consideration of CFETR ITER-like case diagnostic system.

    Science.gov (United States)

    Li, G S; Yang, Y; Wang, Y M; Ming, T F; Han, X; Liu, S C; Wang, E H; Liu, Y K; Yang, W J; Li, G Q; Hu, Q S; Gao, X

    2016-11-01

    Chinese Fusion Engineering Test Reactor (CFETR) is a new superconducting tokamak device being designed in China, which aims at bridging the gap between ITER and DEMO, where DEMO is a tokamak demonstration fusion reactor. Two diagnostic cases, ITER-like case and towards DEMO case, have been considered for CFETR early and later operating phases, respectively. In this paper, some preliminary consideration of ITER-like case will be presented. Based on ITER diagnostic system, three versions of increased complexity and coverage of the ITER-like case diagnostic system have been developed with different goals and functions. Version A aims only machine protection and basic control. Both of version B and version C are mainly for machine protection, basic and advanced control, but version C has an increased level of redundancy necessary for improved measurements capability. The performance of these versions and needed R&D work are outlined.

  1. Preliminary consideration of CFETR ITER-like case diagnostic system

    International Nuclear Information System (INIS)

    Li, G. S.; Liu, Y. K.; Gao, X.; Yang, Y.; Wang, Y. M.; Ming, T. F.; Han, X.; Liu, S. C.; Wang, E. H.; Yang, W. J.; Li, G. Q.; Hu, Q. S.

    2016-01-01

    Chinese Fusion Engineering Test Reactor (CFETR) is a new superconducting tokamak device being designed in China, which aims at bridging the gap between ITER and DEMO, where DEMO is a tokamak demonstration fusion reactor. Two diagnostic cases, ITER-like case and towards DEMO case, have been considered for CFETR early and later operating phases, respectively. In this paper, some preliminary consideration of ITER-like case will be presented. Based on ITER diagnostic system, three versions of increased complexity and coverage of the ITER-like case diagnostic system have been developed with different goals and functions. Version A aims only machine protection and basic control. Both of version B and version C are mainly for machine protection, basic and advanced control, but version C has an increased level of redundancy necessary for improved measurements capability. The performance of these versions and needed R&D work are outlined.

  2. Development and application of reactor noise diagnostics

    Energy Technology Data Exchange (ETDEWEB)

    Karlsson, Joakim K.H

    1999-04-01

    A number of problems in reactor noise diagnostics have been investigated within the framework of the present thesis. The six papers presented cover three relatively different areas, namely the use of analytical calculations of the neutron noise in simple reactor models, some aspects of boiling water reactor (BWR) stability and diagnostics of core barrel motion in pressurized water reactors (PWRs). The noise induced by small vibrations of a strong absorber has been the subject of several previous investigations. For a conventional {delta}-function source model, the equations can not be linearized in the traditional manner. Thus, a new source model, which is called the {epsilon}/d model, was developed. The correct solution has been derived in the {epsilon}/d model for both 1-D and 2-D reactor models. Recently, several reactor diagnostic problems have occurred which include a control rod partially inserted into the reactor core. In order to study such problems, we have developed an analytically solvable, axially non-homogenous, 2-D reactor model. This model has also been used to study the noise induced by a rod maneuvering experiment. Comparisons of the noise with the results of different reactor kinetic approximations have yielded information on the validity of the approximations in this relatively realistic model. In case of an instability event in a BWR, the noise may consist of one or several co-existing modes of oscillation and besides the fundamental mode, a regional first azimuthal mode has been observed in e.g. the Swedish BWR Ringhals-1. In order to determine the different stability characteristics of the different modes separately, it is important to be able to decompose the noise into its mode constituents. A separation method based on factorisation of the flux has been attempted previously, but without success. The reason for the failure of the factorisation method is the presence of the local component of the noise and its axial correlation properties. In

  3. Development and application of reactor noise diagnostics

    International Nuclear Information System (INIS)

    Karlsson, Joakim K.H.

    1999-04-01

    A number of problems in reactor noise diagnostics have been investigated within the framework of the present thesis. The six papers presented cover three relatively different areas, namely the use of analytical calculations of the neutron noise in simple reactor models, some aspects of boiling water reactor (BWR) stability and diagnostics of core barrel motion in pressurized water reactors (PWRs). The noise induced by small vibrations of a strong absorber has been the subject of several previous investigations. For a conventional δ-function source model, the equations can not be linearized in the traditional manner. Thus, a new source model, which is called the ε/d model, was developed. The correct solution has been derived in the ε/d model for both 1-D and 2-D reactor models. Recently, several reactor diagnostic problems have occurred which include a control rod partially inserted into the reactor core. In order to study such problems, we have developed an analytically solvable, axially non-homogenous, 2-D reactor model. This model has also been used to study the noise induced by a rod maneuvering experiment. Comparisons of the noise with the results of different reactor kinetic approximations have yielded information on the validity of the approximations in this relatively realistic model. In case of an instability event in a BWR, the noise may consist of one or several co-existing modes of oscillation and besides the fundamental mode, a regional first azimuthal mode has been observed in e.g. the Swedish BWR Ringhals-1. In order to determine the different stability characteristics of the different modes separately, it is important to be able to decompose the noise into its mode constituents. A separation method based on factorisation of the flux has been attempted previously, but without success. The reason for the failure of the factorisation method is the presence of the local component of the noise and its axial correlation properties. In the paper

  4. Diagnostic system and diagnostic experiences at the Paks Nuclear Power Plant

    International Nuclear Information System (INIS)

    Katona, Tamas

    1986-01-01

    The major functions of the diagnostic system of the first two units of the Paks Nuclear Power Plant are as follows: monitoring the mechanical integrity of the reactor and the primary coolant circuit by means of vibration diagnostics; leakage detection of the primary coolant circuit by means of high frequency sonic analysis; loose parts monitoring based on the analysis of high frequency signals of acceleration detectors; and monitoring the vibration state of the turbines and rotary machines by the latter method or by a procedure based on the detection of mechanical vibrations. Up-to-date vibration diagnostics is based on the information supplied by either acceleration detectors or pressure fluctuation detectors, or in-core and ex-core neutron detectors. (V.N.)

  5. Plasma Diagnostics in High Density Reactors

    International Nuclear Information System (INIS)

    Daltrini, A. M.; Moshkalyov, S.; Monteiro, M. J. R.; Machida, M.; Kostryukov, A.; Besseler, E.; Biasotto, C.; Diniz, J. A.

    2006-01-01

    Langmuir electric probes and optical emission spectroscopy diagnostics were developed for applications in high density plasmas. These diagnostics were employed in two plasma sources: an electron cyclotron resonance (ECR) plasma and an RF driven inductively coupled plasma (ICP) plasma. Langmuir probes were tested using a number of probing dimensions, probe tip materials, circuits for probe bias and filters. Then, the results were compared with the optical spectroscopy measurements. With these diagnostics, analyses of various plasma processes were performed in both reactors. For example, it has been shown that species like NH radicals generated in gas phase can have critical impact on films deposited by ECR plasmas. In the ICP source, plasmas in atomic and molecular gases were shown to have different spatial distributions, likely due to nonlocal electron heating. The low-to-high density transitions in the ICP plasma were also studied. The role of metastables is shown to be significant in Ar plasmas, in contrast to plasmas with additions of molecular gases

  6. IDAS-RR: an incident data base system for research reactors

    International Nuclear Information System (INIS)

    Matsumoto, Kiyoshi; Kohsaka, Atsuo; Kaminaga, Masanori; Murayama, Youji; Ohnishi, Nobuaki; Maniwa, Masaki.

    1990-03-01

    An Incident Data Base System for Research Reactors, IDAS-RR, has been developed. IDAS-RR has information about abnormal incidents (failures, transients, accidents, etc.) of research reactors in the world. Data reference, input, editing and other functions of IDAS-RR are menu driven. The routine processing and data base management functions are performed by the system software and hardware. PC-9801 equipment was selected as the hardware because of its portability and popularity. IDAS-RR provides effective reference information for the following activities. 1) Analysis of abnormal incident of research reactors, 2) Detail analysis of research reactor behavior in the abnormal incident for building the knowledge base of the reactor emergency diagnostic system for research reactor, 3) Planning counter-measure for emergency situation in the research reactor. This report is a user's manual of IDAS-RR. (author)

  7. 'Kazmer' a complex noise diagnostic system for 1000 MWe PWR WWER type nuclear power units

    International Nuclear Information System (INIS)

    Por, G.

    1992-06-01

    Noise diagnostic systems have previously been developed and installed for the WWER-440 type reactors at the Paks Nuclear Power Plant, Hungary. Based on the experiences, the system has been extended and modified for use in 1000 MWe, WWER-1000 type units. KAZMER consists of three subsystem, the KARD reactor noise diagnostic system, ARGUS vibration monitoring system for rotation machinery, and ALMOS acoustic monitoring system. The installation of the KAZMER system at the Kalinin Nuclear Power Station, Russia, and the first operational experiences are outlined. (R.P.) 15 refs.; 9 figs

  8. LOFT advanced control room operator diagnostic and display system (ODDS)

    International Nuclear Information System (INIS)

    Larsen, D.G.; Robb, T.C.

    1980-01-01

    The Loss-of-Fluid Test (LOFT) Reactor Facility in Idaho includes a highly instrumented nuclear reactor operated by the Department of Energy for the purpose of establishing nuclear safety requirements. The results of the development and installation into LOFT of an Operator Diagnostic and Display System (ODDS) are presented. The ODDS is a computer-based graphics display system centered around a PRIME 550 computer with several RAMTEK color graphic display units located within the control room and available to the reactor operators. Use of computer-based color graphics to aid the reactor operator is discussed. A detailed hardware description of the LOFT data system and the ODDS is presented. Methods and problems of backfitting the ODDS equipment into the LOFT plant are discussed

  9. Study for the design method of multi-agent diagnostic system to improve diagnostic performance for similar abnormality

    International Nuclear Information System (INIS)

    Minowa, Hirotsugu; Gofuku, Akio

    2014-01-01

    Accidents on industrial plants cause large loss on human, economic, social credibility. In recent, studies of diagnostic methods using techniques of machine learning such as support vector machine is expected to detect the occurrence of abnormality in a plant early and correctly. There were reported that these diagnostic machines has high accuracy to diagnose the operating state of industrial plant under mono abnormality occurrence. But the each diagnostic machine on the multi-agent diagnostic system may misdiagnose similar abnormalities as a same abnormality if abnormalities to diagnose increases. That causes that a single diagnostic machine may show higher diagnostic performance than one of multi-agent diagnostic system because decision-making considering with misdiagnosis is difficult. Therefore, we study the design method for multi-agent diagnostic system to diagnose similar abnormality correctly. This method aimed to realize automatic generation of diagnostic system where the generation process and location of diagnostic machines are optimized to diagnose correctly the similar abnormalities which are evaluated from the similarity of process signals by statistical method. This paper explains our design method and reports the result evaluated our method applied to the process data of the fast-breeder reactor Monju

  10. Data acquisition, monitoring and diagnostic system for predictive control and protection of rotating components of IEAR-1 reactor by vibration analysis

    International Nuclear Information System (INIS)

    Serra, Reynaldo Cavalcanti; Tecco, Dorival Goncalves

    1996-01-01

    This work presents the vibration and temperature data acquisition, monitoring and diagnostic systems, recently installed in the primary circuit, secondary circuit and emergency generator of the IEA-R1 reactor at IPEN during the course of the first power elevation tests to 5MW. It incorporates a series of routines for equipment configuration, interactive automatic monitoring , data processing and documentation/storage without the exposure of operators in the radiological protection areas. (author)

  11. Diagnostic mirror concept development for use in the complex environment of a fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Krimmer, Andreas Joachim

    2016-07-01

    Light-based diagnostic systems of fusion reactors require optical mirrors to channel light through the structures surrounding the plasma. With increasing plasma volume, power and plasma burn time, the environmental conditions grow more demanding and new requirements arise. In this dissertation, the design of optical mirrors inside the vacuum chamber of the prototype reactor ITER (Latin ''the way'') and future fusion power plants are investigated. Comparing the state of the art with the boundary conditions close to the fusion plasma, existing mirror designs and choices for the reflective surface are evaluated. For the design, it is not the individual boundary conditions that are critical, but rather, their combination and the resulting interactions. Drawing from the existing designs, possible realizations for central functionality are discussed. Included in the discussion are substrate choice, mounting, adjustment and thermal contacting as well as positioning of the mirror assembly compatible with hot cell maintenance. Building on the general discussion, mirror concepts for the charge exchange recombination spectroscopy (CXRS) diagnostic system for the ITER plasma core are proposed and simulated. In addition, prototypes are manufactured and tested to assess critical aspects of the proposed design. Testing includes positioning by pins, manufacturing of a stainless steel substrate with fluid channels adapted to the mirror shape, and tests with an SiO{sub 2} /TiO{sub 2} dielectric coating under selected ITER conditions. As a result of the work, the fusion reactor mirror design considerations given in the principal design discussion can be used as a basis for other diagnostic systems as well. In the case of the core CXRS mirror concept for ITER, the basic suitability was shown and critical topics were identified where additional work is necessary.

  12. Diagnostic mirror concept development for use in the complex environment of a fusion reactor

    International Nuclear Information System (INIS)

    Krimmer, Andreas Joachim

    2016-01-01

    Light-based diagnostic systems of fusion reactors require optical mirrors to channel light through the structures surrounding the plasma. With increasing plasma volume, power and plasma burn time, the environmental conditions grow more demanding and new requirements arise. In this dissertation, the design of optical mirrors inside the vacuum chamber of the prototype reactor ITER (Latin ''the way'') and future fusion power plants are investigated. Comparing the state of the art with the boundary conditions close to the fusion plasma, existing mirror designs and choices for the reflective surface are evaluated. For the design, it is not the individual boundary conditions that are critical, but rather, their combination and the resulting interactions. Drawing from the existing designs, possible realizations for central functionality are discussed. Included in the discussion are substrate choice, mounting, adjustment and thermal contacting as well as positioning of the mirror assembly compatible with hot cell maintenance. Building on the general discussion, mirror concepts for the charge exchange recombination spectroscopy (CXRS) diagnostic system for the ITER plasma core are proposed and simulated. In addition, prototypes are manufactured and tested to assess critical aspects of the proposed design. Testing includes positioning by pins, manufacturing of a stainless steel substrate with fluid channels adapted to the mirror shape, and tests with an SiO_2 /TiO_2 dielectric coating under selected ITER conditions. As a result of the work, the fusion reactor mirror design considerations given in the principal design discussion can be used as a basis for other diagnostic systems as well. In the case of the core CXRS mirror concept for ITER, the basic suitability was shown and critical topics were identified where additional work is necessary.

  13. Optical fibres for fusion plasma diagnostics systems

    International Nuclear Information System (INIS)

    Brichard, B.

    2005-01-01

    The condition to achieve and maintain the ignition of a thermonuclear fusion plasma ignition calls for the construction of a large scale fusion reactor, namely ITER. This reactor is designed to deliver an average fusion power of 500 MW. The burning of fusion plasma at such high power level will release a tremendous amount of energy in the form of particle fluxes and ionising radiation. This energy release, primarily absorbed by the plasma facing components, can significantly degrade the performances of the plasma diagnostic equipment surrounding the machine. To ensure a correct operation of the Tokamak we need to develop highly radiation-resistance devices. In plasma diagnostic systems, optical fibre is viewed as a convenient tool to transport light from the plasma edge to the diagnostic area. Radiation affects the optical performances of the fibre mainly by the occurrence of radiation-induced absorption and luminescence. Both effects degrade the light signal used for plasma diagnostic. SCK-CEN is currently assessing radiation-resistant glasses for optical fibres and is developing the associated qualification procedure. The main objectives of this study were to increase the lifetime of optical components in high radiation background and to develop a radiation resistance optical fibre capable to operate in the radiation background of ITER

  14. The role of neural networks in reactor diagnostics and control

    International Nuclear Information System (INIS)

    Pazsit, I.; Kitamura, M.

    1997-01-01

    Reactor diagnostics and core diagnostics in particular, is an inverse task, just as most other diagnostics. One measures some physical parameter at some position, or the fluctuation thereof, which is given rise by the fluctuation of another parameter, presumably at a different position. In neutron noise diagnostics, the measured quantity is the neutron noise, whereas the cause, fluctuations of the core material, is called the open-quotes noise sourceclose quotes. The relationship between the cause (noise source) and the induced noise (effect) is determined by the physics of the process and can usually be described by a theory. This means that the direct task, calculation of the noise from the noise source, can always be achieved. In diagnostics, however, the process starts from the back end, i.e. one observes the effect of some cause. The task is to infer the cause (noise source) from the effect (induced noise), which is an inverse task (sometimes also called unfolding). The situation is very much similar even regarding control problems. The system state is described by a vector in the parameter space. When it deviates from the desired one, the control parameters need to be changed to bring the system into the desired state. The computation of the change in the state vector, due to a change in the control vector, can be calculated with no difficulty in principle. The control requires however the inverse task to be performed, i.e. to determine which changes of the control vector would bring the state vector into the desired state

  15. Development of a reactor coolant pump monitoring and diagnostic system. Progress report, June 1982-July 1983

    International Nuclear Information System (INIS)

    Morris, D.J.; Sommerfield, G.A.

    1983-12-01

    The quality of operating data has been insufficient to allow proper evaluation of theoretical reactor coolant (RC) pump seal failure mechanisms. The RC pump monitoring and diagnostic system being developed and installed at Toledo Edison's Davis-Besse Nuclear Power Station will examine the relationship between seal failures and three other variables: The rotordynamic behavior of the pump shaft and related components, the internal conditions and performance of the seals, and the plant or pump operating environment (controlled by the plant operator). Interrelationships between these areas will be developed during the data collection task, scheduled to begin in October 1983 (for a full fuel cycle at Davis-Besse). This report describes system software and hardware development, testing, and installation work performed during this period. Also described is a parallel effort being conducted by a B and W/Byron Jackson/Utility group to improve pump seal performance

  16. 9 CFR 75.4 - Interstate movement of equine infectious anemia reactors and approval of laboratories, diagnostic...

    Science.gov (United States)

    2010-01-01

    ... infectious anemia reactors and approval of laboratories, diagnostic facilities, and research facilities. 75.4... IN HORSES, ASSES, PONIES, MULES, AND ZEBRAS Equine Infectious Anemia (swamp Fever) § 75.4 Interstate movement of equine infectious anemia reactors and approval of laboratories, diagnostic facilities, and...

  17. A proof-of-concept transient diagnostic expert system for BWRs [Boiling Water Reactors

    International Nuclear Information System (INIS)

    Yoshida, K.; Naser, J.A.

    1988-05-01

    A proof-of-concept transient diagnostic expert system has been developed to identify the cause and the type of an abnormal transient in a boiling water nuclear power plant. For this expert system development, the calculational results of the simulation code RETRAN were used as the knowledge source. The knowledge extracted from the RETRAN analyses was transformed into IF-THEN rules in the knowledge base for the expert system. An important feature of this expert system is the introduction of certainty factors to allow diagnosis even in the cases where data may be either missing or marked as invalid. To increase the capability of this diagnostic system to distinguish between similiar transients, backward chaining reasoning is used to support the forward chaining reasoning with certainty factors. Through this effort, it has been demonstrated that an expert system can be successfully used to create a transient diagnostic system. It has also successfully demonstrated that RETRAN can be used as the knowledge source for developing the knowledge base of the diagnostic system

  18. Vibration monitoring/diagnostic techniques, as applied to reactor coolant pumps

    International Nuclear Information System (INIS)

    Sculthorpe, B.R.; Johnson, K.M.

    1986-01-01

    With the increased awareness of reactor coolant pump (RCP) cracked shafts, brought about by the catastrophic shaft failure at Crystal River number3, Florida Power and Light Company, in conjunction with Bently Nevada Corporation, undertook a test program at St. Lucie Nuclear Unit number2, to confirm the integrity of all four RCP pump shafts. Reactor coolant pumps play a major roll in the operation of nuclear-powered generation facilities. The time required to disassemble and physically inspect a single RCP shaft would be lengthy, monetarily costly to the utility and its customers, and cause possible unnecessary man-rem exposure to plant personnel. When properly applied, vibration instrumentation can increase unit availability/reliability, as well as provide enhanced diagnostic capability. This paper reviews monitoring benefits and diagnostic techniques applicable to RCPs/motor drives

  19. Research at ITER towards DEMO: Specific reactor diagnostic studies to be carried out on ITER

    Energy Technology Data Exchange (ETDEWEB)

    Krasilnikov, A. V.; Kaschuck, Y. A.; Vershkov, V. A.; Petrov, A. A.; Petrov, V. G.; Tugarinov, S. N. [Institution Project center ITER, Moscow (Russian Federation)

    2014-08-21

    In ITER diagnostics will operate in the very hard radiation environment of fusion reactor. Extensive technology studies are carried out during development of the ITER diagnostics and procedures of their calibration and remote handling. Results of these studies and practical application of the developed diagnostics on ITER will provide the direct input to DEMO diagnostic development. The list of DEMO measurement requirements and diagnostics will be determined during ITER experiments on the bases of ITER plasma physics results and success of particular diagnostic application in reactor-like ITER plasma. Majority of ITER diagnostic already passed the conceptual design phase and represent the state of the art in fusion plasma diagnostic development. The number of related to DEMO results of ITER diagnostic studies such as design and prototype manufacture of: neutron and γ–ray diagnostics, neutral particle analyzers, optical spectroscopy including first mirror protection and cleaning technics, reflectometry, refractometry, tritium retention measurements etc. are discussed.

  20. Research at ITER towards DEMO: Specific reactor diagnostic studies to be carried out on ITER

    Science.gov (United States)

    Krasilnikov, A. V.; Kaschuck, Y. A.; Vershkov, V. A.; Petrov, A. A.; Petrov, V. G.; Tugarinov, S. N.

    2014-08-01

    In ITER diagnostics will operate in the very hard radiation environment of fusion reactor. Extensive technology studies are carried out during development of the ITER diagnostics and procedures of their calibration and remote handling. Results of these studies and practical application of the developed diagnostics on ITER will provide the direct input to DEMO diagnostic development. The list of DEMO measurement requirements and diagnostics will be determined during ITER experiments on the bases of ITER plasma physics results and success of particular diagnostic application in reactor-like ITER plasma. Majority of ITER diagnostic already passed the conceptual design phase and represent the state of the art in fusion plasma diagnostic development. The number of related to DEMO results of ITER diagnostic studies such as design and prototype manufacture of: neutron and γ-ray diagnostics, neutral particle analyzers, optical spectroscopy including first mirror protection and cleaning technics, reflectometry, refractometry, tritium retention measurements etc. are discussed.

  1. Research at ITER towards DEMO: Specific reactor diagnostic studies to be carried out on ITER

    International Nuclear Information System (INIS)

    Krasilnikov, A. V.; Kaschuck, Y. A.; Vershkov, V. A.; Petrov, A. A.; Petrov, V. G.; Tugarinov, S. N.

    2014-01-01

    In ITER diagnostics will operate in the very hard radiation environment of fusion reactor. Extensive technology studies are carried out during development of the ITER diagnostics and procedures of their calibration and remote handling. Results of these studies and practical application of the developed diagnostics on ITER will provide the direct input to DEMO diagnostic development. The list of DEMO measurement requirements and diagnostics will be determined during ITER experiments on the bases of ITER plasma physics results and success of particular diagnostic application in reactor-like ITER plasma. Majority of ITER diagnostic already passed the conceptual design phase and represent the state of the art in fusion plasma diagnostic development. The number of related to DEMO results of ITER diagnostic studies such as design and prototype manufacture of: neutron and γ–ray diagnostics, neutral particle analyzers, optical spectroscopy including first mirror protection and cleaning technics, reflectometry, refractometry, tritium retention measurements etc. are discussed

  2. Investigations of certain problems in reactor noise diagnostics

    International Nuclear Information System (INIS)

    Karlsson, Joakim.

    1997-06-01

    A number of problems in reactor noise diagnostics have been investigated within the framework of the present thesis. The papers presented cover three relatively different areas, namely neutron noise induced by small vibrations of a strong absorber, the decomposition of noise in BWRs with application to stability monitoring, and fluctuation analysis of soft x-ray signals from fusion plasmas. 18 refs, 6 figs

  3. Investigations of certain problems in reactor noise diagnostics

    Energy Technology Data Exchange (ETDEWEB)

    Karlsson, Joakim

    1997-06-01

    A number of problems in reactor noise diagnostics have been investigated within the framework of the present thesis. The papers presented cover three relatively different areas, namely neutron noise induced by small vibrations of a strong absorber, the decomposition of noise in BWRs with application to stability monitoring, and fluctuation analysis of soft x-ray signals from fusion plasmas. 18 refs, 6 figs.

  4. Incipient Transient Detection in Reactor Systems: Experimental and Theoretical Investigation

    Energy Technology Data Exchange (ETDEWEB)

    Lefteri H. Tsoukalas; S.T. Revankar; X Wang; R. Sattuluri

    2005-09-27

    The main goal of this research was to develop a method for detecting reactor system transients at the earliest possible time through a comprehensive experimental, testing and benchmarking program. This approach holds strong promise for developing new diagnostic technologies that are non-intrusive, generic and highly portable across different systems. It will help in the design of new generation nuclear power reactors, which utilize passive safety systems with a reliable and non-intrusive multiphase flow diagnostic system to monitor the function of the passive safety systems. The main objective of this research was to develop an improved fuzzy logic based detection method based on a comprehensive experimental testing program to detect reactor transients at the earliest possible time, practically at their birth moment. A fuzzy logic and neural network based transient identification methodology and implemented in a computer code called PROTREN was considered in this research and was compared with SPRT (Sequentially Probability Ratio Testing) decision and Bayesian inference. The project involved experiment, theoretical modeling and a thermal-hydraulic code assessment. It involved graduate and undergraduate students participation providing them with exposure and training in advanced reactor concepts and safety systems. In this final report, main tasks performed during the project period are summarized and the selected results are presented. Detailed descriptions for the tasks and the results are presented in previous yearly reports (Revankar et al 2003 and Revankar et al 2004).

  5. Integrated control and diagnostic system architectures for future installations

    International Nuclear Information System (INIS)

    Wood, R.; March-Leuba, J.

    2000-01-01

    Nuclear reactors of the 21st century will employ increasing levels of automation and fault tolerance to increase availability, reduce accident risk, and lower operating costs. Key developments in control algorithms, fault diagnostics, fault tolerance, and distributed communications are needed to implement the fully automated plant. It will be equally challenging to integrate developments in separate information and control fields into a cohesive system, which collectively achieves the overall goals of improved safety, reliability, maintainability, and cost-effectiveness. Under the Nuclear Energy Research Initiative (NERI), the US Department of Energy is sponsoring a project to address some of the technical issues involved in meeting the long-range goal of 21st century reactor control systems. This project involves researchers from Oak Ridge National Laboratory, the University of Tennessee, and North Carolina State University. The research tasks under this project focus on some of the first-level breakthroughs in control design, diagnostic techniques, and information system design that will provide a path to enable the design process to be automated in the future. This paper describes the conceptual development of an integrated nuclear plant control and information system architecture, which incorporates automated control system development that can be traced to a set of technical requirements. The expectation is that an integrated plant architecture with optimal control and efficient use of diagnostic information can reduce the potential for operational errors and minimize challenges to the plant safety systems

  6. General review of diagnostic systems in EDF PWR units

    International Nuclear Information System (INIS)

    Chevalier, R.; Brasseur, S.; Ricard, B.

    1998-01-01

    Since the beginning of the French nuclear program, Electricite de France (EDF) has looked for ways to improve the availability and safety of its nuclear units. Therefore, monitoring systems on turbogenerators, reactor coolant pumps, primary circuits and core internal structures were designed by the Research and Development Division and implemented with technologies available during the 1970's. However, mainly because of difficulties for data interpretation by plant personnel, EDF subsequently decided to design and develop different tools to help plant operators to process a diagnosis: - a new generation of the Monitoring and Diagnostic System called PSAD, - expert systems for diagnosis on reactor coolant pumps (RCP) 'DIAPO' and turbogenerator units (TG) 'DIVA', - diagnostic guides written for most equipment pending the installation of new monitoring and diagnosis systems. The first version of PSAD, installed in five units, performs on-line monitoring of the turbogenerator shaft line, steam inlet valves, the reactor coolant pumps and the generator stator. The second version not yet implemented, will include Loose Part Detection (LPD) and Core Internal Structure Monitoring (CISM). The level of diagnosis achieved by PSAD depends on the component monitored. The TG and RCP monitoring functions of PSAD compute high level diagnosis descriptors such as natural frequencies and long term trends but do not elaborate a diagnosis automatically. However, a diagnostic assistance window is available on-line, whenever a warning message is displayed, whether for immediate or later action. The window presents a diagnostic approach whose purpose is to find the causes of the symptoms observed. This diagnosis approach is automated in the DIVA and DIAPO expert systems. Numerous potential faults can be identified for both systems thanks to a hierarchy of abnormal situations. The user interactively answers questions when information is needed to progress in the diagnosis. The resulting

  7. Prodiag--a hybrid artificial intelligence based reactor diagnostic system for process faults

    International Nuclear Information System (INIS)

    Reifman, J.; Wei, T.Y.C.; Vitela, J.E.; Applequist, C. A.; Chasensky, T.M.

    1996-01-01

    Commonwealth Research Corporation (CRC) and Argonne National Laboratory (ANL) are collaborating on a DOE-sponsored Cooperative Research and Development Agreement (CRADA), project to perform feasibility studies on a novel approach to Artificial Intelligence (Al) based diagnostics for component faults in nuclear power plants. Investigations are being performed in the construction of a first-principles physics-based plant level process diagnostic expert system (ES) and the identification of component-level fault patterns through operating component characteristics using artificial neural networks (ANNs). The purpose of the proof-of-concept project is to develop a computer-based system using this Al approach to assist process plant operators during off-normal plant conditions. The proposed computer-based system will use thermal hydraulic (T-H) signals complemented by other non-T-H signals available in the data stream to provide the process operator with the component which most likely caused the observed process disturbance.To demonstrate the scale-up feasibility of the proposed diagnostic system it is being developed for use with the Chemical Volume Control System (CVCS) of a nuclear power plant. A full-scope operator training simulator representing the Commonwealth Edison Braidwood nuclear power plant is being used both as the source of development data and as the means to evaluate the advantages of the proposed diagnostic system. This is an ongoing multi-year project and this paper presents the results to date of the CRADA phase

  8. Diagnostics in the hostile environments of a prototype fusion reactor

    International Nuclear Information System (INIS)

    Osher, J.E.

    1982-01-01

    Various facets of a thermonuclear type plasma that will likely require special considerations or hardening of applied diagnostic instrumentation are reviewed. The discussion will include both on-line diagnostic instrumentation requirements for satisfactory operation and considerations to reduce integrated radiation damage sufficiently for a reasonable diagnostic lifetime. Several new diagnostics aimed specifically at measurements of the plasma characteristics most appropriate to a thermonculear reactor type plasma are discussed. This will include instrumentation needed to make quantitative energy flow measurements during long term operation with the expected high input power sources, and locally very high wall power loadings. The second part of this lecture will broaden diagnostics to include materials damage measurements needed for engineering design studies. This will include needed diagnostic instrumentation to assess first wall damage, sputtering erosion at walls (and high power beam dumps), and radiation damage to components such as insulators

  9. Diagnostic measurement on research reactors

    International Nuclear Information System (INIS)

    Dach, K.; Zbytovsky, A.

    A comparison is made of noise experiments on zero power and power reactors. The general characteristics of noise experiments on power reactors is their ''passivity'', i.e., the experiment does not require any interruption of the normal operating regime of the reactor system. On zero power research reactors where the fission reaction constitutes the dominant noise source such conditions have to be created in the study of noise components as to make the investigated noise dominant and the noise of the fission reaction the background. The simultaneous use of both methods makes it possible to determine the spectral composition of reactivity fluctuations, which facilitates the identification of noise sources. The conditions are described of the recordability of noise components. The possibilities are listed provided for research work in Czechoslovakia and the possibility is studied of setting up an expert team to organize the respective experimental programme on an international scale. Power reactors manufactured in the GDR are considered as the suitable experimental base. (J.P.)

  10. Applications of a surveillance and diagnostics methodology using neutron noise from a pressurized-water reactor

    International Nuclear Information System (INIS)

    Wood, R.T.; Miller, L.F.; Perez, R.B.

    1992-01-01

    Two applications of a noise diagnostic methodology were performed with ex-core neutron detector data from a pressurized-water reactor (PWR). A feedback dynamics model of the neutron power spectral denisty was derived from a low-order whole-plant physical model made stochastic with the Langevin technique. From a functional fit to plant data, the response of the dynamic system to changes in important physical parameters was evaluated by a direct sensitivity analysis. In addition, changes in monitored spectra were related to changes in physical parameters, and detection thresholds using common surveillance discriminants were determined. A resonance model was developed from perturbation theory to give the ex-core neutron detector response for small in-core mechanical motions in terms of a pole-strength factor, a resonance asymmetry (or skewness) factor, a vibration damping factor, and a frequency of vibration. The mechanical motion paramters for several resonances were determined by a functional fit of the model to plant data taken at various times during a fuel cycle and were tracked to determined trends that indicated vibrational changes of reactor internals. In addition, the resonance model gave the ability to separate the resonant components of the power spectral density after the parameters had been identified. As a result, the behavior of several vibration peaks was monitored over a fuel cycle. The noise diagnostic methodology illustrated by these applications can be used in monitoring the condition of the reactor system. Early detection of degraded mechanical components or undesirable operating conditions by using such surveillance and diagnostic techniques would enhance plant safety. 15 refs., 6 figs., 1 tab

  11. Research and Development Program in Reactor Diagnostics and Monitoring with Neutron Noise Methods, Stage 18

    International Nuclear Information System (INIS)

    Pazsit, Imre; Nam, Tran Hoai; Dykin, Victor; Jonsson, Anders

    2013-01-01

    This report constitutes Stage 18 of a long-term research and development program concerning the development of diagnostics and monitoring methods for nuclear reactors. The objective of the research program is to contribute to the strategic research goal of competence and research capacity by building up competence within the Department of Nuclear Engineering at Chalmers University of Technology, regarding reactor physics, reactor dynamics and noise diagnostics. The purpose is also to contribute to the research goal of giving a basis for SSM's supervision by developing methods for identification and localization of perturbations in reactor cores. Results up to Stage 17 were reported in SKI and SSM reports, as listed in the report's summary

  12. Cooperative research and development for artificial intelligence based reactor diagnostic system

    International Nuclear Information System (INIS)

    Reifman, J.; Wei, T.Y.C.; Abboud, R.G.; Chasensky, T.M.

    1994-01-01

    Artificial Intelligence (AI) techniques in the form of knowledge-based Expert Systems (ESs) have been proposed to provide on-line decision-making support for plant operators during both normal and emergency conditions. However, in spite of the great interest in these advanced techniques, their application in the diagnosis of large-scale processes has not yet reached its full potential because of limitations of the knowledge base. These limitations include problems with knowledge acquisition and the use of an event-oriented approach for process diagnosis. To investigate the capabilities of this two-level hierarchical knowledge structure, Commonwealth Research Corporation (CRC) and Argonne National Laboratory (ANL)are collaborating on a DOE-sponsored Cooperative Research and Development Agreement (CRADA) project to perform feasibility studies on the proposed diagnostic system. Investigations are being performed in the construction of a physics-based plant level process diagnostic ES and the characterization of component-level fault project is to develop a computer-based system using this Al approach to assist process plant operators during off-normal plant conditions. The proposed computer-based system will use T-H signals complemented by other non-T-H signals available in the data stream to provide the process operator with the component which most likely caused the observed process disturbance. To demonstrate the scale-up feasibility of the proposed diagnostic system it is being developed for use with the Chemical Volume Control System (CVCS) of a nuclear power plant. This is an ongoing multi-year project and the remainder of this paper presents a mid-term status report

  13. Development of advanced strain diagnostic techniques for reactor environments.

    Energy Technology Data Exchange (ETDEWEB)

    Fleming, Darryn D.; Holschuh, Thomas Vernon,; Miller, Timothy J.; Hall, Aaron Christopher; Urrea, David Anthony,; Parma, Edward J.,

    2013-02-01

    The following research is operated as a Laboratory Directed Research and Development (LDRD) initiative at Sandia National Laboratories. The long-term goals of the program include sophisticated diagnostics of advanced fuels testing for nuclear reactors for the Department of Energy (DOE) Gen IV program, with the future capability to provide real-time measurement of strain in fuel rod cladding during operation in situ at any research or power reactor in the United States. By quantifying the stress and strain in fuel rods, it is possible to significantly improve fuel rod design, and consequently, to improve the performance and lifetime of the cladding. During the past year of this program, two sets of experiments were performed: small-scale tests to ensure reliability of the gages, and reactor pulse experiments involving the most viable samples in the Annulated Core Research Reactor (ACRR), located onsite at Sandia. Strain measurement techniques that can provide useful data in the extreme environment of a nuclear reactor core are needed to characterize nuclear fuel rods. This report documents the progression of solutions to this issue that were explored for feasibility in FY12 at Sandia National Laboratories, Albuquerque, NM.

  14. Space nuclear reactor system diagnosis: Knowledge-based approach

    International Nuclear Information System (INIS)

    Ting, Y.T.D.

    1990-01-01

    SP-100 space nuclear reactor system development is a joint effort by the Department of Energy, the Department of Defense and the National Aeronautics and Space Administration. The system is designed to operate in isolation for many years, and is possibly subject to little or no remote maintenance. This dissertation proposes a knowledge based diagnostic system which, in principle, can diagnose the faults which can either cause reactor shutdown or lead to another serious problem. This framework in general can be applied to the fully specified system if detailed design information becomes available. The set of faults considered herein is identified based on heuristic knowledge about the system operation. The suitable approach to diagnostic problem solving is proposed after investigating the most prevalent methodologies in Artificial Intelligence as well as the causal analysis of the system. Deep causal knowledge modeling based on digraph, fault-tree or logic flowgraph methodology would present a need for some knowledge representation to handle the time dependent system behavior. A proposed qualitative temporal knowledge modeling methodology, using rules with specified time delay among the process variables, has been proposed and is used to develop the diagnostic sufficient rule set. The rule set has been modified by using a time zone approach to have a robust system design. The sufficient rule set is transformed to a sufficient and necessary one by searching the whole knowledge base. Qualitative data analysis is proposed in analyzing the measured data if in a real time situation. An expert system shell - Intelligence Compiler is used to develop the prototype system. Frames are used for the process variables. Forward chaining rules are used in monitoring and backward chaining rules are used in diagnosis

  15. Development of fault diagnostic technique using reactor noise analysis

    International Nuclear Information System (INIS)

    Park, Jin Ho; Kim, J. S.; Oh, I. S.; Ryu, J. S.; Joo, Y. S.; Choi, S.; Yoon, D. B.

    1999-04-01

    The ultimate goal of this project is to establish the analysis technique to diagnose the integrity of reactor internals using reactor noise. The reactor noise analyses techniques for the PWR and CANDU NPP(Nuclear Power Plants) were established by which the dynamic characteristics of reactor internals and SPND instrumentations could be identified, and the noise database corresponding to each plant(both Korean and foreign one) was constructed and compared. Also the change of dynamic characteristics of the Ulchin 1 and 2 reactor internals were simulated under presumed fault conditions. Additionally portable reactor noise analysis system was developed so that real time noise analysis could directly be able to be performed at plant site. The reactor noise analyses techniques developed and the database obtained from the fault simulation, can be used to establish a knowledge based expert system to diagnose the NPP's abnormal conditions. And the portable reactor noise analysis system may be utilized as a substitute for plant IVMS(Internal Vibration Monitoring System). (author)

  16. Development of fault diagnostic technique using reactor noise analysis

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jin Ho; Kim, J. S.; Oh, I. S.; Ryu, J. S.; Joo, Y. S.; Choi, S.; Yoon, D. B

    1999-04-01

    The ultimate goal of this project is to establish the analysis technique to diagnose the integrity of reactor internals using reactor noise. The reactor noise analyses techniques for the PWR and CANDU NPP(Nuclear Power Plants) were established by which the dynamic characteristics of reactor internals and SPND instrumentations could be identified, and the noise database corresponding to each plant(both Korean and foreign one) was constructed and compared. Also the change of dynamic characteristics of the Ulchin 1 and 2 reactor internals were simulated under presumed fault conditions. Additionally portable reactor noise analysis system was developed so that real time noise analysis could directly be able to be performed at plant site. The reactor noise analyses techniques developed and the database obtained from the fault simulation, can be used to establish a knowledge based expert system to diagnose the NPP's abnormal conditions. And the portable reactor noise analysis system may be utilized as a substitute for plant IVMS(Internal Vibration Monitoring System). (author)

  17. Power reactors operational diagnosis

    International Nuclear Information System (INIS)

    Dach, K.; Pecinka, L.

    1976-01-01

    The definition of reactor operational diagnostics is presented and the fundamental trends of research are determined. The possible sources of power reactor malfunctions, the methods of defect detection, the data evaluation and the analysis of the results are discussed in detail. In view of scarcity of a theoretical basis and of insufficient in-core instrumentation, operational diagnostics cannot be as yet incorporated in a computer-aided reactor control system. (author)

  18. Fuel-motion diagnostics and cineradiography

    International Nuclear Information System (INIS)

    DeVolpi, A.

    1982-09-01

    Nuclear and non-nuclear applications of cineradiography are reviewed, with emphasis on diagnostic instrumentation for in-pile transient-reactor safety testing of nuclear fuel motion. The primary instrument for this purpose has been the fast-neutron hodoscope, which has achieved quantitative monitoring of time, location, mass, and velocity of fuel movement under the difficult conditions associated with transient-reactor experiments. Alternative diagnostic devices that have been developed have not matched the capabilities of the hodoscope. Other applications for the fuel-motion diagnostic apparatus are also evolving, including time-integrated radiography and direct time- and space-resolved fuel-pin power monitoring. Although only two reactors are now actively equipped with high-resolution fuel-motion diagnostic systems, studies and tests have been carried out in and for many other reactors

  19. 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

  20. Development and testing of a diagnostic system for intelligen distributed control at EBR-2

    International Nuclear Information System (INIS)

    Edwards, R.M.; Ruhl, D.W.; Klevans, E.H.; Robinson, G.E.

    1990-01-01

    A diagnostic system is under development for demonstration of Intelligent Distributed Control at the Experimental Breeder Reactor (EBR--II). In the first phase of the project a diagnostic system is being developed for the EBR-II steam plant based on the DISYS expert systems approach. Current testing uses recorded plant data and data from simulated plant faults. The dynamical simulation of the EBR-II steam plant uses the Babcock and Wilcox (B ampersand W) Modular Modeling System (MMS). At EBR-II the diagnostic system operates in the UNIX workstation and receives live plant data from the plant Data Acquisition System (DAS). Future work will seek implementation of the steam plant diagnostic in a distributed manner using UNIX based computers and Bailey microprocessor-based control system. 10 refs., 6 figs

  1. 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

  2. Symptom based diagnostic system using artificial neural networks

    International Nuclear Information System (INIS)

    Santosh; Vinod, Gopika; Saraf, R.K.

    2003-01-01

    Nuclear power plant experiences a number of transients during its operations. In case of such an undesired plant condition generally known as an initiating event, the operator has to carry out diagnostic and corrective actions. The operator's response may be too late to mitigate or minimize the negative consequences in such scenarios. The objective of this work is to develop an operator support system based on artificial neural networks that will assist the operator to identify the initiating events at the earliest stages of their developments. A symptom based diagnostic system has been developed to investigate the initiating events. Neutral networks are utilized for carrying out the event identification by continuously monitoring process parameters. Whenever an event is detected, the system will display the necessary operator actions along with the initiating event. The system will also show the graphical trend of process parameters that are relevant to the event. This paper describes the features of the software that is used to monitor the reactor. (author)

  3. Diagnostic systems developed in NPPRI (VUJE) Trnava Inc. for NPPs

    International Nuclear Information System (INIS)

    Kucharek, P.

    1998-01-01

    Since foundation of Nuclear Power Plant Research Institute (NPPRI) in 1977, the department of diagnostics has been dealt with problems related to the theoretical, practical and organizatory questions of operational diagnostics connected with PWR type nuclear components. This department acts directly in locality of NPP Jaslovske Bohunice, but there are performances for all NPP in Slovak or Czech Republic (Dukovany, Mochovce, and Temelin). Besides direct services and achievements for NPP there exist advisory, experts and research activities for the government and supervising authorities, too. In 1985, NPPRI began systematically construct and verify technical means for operational diagnostics of main circulating pumps (MCP) with good results, based on own rich practical experiences and contacts with organisations abroad. In recent years NPPRI as one of recognised qualified and authorised institutions in Slovak Republic has begun to develop a new generation of diagnostic systems for NPP on high technical level but with lower procuring costs in comparison with western countries products. This contribution deals with four following types of diagnostic systems which were not only developed but also delivered and installed on Slovak and Czech nuclear units: - Loose part monitoring system (LPMS), - Humidity monitoring system (HUMON), - Reactor coolant pumps monitoring system (RCPMS), - Primary circuit vibration monitoring system (VMS). Main features of new generation from middle of 1990's of these systems are described in this paper and operational experiences with them too. (author)

  4. Final results from the development of the diagnostic expert system DESYRE

    International Nuclear Information System (INIS)

    Scherer, K.P.; Eggert, H.; Sheleisiek, K.; Stille, P.; Schoeller, H.

    1997-01-01

    In the Kernforschungszentrum Karlsruhe (KfK), a distributed knowledge based diagnostic system is developed to supervise the primary system including the core of the Kompakte Natriumgekuehlte Kernreaktoranlage (KNK II), a 20 MWe experimental fast reactor. The problem is to detect anomalies and disturbances in the beginning state before fault propagation - early diagnosis - and provide the scram analysis to detect the causality when a system shutdwon occurs. (author). 9 refs, 15 figs

  5. Pattern recognition, neural networks, genetic algorithms and high performance computing in nuclear reactor diagnostics. Results and perspectives

    International Nuclear Information System (INIS)

    Dzwinel, W.; Pepyolyshev, N.

    1996-01-01

    The main goal of this paper is the presentation of our experience in development of the diagnostic system for the IBR-2 (Russia - Dubna) nuclear reactor. The authors show the principal results of the system modifications to make it work more reliable and much faster. The former needs the adaptation of new techniques of data processing, the latter, implementation of the newest computational facilities. The results of application of the clustering techniques and a method of visualization of the multi-dimensional information directly on the operator display are presented. The experiences with neural nets, used for prediction of the reactor operation, are discussed. The genetic algorithms were also tested, to reduce the quantity of data nd extracting the most informative components of the analyzed spectra. (authors)

  6. Modern diagnostic systems for loose parts, vibration and leakage monitoring

    International Nuclear Information System (INIS)

    Kunze, U.

    1997-01-01

    The modern diagnostic systems for loose parts, vibration and leakage monitoring of Siemens marked improvements in signal detection, ease of operation, and the display of information. The paper gives an overview on: Loose parts monitoring system KUeS '95 - a computer-based system. The knowledge and experience about loose parts detection incorporated into this system can be characterized as ''intelligence''. Vibration monitoring system SUeS '95 - a fully automated system for early detection of changes in the vibration patterns of the reactor coolant system components and reactor pressure vessel internals. Leak detection system FLUeS - a system that detects even small leaks in steam-carrying components and very accurately determines their location. Leaks are detected on the moisture distribution in a sample air column into which the escaping steam locally diffuses. All systems described represent the latest state of technology. Nevertheless a considerable amount of operational experience can be reported. (author). 5 refs, 10 figs

  7. An Advanced Diagnostic Display for Core Protection Calculator System

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Ji-Hyeon; Jeong, See-Chae; Sohn, Se-Do [Korea Power Engineering Company, Daejeon (Korea, Republic of)

    2008-10-15

    The main purpose of a Nuclear Power Plant Instrumentation and Control (I and C) Display System is to provide operator's interface for I and C systems. The CPCS display(Shin-Kori 1 and 2) provides operators with 1) plant monitoring values of field input and algorithm variables that reflect the reactor core conditions, 2) operation values that operators can change and 3) CPCS status. It will be an optimal case if operators can understand the plant (including CPCS itself) condition intuitively with the displayed values but it is not easy in CPCS. For example, if the CPCS Channel Trouble light is lit, operators need some amount of time to investigate what caused the trouble light because there are more than hundred causes that can generate the channel trouble. If a Display supports diagnostic information that shows what cause the displayed alarms, it will greatly help operators in easy understanding the CPCS status. To provide these diagnostic information, this paper suggests an active self-explanatory display mechanism. This self-explanatory diagnostic display mechanism utilizes an ontology in XML that describes parent child, sibling relationships of display variables, through which in-depth, in-breadth diagnostic tracking is possible. This paper consists of two parts. First, the key features of CPCS Flat Panel Display System (FPDS) are described. Second, the features of active self explanatory diagnostic display are discussed.

  8. An Advanced Diagnostic Display for Core Protection Calculator System

    International Nuclear Information System (INIS)

    Kim, Ji-Hyeon; Jeong, See-Chae; Sohn, Se-Do

    2008-01-01

    The main purpose of a Nuclear Power Plant Instrumentation and Control (I and C) Display System is to provide operator's interface for I and C systems. The CPCS display(Shin-Kori 1 and 2) provides operators with 1) plant monitoring values of field input and algorithm variables that reflect the reactor core conditions, 2) operation values that operators can change and 3) CPCS status. It will be an optimal case if operators can understand the plant (including CPCS itself) condition intuitively with the displayed values but it is not easy in CPCS. For example, if the CPCS Channel Trouble light is lit, operators need some amount of time to investigate what caused the trouble light because there are more than hundred causes that can generate the channel trouble. If a Display supports diagnostic information that shows what cause the displayed alarms, it will greatly help operators in easy understanding the CPCS status. To provide these diagnostic information, this paper suggests an active self-explanatory display mechanism. This self-explanatory diagnostic display mechanism utilizes an ontology in XML that describes parent child, sibling relationships of display variables, through which in-depth, in-breadth diagnostic tracking is possible. This paper consists of two parts. First, the key features of CPCS Flat Panel Display System (FPDS) are described. Second, the features of active self explanatory diagnostic display are discussed

  9. Investigation of neural network paradigms for the development of automatic noise diagnostic/reactor surveillance systems

    International Nuclear Information System (INIS)

    Korsah, K.; Uhrig, R.E.

    1991-01-01

    The use of artificial intelligence (AI) techniques as an aid in the maintenance and operation of nuclear power plant systems has been recognized for the past several years, and several applications using expert systems technology currently exist. The authors investigated the backpropagation paradigm for the recognition of neutron noise power spectral density (PSD) signatures as a possible alternative to current methods based on statistical techniques. The goal is to advance the state of the art in the application of noise analysis techniques to monitor nuclear reactor internals. Continuous surveillance of reactor systems for structural degradation can be quite cost-effective because (1) the loss of mechanical integrity of the reactor internal components can be detected at an early stage before severe damage occurs, (2) unnecessary periodic maintenance can be avoided, (3) plant downtime can be reduced to a minimum, (4) a high level of plant safety can be maintained, and (5) it can be used to help justify the extension of a plant's operating license. The initial objectives were to use neutron noise PSD data from a pressurized water reactor, acquired over a period of ∼2 years by the Oak Ridge National Laboratory (ORNL) Power Spectral Density RECognition (PSDREC) system to develop networks that can (1) differentiate between normal neutron spectral data and anomalous spectral data (e.g., malfunctioning instrumentation); and (2) detect significant shifts in the positions of spectral resonances while reducing the effect of small, random shifts (in neutron noise analysis, shifts in the resonance(s) present in a neutron PSD spectrum are the primary means for diagnosing degradation of reactor internals). 11 refs, 8 figs

  10. Safety aspects of neutron noise diagnostics and loose parts monitoring in WWER reactors

    International Nuclear Information System (INIS)

    Por, G.

    1997-01-01

    In this paper our aim is to give very short introduction on different types of well selected noise diagnostics methods and then mentioning their occurrence in WWER reactors we analyze what impact they might have to operational safety and for ageing (which also affects on safety of the installations). We do not deny, that one of our main aim is to call the attention of management staff of NPP, which deals with safety, safety culture, maintenance and operation proving, that such methods and system can give not only benefit to economy but also impact on safety of nuclear installations. 14 refs, 6 figs

  11. MCNPX simulations of fast neutron diagnostics for accelerator-driven systems

    Energy Technology Data Exchange (ETDEWEB)

    Habob, Moinul

    2005-12-15

    In accelerator-driven systems, the neutron spectrum will extend all the way up to the incident beam energy, i.e., several hundred MeV or even up to GeV energies. The high neutron energy allows novel diagnostics with a set of measurement techniques that can be used in a sub-critical reactor environment. Such measurements are primarily connected to system safety and validation. This report shows that in-core fast-neutron diagnostics can be employed to monitor changes in the position of incidence of the primary proton beam onto the neutron production target. It has also been shown that fast neutrons can be used to detect temperature-dependent density changes in a liquid lead-bismuth target. Fast neutrons can escape the system via the beam pipe for the incident proton beam. Out-of-core monitoring of these so called back-streaming neutrons could potentially be used to monitor beam changes if the target has a suitable shape. Moreover, diagnostics of back-streaming neutrons might be used for validation of the system design.

  12. MCNPX simulations of fast neutron diagnostics for accelerator-driven systems

    International Nuclear Information System (INIS)

    Habib, Moinul

    2005-12-01

    In accelerator-driven systems, the neutron spectrum will extend all the way up to the incident beam energy, i.e., several hundred MeV or even up to GeV energies. The high neutron energy allows novel diagnostics with a set of measurement techniques that can be used in a sub-critical reactor environment. Such measurements are primarily connected to system safety and validation. This report shows that in-core fast-neutron diagnostics can be employed to monitor changes in the position of incidence of the primary proton beam onto the neutron production target. It has also been shown that fast neutrons can be used to detect temperature-dependent density changes in a liquid lead-bismuth target. Fast neutrons can escape the system via the beam pipe for the incident proton beam. Out-of-core monitoring of these so called back-streaming neutrons could potentially be used to monitor beam changes if the target has a suitable shape. Moreover, diagnostics of back-streaming neutrons might be used for validation of the system design

  13. 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.)

  14. Design of a Fast Neutral He Beam System for Feasibility Study of Charge-Exchange Alpha-Particle Diagnostics in a Thermonuclear Fusion Reactor

    CERN Document Server

    Shinto, Katsuhiro; Kitajima, Sumio; Kiyama, Satoru; Nishiura, Masaki; Sasao, Mamiko; Sugawara, Hiroshi; Takenaga, Mahoko; Takeuchi, Shu; Wada, Motoi

    2005-01-01

    For alpha-particle diagnostics in a thermonuclear fusion reactor, neutralization using a fast (~2 MeV) neutral He beam produced by the spontaneous electron detachment of a He- is considered most promising. However, the beam transport of produced fast neutral He has not been studied, because of difficulty for producing high-brightness He- beam. Double-charge-exchange He- sources and simple beam transport systems were developed and their results were reported in the PAC99* and other papers.** To accelerate an intense He- beam and verify the production of the fast neutral He beam, a new test stand has been designed. It consists of a multi-cusp He+

  15. Operation diagnostics of the reactor coolant pumps in the Jaslovske Bohunice nuclear power plant, CSSR

    International Nuclear Information System (INIS)

    Bahna, J.; Jaros, I.; Oksa, G.

    1990-01-01

    The state of the art of the materials basis, the diagnostics methods used, organization of data collection and processing, and some results of routine and specific investigations concerned with diagnosis of the reactor coolant pump in the Jaslovske Bohunice NPP V-1 are presented. Some information is given about the reactor coolant pump monitor developed in the VUJE. (author)

  16. Design Features of the Neutral Particle Diagnostic System for the ITER Tokamak

    Science.gov (United States)

    Petrov, S. Ya.; Afanasyev, V. I.; Melnik, A. D.; Mironov, M. I.; Navolotsky, A. S.; Nesenevich, V. G.; Petrov, M. P.; Chernyshev, F. V.; Kedrov, I. V.; Kuzmin, E. G.; Lyublin, B. V.; Kozlovski, S. S.; Mokeev, A. N.

    2017-12-01

    The control of the deuterium-tritium (DT) fuel isotopic ratio has to ensure the best performance of the ITER thermonuclear fusion reactor. The diagnostic system described in this paper allows the measurement of this ratio analyzing the hydrogen isotope fluxes (performing neutral particle analysis (NPA)). The development and supply of the NPA diagnostics for ITER was delegated to the Russian Federation. The diagnostics is being developed at the Ioffe Institute. The system consists of two analyzers, viz., LENPA (Low Energy Neutral Particle Analyzer) with 10-200 keV energy range and HENPA (High Energy Neutral Particle Analyzer) with 0.1-4.0MeV energy range. Simultaneous operation of both analyzers in different energy ranges enables researchers to measure the DT fuel ratio both in the central burning plasma (thermonuclear burn zone) and at the edge as well. When developing the diagnostic complex, it was necessary to account for the impact of several factors: high levels of neutron and gamma radiation, the direct vacuum connection to the ITER vessel, implying high tritium containment, strict requirements on reliability of all units and mechanisms, and the limited space available for accommodation of the diagnostic hardware at the ITER tokamak. The paper describes the design of the diagnostic complex and the engineering solutions that make it possible to conduct measurements under tokamak reactor conditions. The proposed engineering solutions provide a safe—with respect to thermal and mechanical loads—common vacuum channel for hydrogen isotope atoms to pass to the analyzers; ensure efficient shielding of the analyzers from the ITER stray magnetic field (up to 1 kG); provide the remote control of the NPA diagnostic complex, in particular, connection/disconnection of the NPA vacuum beamline from the ITER vessel; meet the ITER radiation safety requirements; and ensure measurements of the fuel isotopic ratio under high levels of neutron and gamma radiation.

  17. DIVA and DIAPO: two diagnostic knowledge based systems used for French nuclear power plants

    International Nuclear Information System (INIS)

    Porcheron, M.; Ricard, B.; Joussellin, A.

    1997-01-01

    In order to improve monitoring and diagnosis capabilities in nuclear power plants, Electricite de France (EDF) has designed an integrated monitoring and diagnosis assistance system: PSAD-Poste de Surveillance et d'Aide au Diagnostic. The development of such a sophisticated monitoring and data processing systems has emphasized the need for the addition of analysis and diagnosis assistance capabilities. Therefore, diagnostic knowledge based systems have been added to the functions monitored in PSAD: DIVA for turbine generators, and DIAPO for reactor coolant pumps. These systems were designed from a representation of the diagnostic reasoning process of experts and of the supporting knowledge. Diagnosis in both systems relies on an abduction reasoning process applied to component fault models and observations derived from their actual behavior, as provided by the monitoring functions. The basic theoretical elements of this diagnostic model are summarized in a first part. In a second part, DIVA and DIAPO specific elements are described

  18. Component vibration of VVER-reactors - diagnostics and modelling

    International Nuclear Information System (INIS)

    Altstadt, E.; Scheffler, M.; Weiss, F.-P.

    1995-01-01

    Flow induced vibrations of reactor pressure vessel (RPV) internals (control element and core barrel motions) at VVER-440 reactors have led to the development of dedicated methods for on-line monitoring. These methods need a certain developed stage of the faults to be detected. To achieve a real sensitive early detection of mechanical faults of RPV internals, a theoretical vibration model was developed based on finite elements. The model comprises the whole primary circuit including the steam generators (SG). By means of that model all eigenfrequencies up to 30 Hz and the corresponding mode shapes were calculated for the normal vibration behaviour. Moreover the shift of eigenfrequencies and of amplitudes due to the degradation or to the failure of internal clamping and spring elements could be investigated, showing that a recognition of such degradations even inside the RPV is possible by pure excore vibration measurements. A true diagnostic, that is the identification of the failed component, might become possible because different faults influence different and well separated eigenfrequencies. (author)

  19. Modernization incore monitoring system of WWER-1000 reactors (V-320)

    International Nuclear Information System (INIS)

    Mitin, Valentin; Semchenkov, Yurij; Kalinushkin, Andrey

    2008-01-01

    Modern ICIS system for VVER-1000, including a number of sensors, cable runs, corresponding measuring equipment and computer engineering, software, accumulated 30 year experience of interaction researches on VVER reactors and is capable to ensure carrying out of control, protection, informational, diagnostic functions and thus to promote real increase of quality, reliability and safety in nuclear fuel and NPP power units operation

  20. Implementation of New Reactivity Measurement System and New Reactor Noise Analysis Equipment in a VVER-440 Nuclear Power Plant

    Science.gov (United States)

    Vegh, János; Kiss, Sándor; Lipcsei, Sándor; Horvath, Csaba; Pos, István; Kiss, Gábor

    2010-10-01

    The paper deals with two recently developed, high-precision nuclear measurement systems installed at the VVER-440 units of the Hungarian Paks NPP. Both developments were motivated by the reactor power increase to 108%, and by the planned plant service time extension. The first part describes the RMR start-up reactivity measurement system with advanced services. High-precision picoampere meters were installed at each reactor unit and measured ionization chamber current signals are handled by a portable computer providing data acquisition and online reactivity calculation service. Detailed offline evaluation and analysis of reactor start-up measurements can be performed on the portable unit, too. The second part of the paper describes a new reactor noise diagnostics system using state-of-the-art data acquisition hardware and signal processing methods. Details of the new reactor noise measurement evaluation software are also outlined. Noise diagnostics at Paks NPP is a standard tool for core anomaly detection and for long-term noise trend monitoring. Regular application of these systems is illustrated by real plant data, e.g., results of standard reactivity measurements during a reactor startup session are given. Noise applications are also illustrated by real plant measurements; results of core anomaly detection are presented.

  1. An integrated real-time diagnostic concept using expert systems, qualitative reasoning and quantitative analysis

    International Nuclear Information System (INIS)

    Edwards, R.M.; Lee, K.Y.; Kumara, S.; Levine, S.H.

    1989-01-01

    An approach for an integrated real-time diagnostic system is being developed for inclusion as an integral part of a power plant automatic control system. In order to participate in control decisions and automatic closed loop operation, the diagnostic system must operate in real-time. Thus far, an expert system with real-time capabilities has been developed and installed on a subsystem at the Experimental Breeder Reactor (EBR-II) in Idaho, USA. Real-time simulation testing of advanced power plant concepts at the Pennsylvania State University has been developed and was used to support the expert system development and installation at EBR-II. Recently, the US National Science Foundation (NSF) and the US Department of Energy (DOE) have funded a Penn State research program to further enhance application of real-time diagnostic systems by pursuing implementation in a distributed power plant computer system including microprocessor based controllers. This paper summarizes past, current, planned, and possible future approaches to power plant diagnostic systems research at Penn State. 34 refs., 9 figs

  2. 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)

  3. The system for diagnostics and monitoring of the IBR-2 reactor state. Data acquisition, accumulation and storage of information

    International Nuclear Information System (INIS)

    Ermilov, V.G.; Ivanov, V.V.; Korolev, V.S.; Pepelyshev, Yu.N.; Semashko, S.V.; Tulaev, A.B.

    2000-01-01

    The architectural decisions for a developed distributed system of the IBR-2 pulsed reactor conditions monitoring are described. The system is intended for measurement of the basic reactor parameters, acquisition, storage and processing of information, the current reactor state monitoring, analysis of reactor parameters for a long time operation period both in on-line, and in off-line modes. The system is constructed in the architecture client-server using DBMS MS SQL Server 7.0 The basic hardware components of the system are measuring workstations and devices, processing and user workstations and the central server. The software of the system consists of the measuring programs, data flows dispatching services, client applications for data processing and visualization, and means for preparing data for subsequent presentation in WWW. The basic results of the first system operation phase and prospect of its development are discussed. (author)

  4. The graphics-based human interface to the DISYS diagnostic/control guidance system at EBR-2

    International Nuclear Information System (INIS)

    Edwards, R.M.; Chavez, C.; Kamarthi, S.; Dharap, S.; Lindsay, R.W.; Staffon, J.

    1990-01-01

    An initial graphics based interface to the real-time DISYS diagnostic system has been developed using the multi-tasking capabilities of the UNIX operating system and X-Windows 11 Xlib graphics library. This system is interfaced to live plant data at the Experimental Breeder Reactor (EBR-2) for the Argon Cooling System of fuel handling operations and the steam plant. The interface includes an intelligent process schematic which highlights problematic components and sensors based on the results of the diagnostic computations. If further explanation of a faulted component is required, the user can call up a display of the diagnostic computations presented in a tree-like diagram. Numerical data on the process schematic and optional diagnostic tree are updated as new real-time data becomes available. The initial X-Windows 11 based interface will be further enhanced using VI Corporation DATAVIEWS graphical data base software. 5 refs., 6 figs

  5. 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)

  6. The role of risk management in the design of diagnostics for fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ingesson, L. C. [Fusion for Energy, Josep Pla 2, Torres Diagonal Litoral B3, 08019 Barcelona (Spain); Collaboration: F4E Diagnostic Project Team

    2014-08-21

    A project-oriented approach is beneficial for the selection and design of viable diagnostics for fusion reactors because of the associated complex physical and organizational environment. The project-oriented approach includes rigorous risk management. The nature and impact of risks related to technical, organizational and commercial aspects in relation to the development of ITER diagnostics under EU responsibility are analyzed. The majority of risks are related to organizational aspects and technical feasibility issues. The experience with ITER is extrapolated to DEMO and beyond. It should not be taken for granted that technical solutions will be found, while a risk analysis of various diagnostic techniques with quantitative assessments undertaken early in the design of DEMO would be beneficial.

  7. Operational experiences on the Borssele nuclear power plant using computer based surveillance and diagnostic system on-line

    International Nuclear Information System (INIS)

    Turkcan, E.; Quaadvliet, W.H.J.; Peeters, T.T.J.M.; Verhoef, J.P

    1991-06-01

    The on-line monitoring and diagnostics system of Borssele nuclear power plant (NPP), designed and established by the ECN Energy Research Foundation, has been operating continuously since 1983. The system is extended in form of multiprocessing, multi-tasking structure performing real-time monitoring, on-line reactor parameters' calculation, data-base preparation for expert systems and providing early information on possible malfunctions even in the incipient stage making alert by passive alarms. The system realized has already been operating in the course of 7 fuel cycles of the reactor starting from start-up through normal power operation. An expert system operating on the VAX work station is added to the surveillance and diagnostics system for data base management of the observed physical parameters relevant to the NPP under supervision. The paper highlights the surveillance and diagnostic modules involved, in their actual hierarchical form in use, presents theoretical considerations applied to the design of the surveillance system together with the results obtained through the 12th to 17th fuel cycles of the NPP including start-ups and shut-downs and reveals the experience thus gained by both utility and ECN through the application of the system described. (author). 19 refs.; 4 figs

  8. Pressurized water reactor monitoring. Study of detection, diagnostic and estimation (least squares and filtering) methods

    International Nuclear Information System (INIS)

    Gillet, M.

    1986-07-01

    This thesis presents a study for the surveillance of the Primary circuit water inventory of a pressurized water reactor. A reference model is developed for the development of an automatic system ensuring detection and real-time diagnostic. The methods to our application are statistical tests and adapted a pattern recognition method. The estimation of the detected anomalies is treated by the least square fit method, and by filtering. A new projected optimization method with superlinear convergence is developed in this framework, and a segmented linearization of the model is introduced, in view of a multiple filtering. 46 refs [fr

  9. 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)

  10. Capabilities and diagnostics of the Sandia Pelletron-raster system

    International Nuclear Information System (INIS)

    Buckalew, W.H.; Lockwood, G.J.; Luker, S.M.; Ruggles, L.E.; Wyant, F.J.

    1984-07-01

    The radiation capabilities of the PELLETRON Electron Beam Accelerator have been expanded to include a controllable, variable dimension, beam diffusion option. This rastered beam option has been studied in detail. Beam characteristics have been determined as a function of incident electron beam energy, current, and deflection system parameters. The beam diagnostics required to define any given diffuse beam pattern are accurate and predictable. Recently, utility of this added PELLETRON capability was demonstrated by simulating the effects of complex nuclear reactor accident electron environments on electrical insulation materials similar to those used in nuclear power plants

  11. Morphological evolution of copper nanoparticles: Microemulsion reactor system versus batch reactor system

    Science.gov (United States)

    Xia, Ming; Tang, Zengmin; Kim, Woo-Sik; Yu, Taekyung; Park, Bum Jun

    2017-07-01

    In the synthesis of nanoparticles, the reaction rate is important to determine the morphology of nanoparticles. We investigated morphology evolution of Cu nanoparticles in this two different reactors, microemulsion reactor and batch reactor. In comparison with the batch reactor system, the enhanced mass and heat transfers in the emulsion system likely led to the relatively short nucleation time and the highly homogeneous environment in the reaction mixture, resulting in suppressing one or two dimensional growth of the nanoparticles. We believe that this work can offer a good model system to quantitatively understand the crystal growth mechanism that depends strongly on the local monomer concentration, the efficiency of heat transfer, and the relative contribution of the counter ions (Br- and Cl-) as capping agents.

  12. On the survivability of diagnostic windows in the CIT [Compact Ignition Tokamak] reactor

    International Nuclear Information System (INIS)

    Taylor, A.

    1988-11-01

    The problem of radiation induced stresses in CIT diagnostic windows is discussed. Existing data indicate windows of existing design will probably survive if placed on the periphery of the reactor. There is a lack of adequate engineering data upon which the design and survivability of windows can be based. 22 refs., 5 figs., 2 tabs

  13. The Automatic Test Features of the IDiPS Reactor Protection System

    International Nuclear Information System (INIS)

    Hur, Seop; Kim, Dong-Hoon; Hwang, In-Koo; Lee, Cheol-Kwon; Lee, Dong-Young

    2007-01-01

    The reactor protection system (RPS) is designed to minimize a propagation of abnormal or accident conditions of nuclear power plants. A digital RPS (Integrated Digital Protection System (IDiPS) RPS) is being developed in the Korea Nuclear Instrumentation and Control System (KNICS) R and D project. To make good use of the advantages of the digital technology, it is necessary to improve the reliability and availability of a system through automatic test features including an on-line testing, a self-diagnostics, an auto calibration, etc. This paper summarizes the system test strategy and the automatic test features of the IDiPS RPS

  14. Noise Diagnostics of Stationary and Non-Stationary Reactor Processes

    Energy Technology Data Exchange (ETDEWEB)

    Sunde, Carl

    2007-04-15

    This thesis concerns the application of noise diagnostics on different problems in the area of reactor physics involving both stationary and non-stationary core processes. Five different problems are treated, divided into three different parts. The first problem treated in the first part is the classification of two-phase flow regimes from neutron radiographic and visible light images with a neuro-wavelet algorithm. The algorithm consists of wavelet pre-processing and of an artificial neural network. The result indicates that the wavelet pre-processing is improving the training of the neural network. Next, detector tubes which are suspected of impacting on nearby fuel-assemblies in a boiling water reactor (BWR) are identified by both a classical spectral method and wavelet-based methods. It was found that there is good agreement between the different methods as well as with visual inspections of detector tube and fuel assembly damage made during the outage at the plant. The third problem addresses the determination of the decay ratio of a BWR from the auto-correlation function (ACF). Here wavelets are used, with some success, both for de-trending and de-nosing of the ACF and also for direct estimation of the decay ratio from the ACF. The second part deals with the analysis of beam-mode and shell-mode core-barrel vibrations in pressurised water reactors (PWRs). The beam-mode vibrations are analysed by using parameters of the vibration peaks, in spectra from ex core detectors. A trend analysis of the peak amplitude shows that the peak amplitude is changing during the fuel cycle. When it comes to the analysis of the shell-mode vibration, 1-D analytical and numerical calculations are performed in order to calculate the neutron noise induced in the core. The two calculations are in agreement and show that a large local noise component is present in the core which could be used to classify the shell-mode vibrations. However, a measurement made in the PWR Ringhals-3 shows

  15. Noise Diagnostics of Stationary and Non-Stationary Reactor Processes

    International Nuclear Information System (INIS)

    Sunde, Carl

    2007-01-01

    This thesis concerns the application of noise diagnostics on different problems in the area of reactor physics involving both stationary and non-stationary core processes. Five different problems are treated, divided into three different parts. The first problem treated in the first part is the classification of two-phase flow regimes from neutron radiographic and visible light images with a neuro-wavelet algorithm. The algorithm consists of wavelet pre-processing and of an artificial neural network. The result indicates that the wavelet pre-processing is improving the training of the neural network. Next, detector tubes which are suspected of impacting on nearby fuel-assemblies in a boiling water reactor (BWR) are identified by both a classical spectral method and wavelet-based methods. It was found that there is good agreement between the different methods as well as with visual inspections of detector tube and fuel assembly damage made during the outage at the plant. The third problem addresses the determination of the decay ratio of a BWR from the auto-correlation function (ACF). Here wavelets are used, with some success, both for de-trending and de-nosing of the ACF and also for direct estimation of the decay ratio from the ACF. The second part deals with the analysis of beam-mode and shell-mode core-barrel vibrations in pressurised water reactors (PWRs). The beam-mode vibrations are analysed by using parameters of the vibration peaks, in spectra from ex core detectors. A trend analysis of the peak amplitude shows that the peak amplitude is changing during the fuel cycle. When it comes to the analysis of the shell-mode vibration, 1-D analytical and numerical calculations are performed in order to calculate the neutron noise induced in the core. The two calculations are in agreement and show that a large local noise component is present in the core which could be used to classify the shell-mode vibrations. However, a measurement made in the PWR Ringhals-3 shows

  16. 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.)

  17. 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

  18. The computerized reactor period measurement system for China fast burst reactor-II

    International Nuclear Information System (INIS)

    Zhao Wuwen; Jiang Zhiguo

    1996-01-01

    The article simply introduces the hardware, principle, and software of the computerized reactor period measurement system for China Fast Burst Reactor-II (CFBR-II). It also gives the relation between fission yield and pre-reactivity of CFBR-II reactor system of bared reactor with decoupled-component and system of bared reactor with multiple light-material. The computerized measurement system makes the reactor period measurement into automatical and intelligent and also improves the speed and precision of period data on-line process

  19. Computer measurement system of reactor period for China fast burst reactor-II

    International Nuclear Information System (INIS)

    Zhao Wuwen; Jiang Zhiguo

    1997-01-01

    The author simply introduces the hardware, principle, and software of the reactor period computer measure system for China Fast Burst Reactor-II (CFBR-II). It also gives the relation between Fission yield and Pre-reactivity of CFBR-II reactor system of bared reactor with decoupled-component and system of bared reactor with multiple light-material. The computer measure system makes the reactor period measurement into automation and intellectualization and also improves the speed and precision of period data process on-line

  20. CATALisp - an expert system for real-time diagnosis of multiple-failure transients in pressurized water reactors

    International Nuclear Information System (INIS)

    Martin, R.P.; Nassersharif, B.

    1988-01-01

    The state of the art in artificial intelligence (AI) and expert system (ES) technology has matured to a degree that the potential development of a computer-aided/automated diagnostic and transient mitigation system in the area of nuclear reactor operation can be considered. Since traditional methods cannot handle complex systems efficiently, AI techniques provide a means to emulate an expert reactor operator rather than follow mechanistic methods. Computer-aided transient analysis coded in LISP (CATA-Lisp) is a confidence level based expert system written in Common LISP on the SYMBOLICS 3640 computer system. New versions are being developed in Common LISP for the Texas Instruments (TI) Explorer and the Sun microsystems machines, CATALisp manipulates both a knowledge base of transient identifier patterns (tree structured to allow for zooming in diagnostics) and a knowledge base containing a qualitative model of a nuclear power plant. The interference engine used by CATALisp uses the information stored in both knowledge bases to arrive at confidence level values that are used to infer particular plant states

  1. Antenna design for fast ion collective Thomson scattering diagnostic for the international thermonuclear experimental reactor

    DEFF Research Database (Denmark)

    Leipold, Frank; Furtula, Vedran; Salewski, Mirko

    2009-01-01

    Fast ion physics will play an important role for the international thermonuclear experimental reactor (ITER), where confined alpha particles will affect and be affected by plasma dynamics and thereby have impacts on the overall confinement. A fast ion collective Thomson scattering (CTS) diagnostic...

  2. 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

  3. 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.)

  4. Monitoring for shaft cracks on reactor recirculation pumps

    International Nuclear Information System (INIS)

    Kowal, M.G.; O'Brien, J.T. Jr.

    1989-01-01

    The article discusses the vibration characteristics associated with a boiling water reactor (BWR) recirculation pump. It also describes the application of diagnostic techniques and shaft crack theory to an on-line diagnostic monitoring system for reactor recirculation pumps employed at Philadelphia Electric Company's Peach Bottom Atomic Power Station. Specific emphasis is placed on the unique monitoring techniques associated with these variable speed vertical pumps

  5. 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

  6. 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

  7. Development of JT-60 diagnostics system

    International Nuclear Information System (INIS)

    Suzuki, Yasuo

    1988-01-01

    The various kinds of plasma diagnostics have been developed and utilized in the JT-60 experiments. The features of JT-60 diagnostics system and the historical proceeding of the development are described in this paper. Taking account of the design consideration, JT-60 diagnostics system is sorted out into eight groups, which include six diagnostics systems, the data processing system and diagnostics supporting system. The all devices in the JT-60 diagnostics system were instrumented on schedule in the end of the fiscal year of 1985 and have contributed to JT-60 experiments. (author)

  8. 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

  9. Application of neural networks and neutron noise for diagnostics of reactor internals vibration

    International Nuclear Information System (INIS)

    Garis, N.S.; Pazsit, I.; Gloeckler, O.

    1995-01-01

    It has long been known that vibration of reactor internals, in particular excessive vibrations of control rods, can be detected via the neutron noise they induce. Noise measurements are actually suitable to determine important diagnostic parameters such as the location of the vibrating rod and the vibration amplitude. An algorithm was earlier elaborated for this purpose, which is based on inversion of the expression describing the neutron noise as a function of vibration parameters. This inversion procedure is nevertheless complicated and not always unique. It was investigated whether a properly trained neural network can perform the inversion more effectively. It was found that artificial neural networks can be trained effectively to perform vibration diagnostics from neutron noise data fast, effectively and reliably. The present paper gives a description of the development and use of the neural networks for purposes of vibration diagnostics

  10. 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.)

  11. Low activation diagnostic equipment design studies

    International Nuclear Information System (INIS)

    Hopkins, G.R.; Cheng, E.T.; Fisher, R.K.

    1985-01-01

    The low activation fusion concept has been applied to the diagnostic equipment in a fusion reactor. The components where fabrication from low activation materials is feasible have been identified. Other systems where higher activation elements are required can have their activation reduced by design approaches which include shielding and operation only in low flux regions of the reactor. Some components will not operate in a high flux so activation is not a major concern. This low activation diagnostic equipment study completes a series of low activation studies where all the components in a fusion power reactor have now been evaluated. It is concluded that a completely low activation fusion reactor is feasible with all components meeting the functional requirements. This provides an environmentally benign energy source with a high confidence level in meeting safety criteria in operation, maintenance and waste disposal

  12. Monitoring nuclear reactor systems using neural networks and fuzzy logic

    International Nuclear Information System (INIS)

    Ikonomopoulos, A.; Tsoukalas, L.H.; Uhrig, R.E.; Mullens, J.A.

    1991-01-01

    A new approach is presented that demonstrates the potential of trained artificial neural networks (ANNs) as generators of membership functions for the purpose of monitoring nuclear reactor systems. ANN's provide a complex-to-simple mapping of reactor parameters in a process analogous to that of measurement. Through such ''virtual measurements'' the value of parameters with operational significance, e.g., control-valve-disk-position, valve-line-up or performance can be determined. In the methodology presented the output of a virtual measuring device is a set of membership functions which independently represent different states of the system. Utilizing a fuzzy logic representation offers the advantage of describing the state of the system in a condensed form, developed through linguistic descriptions and convenient for application in monitoring, diagnostics and generally control algorithms. The developed methodology is applied to the problem of measuring the disk position of the secondary flow control valve of an experimental reactor using data obtained during a start-up. The enhanced noise tolerance of the methodology is clearly demonstrated as well as a method for selecting the actual output. The results suggest that it is possible to construct virtual measuring devices through artificial neural networks mapping dynamic time series to a set of membership functions and thus enhance the capability of monitoring systems. 8 refs., 11 figs., 1 tab

  13. Real-time reactor coolant system pressure/temperature limit system

    International Nuclear Information System (INIS)

    Newton, D.G.; Schemmel, R.R.; Van Scooter, W.E. Jr.

    1991-01-01

    This patent describes an system, used in controlling the operating of a nuclear reactor coolant system, which automatically calculates and displays allowable reactor coolant system pressure/temperature limits within the nuclear reactor coolant system based upon real-time inputs. It comprises: means for producing signals representative of real-time operating parameters of the nuclear reactor cooling system; means for developing pressure and temperature limits relating the real-time operating parameters of the nuclear reactor coolant system, for normal and emergency operation thereof; means for processing the signals representative of real-time operating parameters of the nuclear reactor coolant system to perform calculations of a best estimate of signals, check manual inputs against permissible valves and test data acquisition hardware for validity and over/under range; and means for comparing the representative signals with limits for the real-time operating parameters to produce a signal for a real-time display of the pressure and temperature limits and of the real-time operating parameters use an operator in controlling the operation of the nuclear reactor coolant system

  14. Neutrino remote diagnostics of in-reactor processes

    CERN Document Server

    Rusov, V D; Shaaban, I

    2002-01-01

    The correlation passive location of spontaneous chain reaction inside reactor sources algorithm structures are obtained. The considered algorithm structures could be the base for practical realisation of neutrino sources passive location system. The automatics distance system of continues control for energy-generation and radiation creep of reactor fuel are considered. The model of a radiation creep is explained within the framework of the mechanism of gliding and climbing dislocations based on the conception of a dislocation as not ideal sink for point radiation defects (PRD). The used model is efficient for installed PRD concentration,considerably exceeding thermally steady state concentration. The gliding of dislocation are describing as due to moving dislocation kinks in Peierls relief. The climbing of dislocation are describing as due to moving dislocation jogs. The complex of the computer programs simulating the radiation creep needed the same output parameters: PRD concentration, which calculated by ne...

  15. 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

  16. 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

  17. Research and Development Program in Reactor Diagnostics and Monitoring with Neutron Noise Methods. Stages 14 and 15

    International Nuclear Information System (INIS)

    Pazsit, Imre; Wihlstrand, Gustav; Tambouratzis, Tatiana; Jonsson, Anders; Dahl, Berit

    2009-12-01

    This report constitutes Stages 14 and 15 of a long-term research and development program concerning the development of diagnostics and monitoring methods for nuclear reactors. Stage 14 was a full one-year project, whereas Stage 15 consisted of a half-year project. The program executed in Stages 14 and 15 consists of the following three parts: - Study of criticality, neutron kinetics and neutron noise in molten salt reactors (MSR) (Stages 14 and 15); - An overview and introduction to fuzzy logics (Stage 14), and an application to two-phase flow identification (Stage 15) - Preparations for and execution of an IAEA-ICTP workshop on Neutron fluctuations, reactor noise and their applications in nuclear reactors (Stage 14)

  18. Research and Development Program in Reactor Diagnostics and Monitoring with Neutron Noise Methods. Stages 14 and 15

    Energy Technology Data Exchange (ETDEWEB)

    Pazsit, Imre; Wihlstrand, Gustav; Tambouratzis, Tatiana; Jonsson, Anders; Dahl, Berit (Chalmers Univ. of Technology, Dept. of Nuclear Engineering, SE-412 96 Goeteborg (Sweden))

    2009-12-15

    This report constitutes Stages 14 and 15 of a long-term research and development program concerning the development of diagnostics and monitoring methods for nuclear reactors. Stage 14 was a full one-year project, whereas Stage 15 consisted of a half-year project. The program executed in Stages 14 and 15 consists of the following three parts: - Study of criticality, neutron kinetics and neutron noise in molten salt reactors (MSR) (Stages 14 and 15); - An overview and introduction to fuzzy logics (Stage 14), and an application to two-phase flow identification (Stage 15) - Preparations for and execution of an IAEA-ICTP workshop on Neutron fluctuations, reactor noise and their applications in nuclear reactors (Stage 14)

  19. Flux distribution measurements in the Bruce B Unit 6 reactor using a transportable traveling flux detector system

    International Nuclear Information System (INIS)

    Leung, T.C.; Drewell, N.H.; Hall, D.S.; Lopez, A.M.

    1987-01-01

    A transportable traveling flux detector (TFD) system for use in power reactors has been developed and tested at Chalk River Nuclear Labs. in Canada. It consists of a miniature fission chamber, a motor drive mechanism, a computerized control unit, and a data acquisition subsystem. The TFD system was initially designed for the in situ calibration of fixed self-powered detectors in operating power reactors and for flux measurements to verify reactor physics calculations. However, this system can also be used as a general diagnostic tool for the investigation of apparent detector failures and flux anomalies and to determine the movement of reactor internal components. This paper describes the first successful use of the computerized TFD system in an operating Canada deuterium uranium (CANDU) power reactor and the results obtained from the flux distribution measurements. An attempt is made to correlate minima in the flux profile with the locations of fuel channels so that future measurements can be used to determine the sag of the channels. Twenty-seven in-core flux detector assemblies in the 855-MW (electric) Unit 6 reactor of the Ontario Hydro Bruce B Generating Station were scanned

  20. Vibration monitoring and diagnostic system developments in the VEIKI for the electric power industry

    International Nuclear Information System (INIS)

    Bessenyei, Z.

    1993-01-01

    Members of the ARGUS family comprising IBM PC-based data logging and data evaluation devices for vibration monitoring and diagnostics have been installed at WWER type nuclear reactors (in Paks, Hungary, and in Kalinin, Russia), and at some fossil fuel power plants in Hungary. The important features of the ARGUS monitoring system are discussed, and application details are presented. (N.T.) 2 figs

  1. Expert system for diagnostics and status monitoring of NPP water chemistry condition

    International Nuclear Information System (INIS)

    Shvedova, M.N.; Kritski, V.G.; Zakharova, S.V.; Benediktov, V.B.; Nikolaev, F.V.

    2002-01-01

    Water chemistry condition (WCC) has been the subject of constant study and improvement up to the present day. It is connected with the presence of a direct relationship between the violation of water chemistry regulation on the one hand and components reliability of the circuit's equipment and cost-effectiveness of their operation on the other. It dictates the necessity to apply different optimization methods in the field of monitoring and use of information analytical and diagnostic systems to assess WCC quality, control and support. LI ''VNIPIET'' employees have, for several years, been developing an expert diagnostic system for supporting WCC and status monitoring of RBMK - reactor NPPs [2]. This system has not only conveniently organized the traditional functions of information acquisition and storage, a complete presentation of information in the form of tables, graphs of a dynamical changes of parameters and formation regular reports, diagnostic functions and issuing recommendations on WCC correction, but it also allows the assessment of confidence in the diagnosis made, relying on a wide range of numerical estimates, which were calculated by the use of expert data, and to make a credible prediction of an existing situation development. (authors)

  2. 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.)

  3. Tendencies in operating power reactors

    International Nuclear Information System (INIS)

    Brinckmann, H.F.

    1987-01-01

    A survey is given about new tendencies in operating power reactors. In order to meet the high demands for control and monitoring of power reactors modern procedures are applicated such as the incore-neutron flux detection by means of electron emission detectors and multi-component activation probes, the noise diagnostics as well as high-efficient automation systems

  4. 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

  5. MFTF-B plasma-diagnostic system

    International Nuclear Information System (INIS)

    Throop, A.L.; Goerz, D.A.; Thomas, S.R.

    1981-01-01

    This paper describes the current design status of the plasma diagnostic system for MFTF-B. In this paper we describe the system requirement changes which have occurred as a result of the funded rescoping of the original MFTF facility into MFTF-B. We outline the diagnostic instruments which are currently planned, and present an overview of the diagnostic system

  6. Intercomparison of techniques for inspection and diagnostics of heavy water reactor pressure tubes. Determination of hydrogen concentration and blister characterization

    International Nuclear Information System (INIS)

    2009-03-01

    Heavy water reactors (HWRs) comprise significant numbers of today's operating nuclear power plants, and more are under construction. Efficient and accurate inspection and diagnostic techniques for various reactor components and systems, especially pressure tubes, are an important factor in ensuring reliable and safe plant operation. To foster international collaboration in the efficient and safe use of nuclear power, the IAEA conducted a Coordinated Research Project (CRP) on Intercomparison of Techniques for HWR Pressure Tube Inspection and Diagnostics. This CRP was carried out within the framework of the IAEA's Technical Working Group on Advanced Technologies for HWRs (the TWG-HWR). The TWG-HWR is a group of experts nominated by their governments and designated by the IAEA to provide advice and to support implementation of IAEA's project on advanced technologies for HWRs. The objective of the CRP was to compare non-destructive inspection and diagnostic techniques, in use and being developed, for structural integrity assessment of HWR pressure tubes. During the first phase of this CRP participants investigated the capability of different techniques to detect and characterize flaws. During the second phase participants collaborated to detect and characterize hydride blisters and to determine the hydrogen concentration in zirconium alloys. The intention was to identify the most effective pressure tube inspection and diagnostic methods and to identify further development needs. The organizations which participated in phase 2 of this CRP are: - Comision Nacional de Energia Atomica (CNEA), Argentina; - Atomic Energy of Canada Ltd. (AECL), Chalk River Laboratories (CRL), Canada; - Bhabha Atomic Research Centre (BARC), India; - Korea Atomic Energy Research Institute (KAERI), Republic of Korea; - National Institute for Research and Development for Technical Physics (NIRDTP), Romania; - Nuclear Non-Destructive Testing Research and Services (NNDT), Romania. IAEA-TECDOC-1499

  7. 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

  8. 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)

  9. Development of a rule-based diagnostic platform on an object-oriented expert system shell

    International Nuclear Information System (INIS)

    Wang, Wenlin; Yang, Ming; Seong, Poong Hyun

    2016-01-01

    Highlights: • Multilevel Flow Model represents system knowledge as a domain map in expert system. • Rule-based fault diagnostic expert system can identify root cause via a causal chain. • Rule-based fault diagnostic expert system can be used for fault simulation training. - Abstract: This paper presents the development and implementation of a real-time rule-based diagnostic platform. The knowledge is acquired from domain experts and textbooks and the design of the fault diagnosis expert system was performed in the following ways: (i) establishing of corresponding classes and instances to build the domain map, (ii) creating of generic fault models based on events, and (iii) building of diagnostic reasoning based on rules. Knowledge representation is a complicated issue of expert systems. One highlight of this paper is that the Multilevel Flow Model has been used to represent the knowledge, which composes the domain map within the expert system as well as providing a concise description of the system. The developed platform is illustrated using the pressure safety system of a pressurized water reactor as an example of the simulation test bed; the platform is developed using the commercial and industrially validated software G2. The emulation test was conducted and it has been proven that the fault diagnosis expert system can identify the faults correctly and in a timely way; this system can be used as a simulation-based training tool to assist operators to make better decisions.

  10. Basic concept of common reactor physics code systems. Final report of working party on common reactor physics code systems (CCS)

    International Nuclear Information System (INIS)

    2004-03-01

    A working party was organized for two years (2001-2002) on common reactor physics code systems under the Research Committee on Reactor Physics of JAERI. This final report is compilation of activity of the working party on common reactor physics code systems during two years. Objectives of the working party is to clarify basic concept of common reactor physics code systems to improve convenience of reactor physics code systems for reactor physics researchers in Japan on their various field of research and development activities. We have held four meetings during 2 years, investigated status of reactor physics code systems and innovative software technologies, and discussed basic concept of common reactor physics code systems. (author)

  11. Monitoring nuclear reactor systems using neural networks and fuzzy logic

    International Nuclear Information System (INIS)

    Ikonomopoulos, A.; Tsoukalas, L.H.; Uhrig, R.E.; Mullens, J.A.

    1992-01-01

    A new approach is presented that demonstrates the potential of trained artificial neural networks (ANNs) as generators of membership functions for the purpose of monitoring nuclear reactor systems. ANN's provide a complex-to-simple mapping of reactor parameters in a process analogous to that of measurement. Through such virtual measurements the value of parameters with operational significance, e.g., control-valve-disk-position, valve-line-up-or performance can be determined. In the methodology presented the output of virtual measuring device is a set of membership functions which independently represent different states of the system. Utilizing a fuzzy logic representation offers the advantage of describing the state of the system in a condensed form, developed through linguistic descriptions and convenient for application in monitoring, diagnostics and generally control algorithms. The developed methodology is applied to the problem of measuring the disk position of the secondary flow control is clearly demonstrated as well as a method for selecting the actual output. The results suggest that it is possible to construct virtual measuring devices through artificial neural networks mapping dynamic time series to a set of membership functions and thus enhance the capability of monitoring systems

  12. Molecular ecology of anaerobic reactor systems

    DEFF Research Database (Denmark)

    Hofman-Bang, H. Jacob Peider; Zheng, D.; Westermann, Peter

    2003-01-01

    Anaerobic reactor systems are essential for the treatment of solid and liquid wastes and constitute a core facility in many waste treatment plants. Although much is known about the basic metabolism in different types of anaerobic reactors, little is known about the microbes responsible for these ......Anaerobic reactor systems are essential for the treatment of solid and liquid wastes and constitute a core facility in many waste treatment plants. Although much is known about the basic metabolism in different types of anaerobic reactors, little is known about the microbes responsible...... to the abundance of each microbe in anaerobic reactor systems by rRNA probing. This chapter focuses on various molecular techniques employed and problems encountered when elucidating the microbial ecology of anaerobic reactor systems. Methods such as quantitative dot blot/fluorescence in-situ probing using various...

  13. 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

  14. Nuclear aspects of diagnostics in RTO/RC ITER

    International Nuclear Information System (INIS)

    Walker, C.I.; Yamamoto, S.; Costley, A.; Kock, L. de; Ebisawa, K.; Janeschitz, G.; Khripunov, V.; Martin, E.; Vayakis, G.

    2000-01-01

    ITER (international thermonuclear experimental reactor) will be the first fusion device where the design of the plasma diagnostic systems will make extensive use of the materials and techniques developed in the nuclear technology field. The designs have to satisfy stringent requirements for tritium confinement, nuclear shielding and vacuum integrity. This paper introduces the requirements for diagnostics in the ITER long pulse, burning plasma environment, and addresses the impact of the reactor environment on the diagnostics and ancillary equipment. These systems necessarily require access to the plasma or first wall, which generally conflicts with the requirements of the basic machine. Holes are required through the first wall, primary shielding, containment boundaries and biological shielding. Components have a limited life and require maintenance. This paper describes the effect of the radiation environment on diagnostic design at different locations. Ex-vessel and in-vessel remote handling, hot cell refurbishment and tritium confinement are also described

  15. Diagnostic technology of PWR plant equipment failures

    International Nuclear Information System (INIS)

    Nakamura, Tetsuo; Tanaka, Mamoru; Okamachi, Masao; Taguchi, Shozo; Nagashima, Kazuhiro; Ishikawa, Satoshi

    1985-01-01

    To confirm the soundness of the important facilities in a nuclear power plant contributes to the reliability of the plant operations and improvement of its operation rate. For this purpose, the following diagnostic techniques have been developed. (1). Vibration and loose parts monitoring: Detection of abnormal structural vibrations in the reactor, estimation of its mode, detection of loose parts in the primary system, and estimation of the position and energy of their collisions against the reactor vessel or the like. (2). Valve leak monitoring: Detection of leaks from primary valves in the primary cooling boundary, such as the pressurizer relief valve and safety valve, and estimation of the form of the leaks. (3). Detector noise response diagnosis: Diagnosis of degradation of principal process detectors during plant operation. Furthermore, a diagnostic system incorporating the above diagnostic technology applicable to actual plants has been experimentally manufactured and successfully verified. (author)

  16. 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

  17. 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

  18. 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

  19. 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)

  20. 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

  1. JAERI thermal reactor standard code system for reactor design and analysis SRAC

    International Nuclear Information System (INIS)

    Tsuchihashi, Keichiro

    1985-01-01

    SRAC, JAERI thermal reactor standard code system for reactor design and analysis, developed in Japan Atomic Energy Research Institute, is for all types of thermal neutron nuclear design and analysis. The code system has undergone extensive verifications to confirm its functions, and has been used in core modification of the research reactor, detailed design of the multi-purpose high temperature gas reactor and analysis of the experiment with a critical assembly. In nuclear calculation with the code system, multi-group lattice calculation is first made with the libraries. Then, with the resultant homogeneous equivalent group constants, reactor core calculation is made. Described are the following: purpose and development of the code system, functions of the SRAC system, bench mark tests and usage state and future development. (Mori, K.)

  2. ITER diagnostic system: Vacuum interface

    Energy Technology Data Exchange (ETDEWEB)

    Patel, K.M., E-mail: Kaushal.Patel@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-Lez-Durance (France); Udintsev, V.S.; Hughes, S.; Walker, C.I.; Andrew, P.; Barnsley, R.; Bertalot, L. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-Lez-Durance (France); Drevon, J.M. [Bertin Technologies, BP 22, 13762 Aix-en Provence cedex 3 (France); Encheva, A. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-Lez-Durance (France); Kashchuk, Y. [Institution “PROJECT CENTER ITER”, 1, Akademika Kurchatova pl., Moscow (Russian Federation); Maquet, Ph. [Bertin Technologies, BP 22, 13762 Aix-en Provence cedex 3 (France); Pearce, R.; Taylor, N.; Vayakis, G.; Walsh, M.J. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-Lez-Durance (France)

    2013-10-15

    Diagnostics play an essential role for the successful operation of the ITER tokamak. They provide the means to observe control and to measure plasma during the operation of ITER tokamak. The components of the diagnostic system in the ITER tokamak will be installed in the vacuum vessel, in the cryostat, in the upper, equatorial and divertor ports, in the divertor cassettes and racks, as well as in various buildings. Diagnostic components that are placed in a high radiation environment are expected to operate for the life of ITER. There are approx. 45 diagnostic systems located on ITER. Some diagnostics incorporate direct or independently pumped extensions to maintain their necessary vacuum conditions. They require a base pressure less than 10{sup −7} Pa, irrespective of plasma operation, and a leak rate of less than 10{sup −10} Pa m{sup 3} s{sup −1}. In all the cases it is essential to maintain the ITER closed fuel cycle. These directly coupled diagnostic systems are an integral part of the ITER vacuum containment and are therefore subject to the same design requirements for tritium and active gas confinement, for all normal and accidental conditions. All the diagnostics, whether or not pumped, incorporate penetration of the vacuum boundary (i.e. window assembly, vacuum feedthrough etc.) and demountable joints. Monitored guard volumes are provided for all elements of the vacuum boundary that are judged to be vulnerable by virtue of their construction, material, load specification etc. Standard arrangements are made for their construction and for the monitoring, evacuating and leak testing of these volumes. Diagnostic systems are incorporated at more than 20 ports on ITER. This paper will describe typical and particular arrangements of pumped diagnostic and monitored guard volume. The status of the diagnostic vacuum systems, which are at the start of their detailed design, will be outlined and the specific features of the vacuum systems in ports and extensions

  3. ITER diagnostic system: Vacuum interface

    International Nuclear Information System (INIS)

    Patel, K.M.; Udintsev, V.S.; Hughes, S.; Walker, C.I.; Andrew, P.; Barnsley, R.; Bertalot, L.; Drevon, J.M.; Encheva, A.; Kashchuk, Y.; Maquet, Ph.; Pearce, R.; Taylor, N.; Vayakis, G.; Walsh, M.J.

    2013-01-01

    Diagnostics play an essential role for the successful operation of the ITER tokamak. They provide the means to observe control and to measure plasma during the operation of ITER tokamak. The components of the diagnostic system in the ITER tokamak will be installed in the vacuum vessel, in the cryostat, in the upper, equatorial and divertor ports, in the divertor cassettes and racks, as well as in various buildings. Diagnostic components that are placed in a high radiation environment are expected to operate for the life of ITER. There are approx. 45 diagnostic systems located on ITER. Some diagnostics incorporate direct or independently pumped extensions to maintain their necessary vacuum conditions. They require a base pressure less than 10 −7 Pa, irrespective of plasma operation, and a leak rate of less than 10 −10 Pa m 3 s −1 . In all the cases it is essential to maintain the ITER closed fuel cycle. These directly coupled diagnostic systems are an integral part of the ITER vacuum containment and are therefore subject to the same design requirements for tritium and active gas confinement, for all normal and accidental conditions. All the diagnostics, whether or not pumped, incorporate penetration of the vacuum boundary (i.e. window assembly, vacuum feedthrough etc.) and demountable joints. Monitored guard volumes are provided for all elements of the vacuum boundary that are judged to be vulnerable by virtue of their construction, material, load specification etc. Standard arrangements are made for their construction and for the monitoring, evacuating and leak testing of these volumes. Diagnostic systems are incorporated at more than 20 ports on ITER. This paper will describe typical and particular arrangements of pumped diagnostic and monitored guard volume. The status of the diagnostic vacuum systems, which are at the start of their detailed design, will be outlined and the specific features of the vacuum systems in ports and extensions will be described

  4. Diagnostics and control for the steady state and pulsed tokamak DEMO

    Czech Academy of Sciences Publication Activity Database

    Orsitto, F.P.; Villari, R.; Moro, F.; Todd, T.N.; Lilley, S.; Jenkins, I.; Felton, R.; Biel, W.; Silva, A.; Scholz, M.; Rzadkiewicz, J.; Ďuran, Ivan; Tardocchi, M.; Gorini, G.; Morlock, C.; Federici, G.; Litnovsky, A.

    2016-01-01

    Roč. 56, č. 2 (2016), č. článku 026009. ISSN 0029-5515 Institutional support: RVO:61389021 Keywords : measurement systems, fusion reactor, fusion plasma diagnostics * fusion reactor * fusion plasma diagnostics * DEMO * Hall sensors * tokamak Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: Fluids and plasma physics (including surface physics) Impact factor: 3.307, year: 2016 http://iopscience.iop.org/article/10.1088/0029-5515/56/2/026009

  5. Pressurized water reactor monitoring. Study of detection, diagnostic and estimation methods (least error squares and filtering)

    International Nuclear Information System (INIS)

    Gillet, M.

    1986-07-01

    This thesis presents a study for the surveillance of the ''primary coolant circuit inventory monitoring'' of a pressurized water reactor. A reference model is developed in view of an automatic system ensuring detection and diagnostic in real time. The methods used for the present application are statistical tests and a method related to pattern recognition. The estimation of failures detected, difficult owing to the non-linearity of the problem, is treated by the least error squares method of the predictor or corrector type, and by filtering. It is in this frame that a new optimized method with superlinear convergence is developed, and that a segmented linearization of the model is introduced, in view of a multiple filtering [fr

  6. RSAS: a Reactor Safety Assessment System

    International Nuclear Information System (INIS)

    Sebo, D.E.; Dixon, B.W.; Bray, M.A.

    1985-01-01

    The Reactor Safety Assessment System (RSAS) is an expert system under development for the United States Nuclear Regulatory Commission (NRC). RSAS is being developed for use at the NRC's Operations Center in the event of a serious incident at a licensed nuclear power plant. The system generates situation assessments for the NRC Reactor Safety Team based on a limited number of plant parameters, known operator actions, and plant status data. The RSAS rule base currently covers one reactor type. The extension of the rule base to other reactor types is also discussed

  7. Small space reactor power systems for unmanned solar system exploration missions

    International Nuclear Information System (INIS)

    Bloomfield, H.S.

    1987-12-01

    A preliminary feasibility study of the application of small nuclear reactor space power systems to the Mariner Mark II Cassini spacecraft/mission was conducted. The purpose of the study was to identify and assess the technology and performance issues associated with the reactor power system/spacecraft/mission integration. The Cassini mission was selected because study of the Saturn system was identified as a high priority outer planet exploration objective. Reactor power systems applied to this mission were evaluated for two different uses. First, a very small 1 kWe reactor power system was used as an RTG replacement for the nominal spacecraft mission science payload power requirements while still retaining the spacecraft's usual bipropellant chemical propulsion system. The second use of reactor power involved the additional replacement of the chemical propulsion system with a small reactor power system and an electric propulsion system. The study also provides an examination of potential applications for the additional power available for scientific data collection. The reactor power system characteristics utilized in the study were based on a parametric mass model that was developed specifically for these low power applications. The model was generated following a neutronic safety and operational feasibility assessment of six small reactor concepts solicited from U.S. industry. This assessment provided the validation of reactor safety for all mission phases and generatad the reactor mass and dimensional data needed for the system mass model

  8. Diagnostic techniques for measuring temperature transients and stress transients in the first wall of an ICF reactor

    International Nuclear Information System (INIS)

    Melamed, N.T.; Taylor, L.H.

    1983-01-01

    The primary challenge in the design of an Inertial Confinement Fusion (ICF) power reactor is to make the first wall survive the frequent explosions of the pellets. Westinghouse has proposed a dry wall design consisting of steel tubes coated with tantalum. This report describes the design of a test chamber and two diagnostic procedures for experimentally determining the reliability of the Westinghouse design. The test chamber simulates the x-ray and ion pulse irradiation of the wall due to a pellet explosion. The diagnostics consist of remote temperature sensing and surface deformation measurements. The chamber and diagnostics can also be used to test other first-wall designs

  9. 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

  10. 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

  11. Acoustic and vibration monitoring systems and its place in the general conception of NPP diagnostics

    International Nuclear Information System (INIS)

    Dmitriev, V.; Kudryavtsev, B.; Morozov, S.; Finkel, B.

    1993-01-01

    Concept of diagnostics as a matter is an identification of defects and malfunctions in operating equipment by analyzing some data containing the information about these defects or malfunctions, but received without their direct observation. The direct observation of some parameters may be impossible in principle (for example, coolant saturation temperature), or because of inaccessibility for its measuring (for example, temperature of fuel elements), or because of the absence of a suitable sensor at a proper place. And, on the contrary, in some cases there is too much information for its simple interpretation. This report describes various monitoring systems of reactor components.Application is made to a pwr type reactor

  12. 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

  13. 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.

  14. 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

  15. 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.

  16. 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

  17. 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.

  18. Elaboration of the configuration and programming of the interlocks system of the TRIGA Mark III reactor

    International Nuclear Information System (INIS)

    Mejia C, M. A.

    2016-01-01

    The modernization of the TRIGA Mark III reactor interlock system requires a system that provides high reliability, flexibility and ease of operation during reactor operation. With this modernization of the system, is intended to prevent, control and mitigate the causes of probable accidents reported in the reactor accident analysis. On the other hand, is foreseen the ease reactor operation in a simple, safe and efficient way. The programmable logic controller can be programmed by programming instructions using simple language and easy to develop, these can be modified from a computer using the programming software. In addition, another of the advantages offered by the controller is that can be modified from a touch screen (human-machine interface) that allows adjustment, without the need to use programming software and diagnostic functions during the process. As a result of the present work, a situation of improvement in the reactor operation was generated, facilitating the handling of the bridge and increasing the efficiency of the system in the execution of the operating conditions of the installations external to the reactor. A modern, more reliable and much less expensive system was achieved than the previous one, avoiding that the maintenance to the system generates high expenses. With respect to the development of the application programming, a control was implemented that allows to select a zone of the five that have inside the pool to carry out the displacement of automatic way and later to be located in that zone, having in this way a greater efficiency and ease in bridge control. (Author)

  19. Nuclear reactor auxiliary heat removal system

    International Nuclear Information System (INIS)

    Thompson, R.E.; Pierce, B.L.

    1977-01-01

    An auxiliary heat removal system to remove residual heat from gas-cooled nuclear reactors is described. The reactor coolant is expanded through a turbine, cooled in a heat exchanger and compressed by a compressor before reentering the reactor coolant. The turbine powers both the compressor and the pump which pumps a second fluid through the heat exchanger to cool the reactor coolant. A pneumatic starter is utilized to start the turbine, thereby making the auxiliary heat removal system independent of external power sources

  20. Perspectives of expert systems in the field of reactor safety

    International Nuclear Information System (INIS)

    Bastl, W.

    1989-01-01

    The furure potential of expert systems is based on the following factors: the efficient storage of information in the knowledge basis, the efficient use of comprehensive information bases, the interactive approach, the rapid production of prototypes. The main problems which are encountered at present relate to the input of knowledge derived from experience, the qualification of the contents of the knowledge bases and to the interfacing to technical processes, as real time work is required in such cases. However, the practical use of expert systems in reactor safety is expected to make considerable progress. The following preferred fields should be mentioned: knowledge bases and analysis tools for safety investigations, diagnostic and practising systems for safe operation and, above all in the field of accident management, trainers, in-situ guiding systems or information systems in supraregional guiding centers. (orig./DG) [de

  1. Scaling-Up the Functional Diagnostic Systems

    International Nuclear Information System (INIS)

    Mohamed, A.H.

    2008-01-01

    Functional diagnostic systems received a lot of attention in the last decade. They have proven their powerful for diagnosis the new faults of some complex systems. But, they still have some complexity in both the representation and reasoning about the large-scale systems. This paper introduces a new functional diagnostic system that can divide its small functions into main and auxiliary ones. This process enables the diagnostic system to scale -up the representation of the tested system and simplify the diagnostic mechanism tasks. Thus, it can improve both the representation and reasoning complexity. Also,it can decrease the required analysis, cost, and time. Proposed system can be applied for a wide area of the large-scale systems. It has been proven its acceptance to be applied practically for the Complex real-time systems

  2. 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)

  3. Enhancing the functionality of reactor protection systems to provide diagnostic and monitoring information: The ISATTM approach

    International Nuclear Information System (INIS)

    Baldwin, J.A.; Rowe, B.J.; Jones, C.D.

    1996-01-01

    The ISAT TM architecture has been successfully implemented as the Single Channel Trip System (SCTS), part of the primary protection system of Nuclear Electric's Dungeness 'B' Advanced Gas-Cooled Reactors. The system is the first computer-based protection system licensed on a UK civil nuclear reactor. The system provides protection against single channel faults resulting in high coolant gas outlet temperature. The SCTS was designed to output data at several points in the system to an Ethernet to allow checks to be made on the operation of parts of the protection system and the system as a whole. In order to monitor the performance of this shutdown system a PC based monitoring system was developed to take input as data from the Ethernet, check its integrity and then analyze the data to provide information of the state of the system and subsystems. The SCTS monitor was basically intended to alert the operator to any fault on the safety system and indicate its source, provide a diagnosis of the cause of any trip initiated by the safety system, and log the occurrences of these incidents for later inspection. The intention was also to provide accurate real-time information on the thermocouple readings and to decrease the effort required to maintain the safety system. This paper will describe briefly the development of the ISAT TM monitoring system: how its requirements were arrived at, and how the design, code and testing were carried out to ensure approval for this application. It will then go on to report how the ISAT TM monitor has performed during its time in service: how more functionality has been added over and above its original requirements. Features of additional monitors for the SCTS and other ISAT TM systems will also be described. (author). 2 refs, 5 figs

  4. The TEX-I real-time expert system, applied to situation assessment for the SNR-300 reactor

    International Nuclear Information System (INIS)

    Schmal, N.; Leder, H.J.; Schade, H.J.

    1988-01-01

    Interatom, a subsidiary company of Siemens, is developing expert systems for the technical domain. These systems are operating in various industrial applications like flexible manufacturing or plant configuration, based on a domain-specific expert system shell, developed by Interatom. Additional projects are focusing on real-time diagnostics, e.g., for nuclear power plants. The authors report in this paper about a diagnosis expert system for the liquid-metal fast breeder reactor SNR-300, which uses new real-time tools, developed within the German TEX-I project (technical expert systems for data interpretation, diagnosis, and process control). The purpose of the system is to support the reactor operators in assessing plant status in real time, based on readings from many sensors. By on-line connection to the process control computer, it can monitor all incoming signal values, check the consistency of data, continuously diagnose the current plant status, detect unusual trends prior to accidents, localize faulty components, and recommended operator response in abnormal conditions. In the present knowledge acquisition and test phase, the expert system is connected to a real-time simulation of the reactor. The simulator is based on a thermohydraulic code for simulation of the transient behavior of temperatures and flow rates in the reactor core, plena, pipes, pumps, valves, intermediate heat exchangers, and cooling components. Additionally, the system's response to an asynchronous operator interaction can be simulated

  5. Plasma Diagnostics

    Energy Technology Data Exchange (ETDEWEB)

    Zaveryaev, V [Kurchatov Institute, Moscow (Russian Federation); others, and

    2012-09-15

    The success in achieving peaceful fusion power depends on the ability to control a high temperature plasma, which is an object with unique properties, possibly the most complicated object created by humans. Over years of fusion research a new branch of science has been created, namely plasma diagnostics, which involves knowledge of almost all fields of physics, from electromagnetism to nuclear physics, and up-to-date progress in engineering and technology (materials, electronics, mathematical methods of data treatment). Historically, work on controlled fusion started with pulsed systems and accordingly the methods of plasma parameter measurement were first developed for short lived and dense plasmas. Magnetically confined hot plasmas require the creation of special experimental techniques for diagnostics. The diagnostic set is the most scientifically intensive part of a plasma device. During many years of research operation some scientific tasks have been solved while new ones arose. New tasks often require significant changes in the diagnostic system, which is thus a very flexible part of plasma machines. Diagnostic systems are designed to solve several tasks. As an example here are the diagnostic tasks for the International Thermonuclear Experimental Reactor - ITER: (1) Measurements for machine protection and basic control; (2) Measurements for advanced control; (3) Additional measurements for performance evaluation and physics. Every new plasma machine is a further step along the path to the main goal - controlled fusion - and nobody knows in advance what new phenomena will be met on the way. So in the planning of diagnostic construction we should keep in mind further system upgrading to meet possible new scientific and technical challenges. (author)

  6. Nova target diagnostics control system

    International Nuclear Information System (INIS)

    Severyn, J.R.

    1985-01-01

    During the past year the Nova target diagnostics control system was finished and put in service. The diagnostics loft constructed to the north of the target room provides the environmental conditions required to collect reliable target diagnostic data. These improvements include equipment cooling and isolation of the power source with strict control of instrumentation grounds to eliminate data corruption due to electromagnetic pulses from the laser power-conditioning system or from target implosion effects

  7. Applying Diagnostics to Enhance Cable System Reliability (Cable Diagnostic Focused Initiative, Phase II)

    Energy Technology Data Exchange (ETDEWEB)

    Hartlein, Rick [Georgia Tech Research Corporation (GTRC), Atlanta, GA (United States). National Electric Energy Testing, Research and Applications Center (NEETRAC); Hampton, Nigel [Georgia Tech Research Corporation (GTRC), Atlanta, GA (United States). National Electric Energy Testing, Research and Applications Center (NEETRAC); Perkel, Josh [Georgia Tech Research Corporation (GTRC), Atlanta, GA (United States). National Electric Energy Testing, Research and Applications Center (NEETRAC); Hernandez, JC [Univ. de Los Andes, Merida (Venezuela); Elledge, Stacy [Georgia Tech Research Corporation (GTRC), Atlanta, GA (United States). National Electric Energy Testing, Research and Applications Center (NEETRAC); del Valle, Yamille [Georgia Tech Research Corporation (GTRC), Atlanta, GA (United States). National Electric Energy Testing, Research and Applications Center (NEETRAC); Grimaldo, Jose [Georgia Inst. of Technology, Atlanta, GA (United States). School of Electrical and Computer Engineering; Deku, Kodzo [Georgia Inst. of Technology, Atlanta, GA (United States). George W. Woodruff School of Mechanical Engineering

    2016-02-01

    The Cable Diagnostic Focused Initiative (CDFI) played a significant and powerful role in clarifying the concerns and understanding the benefits of performing diagnostic tests on underground power cable systems. This project focused on the medium and high voltage cable systems used in utility transmission and distribution (T&D) systems. While many of the analysis techniques and interpretations are applicable to diagnostics and cable systems outside of T&D, areas such as generating stations (nuclear, coal, wind, etc.) and other industrial environments were not the focus. Many large utilities in North America now deploy diagnostics or have changed their diagnostic testing approach as a result of this project. Previous to the CDFI, different diagnostic technology providers individually promoted their approach as the “the best” or “the only” means of detecting cable system defects.

  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. 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

  10. Mirror reactor surface study

    Energy Technology Data Exchange (ETDEWEB)

    Hunt, A. L.; Damm, C. C.; Futch, A. H.; Hiskes, J. R.; Meisenheimer, R. G.; Moir, R. W.; Simonen, T. C.; Stallard, B. W.; Taylor, C. E.

    1976-09-01

    A general survey is presented of surface-related phenomena associated with the following mirror reactor elements: plasma first wall, ion sources, neutral beams, director converters, vacuum systems, and plasma diagnostics. A discussion of surface phenomena in possible abnormal reactor operation is included. Several studies which appear to merit immediate attention and which are essential to the development of mirror reactors are abstracted from the list of recommended areas for surface work. The appendix contains a discussion of the fundamentals of particle/surface interactions. The interactions surveyed are backscattering, thermal desorption, sputtering, diffusion, particle ranges in solids, and surface spectroscopic methods. A bibliography lists references in a number of categories pertinent to mirror reactors. Several complete published and unpublished reports on surface aspects of current mirror plasma experiments and reactor developments are also included.

  11. Mirror reactor surface study

    International Nuclear Information System (INIS)

    Hunt, A.L.; Damm, C.C.; Futch, A.H.; Hiskes, J.R.; Meisenheimer, R.G.; Moir, R.W.; Simonen, T.C.; Stallard, B.W.; Taylor, C.E.

    1976-01-01

    A general survey is presented of surface-related phenomena associated with the following mirror reactor elements: plasma first wall, ion sources, neutral beams, director converters, vacuum systems, and plasma diagnostics. A discussion of surface phenomena in possible abnormal reactor operation is included. Several studies which appear to merit immediate attention and which are essential to the development of mirror reactors are abstracted from the list of recommended areas for surface work. The appendix contains a discussion of the fundamentals of particle/surface interactions. The interactions surveyed are backscattering, thermal desorption, sputtering, diffusion, particle ranges in solids, and surface spectroscopic methods. A bibliography lists references in a number of categories pertinent to mirror reactors. Several complete published and unpublished reports on surface aspects of current mirror plasma experiments and reactor developments are also included

  12. 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

  13. A modular multi-microcomputer system for on-line vibration diagnostics

    International Nuclear Information System (INIS)

    Saedtler, E.

    1988-01-01

    A new modular multi-microprocessor system for on-line vibration monitoring and diagnostics of PWRs is described. The aim of the system is to make feasible an early detection of increasing failures in relevant regions of a reactor plant, to verify the mechanical integrity of the investigated components, and to improve therefore the operational safety of the plant. After a discussion of the implemented surveillance methods and algorithms, which are based on hierarchical structured identification (estimation) and statistical pattern recognition tools, the system architecture (software and hardware) is portrayed. The classification scheme itself works sequential so that samples (or features) can arrive on-line. This on-line classification is important in order to take necessary actions in time. Furthermore, the system has learning capabilities, which means it is adaptable to different, varying states and plant conditions. The main features of the system are presented and its contribution to an automation of complex surveillance and monitoring tasks is shown. (author)

  14. Coolant clean-up system in nuclear reactor

    International Nuclear Information System (INIS)

    Tsuburaya, Hirobumi; Akita, Minoru; Shiraishi, Tadashi; Kinoshita, Shoichiro; Okura, Minoru; Tsuji, Akio.

    1987-01-01

    Purpose: To ensure a sufficient urging pressure at the inlet of a coolant clean-up system pump in a nuclear reactor and eliminate radioactive contaminations to the pump. Constitution: Coolant clean-up system (CUW) pump in a nuclear reactor is disposed to the downstream of a filtration desalter and, for compensating the insufficiency of the urging pressure at the pump inlet, the reactor water intake port to the clean-up system is disposed to the downstream of the after-heat removing pump and the heat exchanger. By compensating the net positive suction head (NPSH) of the clean-up system from the residual heat removing system, the problems of insufficient NPSH for the CUW pump upon reactor shut-down can be dissolved and, accordingly, the reactor clean-up system can be arranged in the order of the heat exchanger, clean-up device and pump. Thus, the CUW pump acts on reactor water after cleaned-up in the clean-up device to reduce the radioactivity contamination to the pump. (Kawakami, Y.)

  15. Studies on WWER core diagnostics

    International Nuclear Information System (INIS)

    Lunin, G.L.; Mitin, V.I.; Bulavin, V.V.

    1987-01-01

    The reliability and safety of nuclear power plants have decisive meaning under the situation that nuclear power generation steadily increases, and among various measures aiming at ensuring the reliability and safety in the operation of nuclear power plants, the countermeasures for protecting reactor core, main process equipment and high pressure circuits from damage have the important role, and the monitoring of condition and the organization of forecast, which are carried out continuously or periodically during the operation of nuclear power stations using the diagnostic expert system specially developed for the purpose, are included in them. Such monitoring enables the early detection of mechanical damage, increase of vibration, defects caused during operation and so on in reactor cores and primary and secondary circuits, and the continuous watching of defect developments. Also boiling in a core is detected, the place of abnormality occurrence is identified, and the intensity and characteristics of boiling are determined, thus the occurrence of dangerous condition is prevented. The developments of an in-core monitoring system and noise diagnostic systems are reported. (Kako, I.)

  16. 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.)

  17. High-Temperature Gas-cooled Reactor steam-cycle/cogeneration lead plant reactor vessel: system design description

    International Nuclear Information System (INIS)

    1983-01-01

    The Reactor Vessel System contains the primary coolant inventory within a gas-tight pressure boundary, and provides the necessary flow paths and overpressure protection for this pressure boundary. The Reactor Vessel System also houses the components of the Reactor System, the Heat Transport System, and the Auxiliary Heat Removal System. The scope of the Reactor Vessel System includes the prestressed concrete reactor vessel (PCRV) structure with its reinforcing steel and prestressing components; liners, penetrations, closures, and cooling water tubes attached to the concrete side of the liner; the thermal barrier (insulation) on the primary coolant side of the liner; instrumentation for structural monitoring; and a pressure relief system. Specifications are presented

  18. Some aspects of diagnostic systems perspective

    International Nuclear Information System (INIS)

    Korosec, D.

    1998-01-01

    The integrity and safety of all nuclear power plant systems and components is guaranteed by the high requirements to quality assurance during all phases of design, fabrication, construction and operation. Many of the countries operating nuclear facilities, introduced advanced, sophisticated diagnostic systems for continuous monitoring safety important process parameters. The licensee should perform an assessment of the existing diagnostic systems, often supplied by the original design, their reliability and the need for the introduction of the additional monitoring/diagnostic systems. The operating experience should be taken into account and the assessment of the further needs. On this field has to be made on the results of PSA studies. In addition to the cost benefit analysis the evaluation of the new diagnostic systems in the light of nuclear safety should be also made. Experience, gained from the utilities, which have already installed this kind of the equipment should be very useful. Introducing new diagnostic systems will require often a safety assessment of the necessary modifications. Licensing process should be based on the existing nuclear legislation with certain additional requirements. (author)

  19. Fully 3D printed integrated reactor array for point-of-care molecular diagnostics.

    Science.gov (United States)

    Kadimisetty, Karteek; Song, Jinzhao; Doto, Aoife M; Hwang, Young; Peng, Jing; Mauk, Michael G; Bushman, Frederic D; Gross, Robert; Jarvis, Joseph N; Liu, Changchun

    2018-06-30

    Molecular diagnostics that involve nucleic acid amplification tests (NAATs) are crucial for prevention and treatment of infectious diseases. In this study, we developed a simple, inexpensive, disposable, fully 3D printed microfluidic reactor array that is capable of carrying out extraction, concentration and isothermal amplification of nucleic acids in variety of body fluids. The method allows rapid molecular diagnostic tests for infectious diseases at point of care. A simple leak-proof polymerization strategy was developed to integrate flow-through nucleic acid isolation membranes into microfluidic devices, yielding a multifunctional diagnostic platform. Static coating technology was adopted to improve the biocompatibility of our 3D printed device. We demonstrated the suitability of our device for both end-point colorimetric qualitative detection and real-time fluorescence quantitative detection. We applied our diagnostic device to detection of Plasmodium falciparum in plasma samples and Neisseria meningitides in cerebrospinal fluid (CSF) samples by loop-mediated, isothermal amplification (LAMP) within 50 min. The detection limits were 100 fg for P. falciparum and 50 colony-forming unit (CFU) for N. meningitidis per reaction, which are comparable to that of benchtop instruments. This rapid and inexpensive 3D printed device has great potential for point-of-care molecular diagnosis of infectious disease in resource-limited settings. Copyright © 2018 Elsevier B.V. All rights reserved.

  20. 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

  1. Space reactor electric systems: system integration studies, Phase 1 report

    International Nuclear Information System (INIS)

    Anderson, R.V.; Bost, D.; Determan, W.R.; Harty, R.B.; Katz, B.; Keshishian, V.; Lillie, A.F.; Thomson, W.B.

    1983-01-01

    This report presents the results of preliminary space reactor electric system integration studies performed by Rockwell International's Energy Systems Group (ESG). The preliminary studies investigated a broad range of reactor electric system concepts for powers of 25 and 100 KWe. The purpose of the studies was to provide timely system information of suitable accuracy to support ongoing mission planning activities. The preliminary system studies were performed by assembling the five different subsystems that are used in a system: the reactor, the shielding, the primary heat transport, the power conversion-processing, and the heat rejection subsystems. The subsystem data in this report were largely based on Rockwell's recently prepared Subsystem Technology Assessment Report. Nine generic types of reactor subsystems were used in these system studies. Several levels of technology were used for each type of reactor subsystem. Seven generic types of power conversion-processing subsystems were used, and several levels of technology were again used for each type. In addition, various types and levels of technology were used for the shielding, primary heat transport, and heat rejection subsystems. A total of 60 systems were studied

  2. 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

  3. 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)

  4. 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)

  5. Investigations with diagnostic fuel rod bundles on Rheinsberg NPP

    International Nuclear Information System (INIS)

    Krauze, F.; Rudolf, G.; Shajfler, V.; Tsimke, K.

    1982-01-01

    In 70MW pressurized water reactor of Rheinsberg NPP diagnostic fuel rod bundles have been installed: first of DK 1 type and then of DK 2 advanced type. Three rounds of measurement were run with DK 1 bundle and one with DK 2. The diagnostic bundles are equiped with various sensors for temperature, pressure, neutron flux and mechanical stress measurements as well as with special flow rate control system which allows to reach coolant boiling within the bundle. Qualitative and quantitative description of the sensors performance during reactor operation is given. The presented experimental results are connected with: 1) working capability of the measuring devices and their calibration; 2) throttling and boiling in two regimes: a) stationary and non-stationary flow rate throbgh DK during stationary reactor operation; b) various constant levels of flow rate through DK during non-stationary reactor operation regime [ru

  6. Nuclear reactor safety program in US department of energy and future perspectives

    International Nuclear Information System (INIS)

    Song, Y.T.

    1988-01-01

    The US Department of Energy (DOE) establishes policy, issues orders, and assures compliance with requirements. The contractors who design, construct, modify, operate, maintain and decommission DOE reactors, set forth the assessment of the safety of cognizant reactors and implement DOE orders. Teams of experts in the Department, through scheduled and unscheduled review programs, reassess the safety of reactors in every phases of their lives. As new technology develops, the safety programs are reevaluated and policies are modified to accommodate these new technologies. The diagnostic capabilities of the computer using multiple alarms to enhance detection of defects and control of a reactor have been greatly utilized in reactor operating systems. The Application of artificial intelligence technologies for diagnostic and even for the decision making process in the event of reactor accidents would be one of the future trends in reactor safety programs

  7. Thioaptamer Diagnostic System, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — AM Biotechnologies (AM) will develop a diagnostic system in response to SBIR Topic X10.01 Reusable Diagnostic Lab Technology that will simultaneously detect and...

  8. Building reactor operator sustain expert system with C language integrated production system

    International Nuclear Information System (INIS)

    Ouyang Qin; Hu Shouyin; Wang Ruipian

    2002-01-01

    The development of the reactor operator sustain expert system is introduced, the capability of building reactor operator sustain expert system is discussed with C Language Integrated Production System (Clips), and a simple antitype of expert system is illustrated. The limitation of building reactor operator sustain expert system with Clips is also discussed

  9. The unique safety challenges of space reactor systems

    International Nuclear Information System (INIS)

    Lanes, S.J.; Marshall, A.C.

    1991-01-01

    Compact reactor systems can provide high levels of power for extended periods in space environments. Their relatively low mass and their ability to operate independently of their proximity to the sun make reactor power systems high desirable for many civilian and military space missions. The US Department of Energy is developing reactor system technologies to provide electrical power for space applications. In addition, reactors are now being considered to provide thermal power to a hydrogen propellant for nuclear thermal rocketry. Space reactor safety issues differ from commercial reactor issues, in some areas, because of very different operating requirements and environments. Accidents similar to those postulated for commercial reactors must be considered for space reactors during their operational phase. Safety strategies will need to be established that account for the consequences of the loss of essential power

  10. Probabilistic techniques using Monte Carlo sampling for multi- component system diagnostics

    International Nuclear Information System (INIS)

    Aumeier, S.E.; Lee, J.C.; Akcasu, A.Z.

    1995-01-01

    We outline the structure of a new approach at multi-component system fault diagnostics which utilizes detailed system simulation models, uncertain system observation data, statistical knowledge of system parameters, expert opinion, and component reliability data in an effort to identify incipient component performance degradations of arbitrary number and magnitude. The technique involves the use of multiple adaptive Kalman filters for fault estimation, the results of which are screened using standard hypothesis testing procedures to define a set of component events that could have transpired. Latin Hypercube sample each of these feasible component events in terms of uncertain component reliability data and filter estimates. The capabilities of the procedure are demonstrated through the analysis of a simulated small magnitude binary component fault in a boiling water reactor balance of plant. The results show that the procedure has the potential to be a very effective tool for incipient component fault diagnosis

  11. 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

  12. Gas-cooled reactor for space power systems

    International Nuclear Information System (INIS)

    Walter, C.E.; Pearson, J.S.

    1987-05-01

    Reactor characteristics based on extensive development work on the 500-MWt reactor for the Pluto nuclear ramjet are described for space power systems useful in the range of 2 to 20 MWe for operating times of 1 y. The modest pressure drop through the prismatic ceramic core is supported at the outlet end by a ceramic dome which also serves as a neutron reflector. Three core materials are considered which are useful at temperatures up to about 2000 K. Most of the calculations are based on a beryllium oxide with uranium dioxide core. Reactor control is accomplished by use of a burnable poison, a variable-leakage reflector, and internal control rods. Reactivity swings of 20% are obtained with a dozen internal boron-10 rods for the size cores studied. Criticality calculations were performed using the ALICE Monte Carlo code. The inherent high-temperature capability of the reactor design removes the reactor as a limiting condition on system performance. The low fuel inventories required, particularly for beryllium oxide reactors, make space power systems based on gas-cooled near-thermal reactors a lesser safeguard risk than those based on fast reactors

  13. 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.)

  14. 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

  15. Safety analysis of reactor's cooling system

    International Nuclear Information System (INIS)

    1999-01-01

    Results of the analysis of reactor's RBMK-1500 coolant system during normal operation mode, hydrodynamic testing and in the case of earthquake are presented. Analysis was performed using RELAP5 code. Calculations showed the most vulnerable place in the reactor's coolant system. It was found that in the case of earthquake the horizontal support system of drum separator could be damaged

  16. Gaseous fuel reactors for power systems

    Science.gov (United States)

    Kendall, J. S.; Rodgers, R. J.

    1977-01-01

    Gaseous-fuel nuclear reactors have significant advantages as energy sources for closed-cycle power systems. The advantages arise from the removal of temperature limits associated with conventional reactor fuel elements, the wide variety of methods of extracting energy from fissioning gases, and inherent low fissile and fission product in-core inventory due to continuous fuel reprocessing. Example power cycles and their general performance characteristics are discussed. Efficiencies of gaseous fuel reactor systems are shown to be high with resulting minimal environmental effects. A technical overview of the NASA-funded research program in gaseous fuel reactors is described and results of recent tests of uranium hexafluoride (UF6)-fueled critical assemblies are presented.

  17. Elaboration of the configuration and programming of the interlocks system of the TRIGA Mark III reactor; Elaboracion de la configuracion y programacion del sistema de interbloqueos del reactor Triga Mark III

    Energy Technology Data Exchange (ETDEWEB)

    Mejia C, M. A.

    2016-07-01

    The modernization of the TRIGA Mark III reactor interlock system requires a system that provides high reliability, flexibility and ease of operation during reactor operation. With this modernization of the system, is intended to prevent, control and mitigate the causes of probable accidents reported in the reactor accident analysis. On the other hand, is foreseen the ease reactor operation in a simple, safe and efficient way. The programmable logic controller can be programmed by programming instructions using simple language and easy to develop, these can be modified from a computer using the programming software. In addition, another of the advantages offered by the controller is that can be modified from a touch screen (human-machine interface) that allows adjustment, without the need to use programming software and diagnostic functions during the process. As a result of the present work, a situation of improvement in the reactor operation was generated, facilitating the handling of the bridge and increasing the efficiency of the system in the execution of the operating conditions of the installations external to the reactor. A modern, more reliable and much less expensive system was achieved than the previous one, avoiding that the maintenance to the system generates high expenses. With respect to the development of the application programming, a control was implemented that allows to select a zone of the five that have inside the pool to carry out the displacement of automatic way and later to be located in that zone, having in this way a greater efficiency and ease in bridge control. (Author)

  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. 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)

  20. 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

  1. 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.

  2. Studying the effects of operators' problem solving behaviour when using a diagnostic expert system developed for the nuclear industry

    International Nuclear Information System (INIS)

    Holmstroem, C.B.O.; Volden, F.S.; Endestad, T.

    1992-01-01

    This paper describes an experiment with the purpose to also illustrate and discuss some of the methodological problems when empirically studying problem solving. The experiment which was the second in a series, conducted at the OECD Halden Reactor Project, aimed to assess the effect on nuclear power plant operators diagnostic behaviour when using a rule-based diagnostic expert system. The rule-based expert system used in the experiment is called DISKET (Diagnosis System Using Knowledge Engineering Technique) and was originally developed by the Japan Atomic Energy Research Institute (JAERI). The experiment was performed in the Halden man-machine laboratory using a full scope pressurized water reactor simulator. Existing data collection methods and experimental design principles includes possibilities but also limitations. This is discussed and experiences are presented. Operator performance in terms of quality of diagnosis is improved by the use of DISKET. The use of the DISKET system also influences operators problem solving behaviour. The main difference between the two experimental conditions can be characterized as while the DISKET users during the diagnosis process are following a strategy which is direct and narrowed, the non-DISKET users are using a much broader and less focused search when trying to diagnose a disturbance. (author)

  3. 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

  4. Case-Based Fault Diagnostic System

    International Nuclear Information System (INIS)

    Mohamed, A.H.

    2014-01-01

    Nowadays, case-based fault diagnostic (CBFD) systems have become important and widely applied problem solving technologies. They are based on the assumption that “similar faults have similar diagnosis”. On the other hand, CBFD systems still suffer from some limitations. Common ones of them are: (1) failure of CBFD to have the needed diagnosis for the new faults that have no similar cases in the case library. (2) Limited memorization when increasing the number of stored cases in the library. The proposed research introduces incorporating the neural network into the case based system to enable the system to diagnose all the faults. Neural networks have proved their success in the classification and diagnosis problems. The suggested system uses the neural network to diagnose the new faults (cases) that cannot be diagnosed by the traditional CBR diagnostic system. Besides, the proposed system can use the another neural network to control adding and deleting the cases in the library to manage the size of the cases in the case library. However, the suggested system has improved the performance of the case based fault diagnostic system when applied for the motor rolling bearing as a case of study

  5. Flexible Conversion Ratio Fast Reactor Systems Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Neil Todreas; Pavel Hejzlar

    2008-06-30

    Conceptual designs of lead-cooled and liquid salt-cooled fast flexible conversion ratio reactors were developed. Both concepts have cores reated at 2400 MWt placed in a large-pool-type vessel with dual-free level, which also contains four intermediate heat exchanges coupling a primary coolant to a compact and efficient supercritical CO2 Brayton cycle power conversion system. Decay heat is removed passively using an enhanced Reactor Vessel Auxiliary Cooling System and a Passive Secondary Auxiliary Cooling System. The most important findings were that (1) it is feasible to design the lead-cooled and salt-cooled reactor with the flexible conversion ratio (CR) in the range of CR=0 and CR=1 n a manner that achieves inherent reactor shutdown in unprotected accidents, (2) the salt-cooled reactor requires Lithium thermal Expansion Modules to overcme the inherent salt coolant's large positive coolant temperature reactivity coefficient, (3) the preferable salt for fast spectrum high power density cores is NaCl-Kcl-MgCl2 as opposed to fluoride salts due to its better themal-hydraulic and neutronic characteristics, and (4) both reactor, but attain power density 3 times smaller than that of the sodium-cooled reactor.

  6. Flexible Conversion Ratio Fast Reactor Systems Evaluation

    International Nuclear Information System (INIS)

    Neil Todreas; Pavel Hejzlar

    2008-01-01

    Conceptual designs of lead-cooled and liquid salt-cooled fast flexible conversion ratio reactors were developed. Both concepts have cores treated at 2400 MWt placed in a large-pool-type vessel with dual-free level, which also contains four intermediate heat exchanges coupling a primary coolant to a compact and efficient supercritical CO2 Brayton cycle power conversion system. Decay heat is removed passively using an enhanced Reactor Vessel Auxiliary Cooling System and a Passive Secondary Auxiliary Cooling System. The most important findings were that (1) it is feasible to design the lead-cooled and salt-cooled reactor with the flexible conversion ratio (CR) in the range of CR=0 and CR=1 n a manner that achieves inherent reactor shutdown in unprotected accidents, (2) the salt-cooled reactor requires Lithium thermal Expansion Modules to overcome the inherent salt coolant's large positive coolant temperature reactivity coefficient, (3) the preferable salt for fast spectrum high power density cores is NaCl-Kcl-MgCl2 as opposed to fluoride salts due to its better thermal-hydraulic and neutronic characteristics, and (4) both reactor, but attain power density 3 times smaller than that of the sodium-cooled reactor

  7. 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)

  8. Application of expert system to evaluating reactor water cleanup system performance

    International Nuclear Information System (INIS)

    Maeda, Katsuji; Nakamura, Masahiro; Nagasawa, Katsumi; Fushiki, Sumiyuki.

    1991-01-01

    Expert systems employing artificial intelligence (AI) have been developed for finding and elucidating causes of anomalies and malfunctions, presenting pertinent recommendation for countermeasures and for making precautionary diagnosis. On the other hand, further improvements in reliabilities for chemical control are required to promote BWR plant reliability and advancement. Especially, it is necessary to maintain the reactor water purity in high quality to minimize stress corrosion cracking (SCC) in primary cooling system, fuel performance degradation and radiation buildup. The reactor water quality is controlled by the reactor water cleanup (RWCU) system. So, it is very important to maintain the RWCU performance, in order to keep good reactor water quality. This paper describes an expert system used for evaluating RWCU system performance in BWR plants. (author)

  9. Integral reactor system and method for fuel cells

    Science.gov (United States)

    Fernandes, Neil Edward; Brown, Michael S; Cheekatamarla, Praveen; Deng, Thomas; Dimitrakopoulos, James; Litka, Anthony F

    2013-11-19

    A reactor system is integrated internally within an anode-side cavity of a fuel cell. The reactor system is configured to convert hydrocarbons to smaller species while mitigating the lower production of solid carbon. The reactor system may incorporate one or more of a pre-reforming section, an anode exhaust gas recirculation device, and a reforming section.

  10. 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

  11. 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

  12. 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)

  13. Psychometric perspectives on diagnostic systems

    NARCIS (Netherlands)

    Borsboom, D.

    2008-01-01

    The author identifies four conceptualizations of the relation between symptoms and disorders as utilized in diagnostic systems such as the Diagnostic and Statistical Manual of Mental Disorders, Fourth Edition (DSM-IV; American Psychiatric Association, 1994): A constructivist perspective, which holds

  14. A new diagnostic concept for the primary circuit of WWER 1000 reactors

    International Nuclear Information System (INIS)

    Streicher, V.; Liska, J.

    1993-01-01

    The new concept developed by the Skoda and Siemens companies is based on their own experience, the requirements of the Czech Power Board/NPP Temelin Diagnostic Department, and on the research work of various institutes in Czechoslovakia. The development of a complex diagnostic concept is an iterative process and includes parts with different stages of experience and different goals. They can be divided into five groups: PC-based diagnostic systems with continuous or periodic functions, equipment for condition monitoring, PC-based data collectors, studies, analyses and measures which are mandatory for the correct performance and interpretation of the diagnostic systems, and the integration of all subsystems and data acquisition/evaluation units into a Local Area Network with a graphic workstation. (Z.S.) 2 figs., 4 refs

  15. 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)

  16. Nuclear Reactor RA Safety Report, Vol. 8, Auxiliary system

    International Nuclear Information System (INIS)

    1986-11-01

    This volume describes RA reactor auxiliary systems, as follows: special ventilation system, special drainage system, hot cells, systems for internal transport. Ventilation system is considered as part of the reactor safety and protection system. Its role is eliminate possible radioactive particles dispersion in the environment. Special drainage system includes pipes and reservoirs with the safety role, meaning absorption or storage of possible radioactive waste water from the reactor building. Hot cells existing in the RA reactor building are designed for production of sealed radioactive sources, including packaging and transport [sr

  17. Reactor power system deployment and startup

    International Nuclear Information System (INIS)

    Wetch, J.R.; Nelin, C.J.; Britt, E.J.; Klein, G.; Rasor Associates, Inc., Sunnyvale, CA; California Institute of Technology, Pasadena)

    1985-01-01

    This paper addresses issues that should receive further examination in the near-term as concept selection for development of a U.S. space reactor power system is approached. The issues include: the economics, practicality and system reliability associated with transfer of nuclear spacecraft from low earth shuttle orbits to operational orbits, via chemical propulsion versus nuclear electric propulsion; possible astronaut supervised reactor and nuclear electric propulsion startup in low altitude Shuttle orbit; potential deployment methods for nuclear powered spacecraft from Shuttle; the general public safety of low altitude startup and nuclear safe and disposal orbits; the question of preferred reactor power level; and the question of frozen versus molten alkali metal coolant during launch and deployment. These issues must be considered now because they impact the SP-100 concept selection, power level selection, weight and size limits, use of deployable radiators, reliability requirements, and economics, as well as the degree of need for and the urgency of developing space reactor power systems. 5 references

  18. Gaseous fuel reactors for power systems

    International Nuclear Information System (INIS)

    Helmick, H.H.; Schwenk, F.C.

    1978-01-01

    The Los Alamos Scientific Laboratory is participating in a NASA-sponsored program to demonstrate the feasibility of a gaseous uranium fueled reactor. The work is aimed at acquiring experimental and theoretical information for the design of a prototype plasma core reactor which will test heat removal by optical radiation. The basic goal of this work is for space applications, however, other NASA-sponsored work suggests several attractive applications to help meet earth-bound energy needs. Such potential benefits are small critical mass, on-site fuel processing, high fuel burnup, low fission fragment inventory in reactor core, high temperature for process heat, optical radiation for photochemistry and space power transmission, and high temperature for advanced propulsion systems. Low power reactor experiments using uranium hexafluoride gas as fuel demonstrated performance in accordance with reactor physics predictions. The final phase of experimental activity now in progress is the fabrication and testing of a buffer gas vortex confinement system

  19. 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)

  20. 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

  1. An approach to build a knowledge base for reactor diagnostic system using statistical method

    International Nuclear Information System (INIS)

    Yokobayashi, Masao; Matsumoto, Kiyoshi; Kohsaka, Atsuo

    1988-01-01

    In the development of a rule-based expert system, one of the key issues is how to acquire knowledge and to build a knowledge base. When the knowledge base of DISKET was built, which is an expert system for nuclear reactor accident diagnosis developed in Japan Atomic Energy Research Institute, several problems have been experienced. To write rules is a time-consuming task, and it was difficult to keep the objectivity and consistency of rules as the number of rules increased. Certainty factors must be determined often according to engineering judgement, i.e. empirically or intuitively. A systematic approach was attempted to cope with these difficulties and to build efficiently an objective knowledge base. The approach described in this paper is based on the concept that a prototype knowledge base, colloquially speaking an initial guess, should first be generated in a systematic way, then it is modified or improved by human experts for practical use. Factor analysis was used as the systematic way. DISKET system, the procedure of building a knowledge base, and the verification of the approach are reported. (Kako, I.)

  2. Reactor Shutdown Mechanism by Top-mounted Hydraulic System

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sang Haun; Cho, Yeong Garp; Choi, Myoung Hwan; Lee, Jin Haeng; Huh, Hyung; Kim, Jong In [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-05-15

    There are two types of reactor shutdown mechanisms in HANARO. One is the mechanism driven by a hydraulic system, and the other is driven by a stepping motor. In HANARO, there are four Control Rod Drive Mechanisms (CRDMs) with an individual step motor and four Shutoff (SO) Units with an individual hydraulic system located at the top of reactor pool. The absorber rods in SO units are poised at the top of the core by the hydraulic force during normal operation. The rods of SO units drop by gravity as the first reactor showdown mechanism when a trip is commended by the reactor protection system (RPS). The rods in CRDMs also drop by gravity together as a redundant shutdown mechanism. When a trip is commended by the reactor regulating system (RRS), the absorber rods of CRDM only drop; while the absorber rods of SO units stay at the top of the core by the hydraulic system. The reactivity control mechanisms of in JRTR, one of the new research reactor with plate type fuels, consist of four CRDMs driven by an individual step motor and two second shutdown drive mechanisms (SSDMs) driven by an individual hydraulic system as shown in Fig. 1. The CRDMs act as the first reactor shutdown mechanism and reactor regulating as well. The top-mounted SSDM driven by the hydraulic system for the JRTR is under design in KAERI. The SSDM provides an alternate and independent means of reactor shutdown. The second shutdown rods (SSRs) of the SSDM are poised at the top of the core by the hydraulic system during the normal operation and drop by gravity for the reactor trip. Based on the proven technology of the design, operation and maintenance for HANARO, the SSDM for the JRTR has been optimized by the design improvement from the experience and test. This paper aims for the introduction of the SSDM in the process of the basic design. The major differences of the shutdown mechanisms by the hydraulic system are compared between HANARO and JRTR, and the design features, system, structure and

  3. 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)

  4. Experimental Development and Demonstration of Ultrasonic Measurement Diagnostics for Sodium Fast Reactor Thermal-hydraulics

    Energy Technology Data Exchange (ETDEWEB)

    Tokuhiro, Akira; Jones, Byron

    2013-09-13

    This research project will address some of the principal technology issues related to sodium-cooled fast reactors (SFR), primarily the development and demonstration of ultrasonic measurement diagnostics linked to effective thermal convective sensing under normatl and off-normal conditions. Sodium is well-suited as a heat transfer medium for the SFR. However, because it is chemically reactive and optically opaque, it presents engineering accessibility constraints relative to operations and maintenance (O&M) and in-service inspection (ISI) technologies that are currently used for light water reactors. Thus, there are limited sensing options for conducting thermohydraulic measurements under normal conditions and off-normal events (maintenance, unanticipated events). Acoustic methods, primarily ultrasonics, are a key measurement technology with applications in non-destructive testing, component imaging, thermometry, and velocimetry. THis project would have yielded a better quantitative and qualitative understanding of the thermohydraulic condition of solium under varied flow conditions. THe scope of work will evaluate and demonstrate ultrasonic technologies and define instrumentation options for the SFR.

  5. Reactor protection system design using micro-computers

    International Nuclear Information System (INIS)

    Fairbrother, D.B.

    1977-01-01

    Reactor Protection Systems for Nuclear Power Plants have traditionally been built using analog hardware. This hardware works quite well for single parameter trip functions; however, optimum protection against DNBR and KW/ft limits requires more complex trip functions than can easily be handled with analog hardware. For this reason, Babcock and Wilcox has introduced a Reactor Protection System, called the RPS-II, that utilizes a micro-computer to handle the more complex trip functions. This paper describes the design of the RPS-II and the operation of the micro-computer within the Reactor Protection System

  6. Reactor protection system design using micro-computers

    International Nuclear Information System (INIS)

    Fairbrother, D.B.

    1976-01-01

    Reactor protection systems for nuclear power plants have traditionally been built using analog hardware. This hardware works quite well for single parameter trip functions; however, optimum protection against DNBR and KW/ft limits requires more complex trip functions than can easily be handled with analog hardware. For this reason, Babcock and Wilcox has introduced a Reactor Protection System, called the RPS-II, that utilizes a micro-computer to handle the more complex trip functions. The paper describes the design of the RPS-II and the operation of the micro-computer within the Reactor Protection System

  7. Seismic attenuation system for a nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Liszkai, Tamas; Cadell, Seth

    2018-01-30

    A system for attenuating seismic forces includes a reactor pressure vessel containing nuclear fuel and a containment vessel that houses the reactor pressure vessel. Both the reactor pressure vessel and the containment vessel include a bottom head. Additionally, the system includes a base support to contact a support surface on which the containment vessel is positioned in a substantially vertical orientation. An attenuation device is located between the bottom head of the reactor pressure vessel and the bottom head of the containment vessel. Seismic forces that travel from the base support to the reactor pressure vessel via the containment vessel are attenuated by the attenuation device in a direction that is substantially lateral to the vertical orientation of the containment vessel.

  8. Nuclear reactor safety program in U.S. Department of Energy and future perspectives

    International Nuclear Information System (INIS)

    Song, Y.T.

    1987-01-01

    The U.S. Department of Energy (DOE) establishes policy, issues orders, and assures compliance with requirements. The contractors who design, construct, modify, operate, maintain and decommission DOE reactors, set forth the assessment of the safety of cognizant reactors and impliment DOE orders. Teams of experts in the Depatment, through scheduled and unscheduled review programs, reassess the safety of reactors in every phases of their lives. As new technology develops, the safety programs are reevaluated and policies are modified to accommodate these new technologies. The diagnostic capabilities of the computer using multiple alarms to enhance detection of defects and control of a reactor have been greatly utilized in reactor operating systems. The application of artificial intelligence (AI) technologies for diagnostic and even for the decision making process in the event of reactor accidents would be one of the future trends in reactor safety programs. (author)

  9. Expert System for Diagnostics and Status Monitoring of NPP Water Chemistry Condition

    International Nuclear Information System (INIS)

    Shvedova, M.N.; Kritski, V.G.; Zakharova, S.V.; Nikolaev, F.V.; Benediktov, V.B.

    2002-01-01

    Water chemistry condition (WCC) has been the subject of constant study and improvement up to the present day. It is connected with the presence of a direct relationship between the violation of water chemistry regulation on the one hand and components reliability of the circuit's equipment and cost-effectiveness of their operation on the other. It dictates the necessity to apply different optimization methods in the field of monitoring and use of information - analytical and diagnostic systems to assess WCC quality, control and support. By now NPP experts have broad experience in revealing and removing the causes of WCC disturbances. However this knowledge is often of an intuitive, non-classified nature, scattered among various working documents, which makes their transfer difficult. Based on what has been mentioned above, special attention is currently being paid to the problem of creating expert diagnostic systems for supporting the optimum WCC. The existing developments in this field (DIWA, Smart chem Works, the water quality control system at the Onagava NPP etc. [1,3,4,5] are based on wide use of experts' knowledge. Such expert diagnostic systems for supporting WCC refer to the new generation of intellectual control methods, which allow the incorporation of the latest achievements both in the field of water chemistry simulation and in the field of artificial intelligence and computer technologies. LI 'VNIPIET' employees have, for several years, been developing an expert diagnostic system for supporting WCC and status monitoring of RBMK - reactor NPPs [2]. This system has not only conveniently organized the traditional functions of information acquisition and storage, a complete presentation of information in the form of tables, graphs of a dynamical changes of parameters and formation regular reports, diagnostic functions and issuing recommendations on WCC correction, but it also allows the assessment of confidence in the diagnosis made, relying on a wide

  10. Study on secondary shutdown systems in Tehran research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jalali, H.R.; Fadaei, A.H., E-mail: Fadaei_amir@aut.ac.ir; Gharib, M.

    2015-09-15

    Highlights: • A study was undertaken to summarize the techniques for secondary shutdown systems (SSS). • Neutronic calculation performed for proposed systems as SSS. • Dumping the heavy water stored in the reflector vessel is capable to shut down reactor. • Neutronic and transient calculation was done for validating the selected SSS. • All calculation shown that this system has advantages in safety and neutron economy. - Abstract: One important safety aspect of any research reactor is the ability to shut down the reactor. Usually, research reactors, currently in operation, have a single shutdown system based on the simultaneous insertion of the all control rods into the reactor core through gravity. Nevertheless, the International Atomic Energy Agency currently recommends use of two shutdown systems which are fully independent from each other to guarantee secure shutdown when one of them fails. This work presents an investigative study into secondary shutdown systems, which will be an important safety component in the research reactor and will provide another alternative way to shut down the reactor emergently. As part of this project, a study was undertaken to summarize the techniques that are currently used at world-wide research reactors for recognizing available techniques to consider in research reactors. Removal of the reflector, removal of the fuels, change in critical shape of reactor core and insertion of neutron absorber between the core and reflector are selected as possible techniques in mentioned function. In the next step, a comparison is performed for these methods from neutronic aspects. Then, chosen method is studied from the transient behavior point of view. Tehran research reactor which is a 5 MW open-pool reactor selected as a case study and all calculations are carried out for it. It has 5 control rods which serve the purpose of both reactivity control and shutdown of reactor under abnormal condition. Results indicated that heavy

  11. Cryogenic system design for a compact tokamak reactor

    International Nuclear Information System (INIS)

    Slack, D.S.; Kerns, J.A.; Miller, J.R.

    1988-01-01

    The International Tokamak Engineering Reactor (ITER) is a program presently underway to design a next-generation tokamak reactor. The cryogenic system for this reactor must meet unusual and new requirements. Unusually high heat loads (100 kW at 4.5 K) must be handled because neutron shielding has been limited to save space in the reactor core. Also, large variations in the cryogenics loads occur over short periods of time because of the pulsed nature of some of the operating scenarios. This paper describes a workable cryogenic system design for a compact tokamak reactor such as ITER. A design analysis is presented dealing with a system that handles transient loads, coil quenches, reactor cool-down and the effect of variations in helium-supply temperatures on the cryogenic stability of the coils. 5 refs., 4 figs., 1 tab

  12. CRNL research reactor retrofit Emergency Filtration System

    International Nuclear Information System (INIS)

    Philippi, H.M.

    1990-01-01

    This paper presents a brief history of NRX and NRU research reactor effluent air treatment systems before describing the selection and design of an appropriate retrofit Emergency Filtration System (EFS) to serve these reactors and the future MX-10 isotope production reactor. The conceptual design of the EFS began in 1984. A standby concrete shielding filter-adsorber system, sized to serve the reactor with the largest exhaust flow, was selected. The standby system, bypassed under normal operating conditions, is equipped with normal exhaust stream shutoff and diversion valves to be activated manually when an emergency is anticipated, or automatically when emergency levels of gamma radiation are detected in the exhaust stream. The first phase of the EFS installation, that is the construction of the EFS and the connection of NRU to the system, was completed in 1987. The second phase of construction, which includes the connection of NRX and provisions for the future connection of MX-10, is to be completed in 1990

  13. 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.)

  14. Bond graph modeling of nuclear reactor dynamics

    International Nuclear Information System (INIS)

    Tylee, J.L.

    1981-01-01

    A tenth-order linear model of a pressurized water reactor (PWR) is developed using bond graph techniques. The model describes the nuclear heat generation process and the transfer of this heat to the reactor coolant. Comparisons between the calculated model response and test data from a small-scale PWR show the model to be an adequate representation of the actual plant dynamics. Possible application of the model in an advanced plant diagnostic system is discussed

  15. 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

  16. 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)

  17. Materials degradation in fission reactors: Lessons learned of relevance to fusion reactor systems

    International Nuclear Information System (INIS)

    Was, Gary S.

    2007-01-01

    The management of materials in power reactor systems has become a critically important activity in assuring the safe, reliable and economical operation of these facilities. Over the years, the commercial nuclear power reactor industry has faced numerous 'surprises' and unexpected occurrences in materials. Mitigation strategies have sometimes solved one problem at the expense of creating another. Other problems have been solved successfully and have motivated the development of techniques to foresee problems before they occur. This paper focuses on three aspects of fission reactor experience that may benefit future fusion systems. The first is identification of parameters and processes that have had a large impact on the behavior of materials in fission systems such as temperature, dose rate, surface condition, gradients, metallurgical variability and effects of the environment. The second is the development of materials performance and failure models to provide a basis for assuring component integrity. Last is the development of proactive materials management programs that identify and pre-empt degradation processes before they can become problems. These aspects of LWR experience along with the growing experience with materials in the more demanding advanced fission reactor systems form the basis for a set of 'lessons learned' to aid in the successful management of materials in fusion reactor systems

  18. Containment atmosphere cooling system for experimental fast reactor 'JOYO'

    International Nuclear Information System (INIS)

    Sasaki, Mikio; Hoshi, Akio; Sato, Morihiko; Takeuchi, Kaoru

    1979-01-01

    The experimental fast reactor ''JOYO'', the first sodium-cooled fast reactor in Japan, achieved the initially licensed full power operation (50 MW) in July 1978 and is now under steady operation. Toshiba has participated in the construction of this reactor as a leading manufacturer and supplied various systems. This article outlines the design philosophy, system concepts and the operating experience of the containment atmosphere cooling system which has many design interfaces throughout the whole plant and requires especially high reliability. The successful performance of this system during the reactor full-power operation owes to the spot cooling design philosophy and to the preoperational adjustment of heat load during the preheating period of reactor cooling system peculiar to FBR. (author)

  19. Intercomparison of techniques for inspection and diagnostics of heavy water reactor pressure tubes: Flaw detection and characterization [Phase 1

    International Nuclear Information System (INIS)

    2006-05-01

    Nuclear power plants with heavy water reactors (HWRs) comprise nine percent of today's operating nuclear units, and more are under construction. Efficient and accurate inspection and diagnostic techniques for various reactor components and systems are an important factor in assuring reliable and safe plant operation. To foster international collaboration in the efficient and safe use of nuclear power, the IAEA conducted a Coordinated Research Programme (CRP) on Inter-comparison of Techniques for HWR Pressure Tube Inspection and Diagnostics. This CRP was carried out within the frame of the IAEA Department of Nuclear Energy's Technical Working Group on Advanced Technologies for HWRs (the TWG-HWR). The TWG-HWR is a group of experts nominated by their governments and designated by the IAEA to provide advice and to support implementation of the IAEA's project on advanced technologies for HWRs. The objective of the CRP was to inter-compare non-destructive inspection and diagnostic techniques, in use and being developed, for structural integrity assessment of HWR pressure tubes. During the first phase of this CRP, participants have investigated the capability of different techniques to detect and characterize flaws. During the second phase of this CRP, participants collaborated to detect and characterize hydride blisters and to determine the hydrogen concentration in Zirconium alloys. The intent was to identify the most effective pressure tube inspection and diagnostic methods, and to identify further development needs. The organizations that have participated in this CRP are: - The Comision Nacional de Energia Atomica (CNEA), Argentina; - Atomic Energy of Canada Ltd. (AECL); Chalk River Laboratories (CRL), Canada; - The Research Institute of Nuclear Power Operations (RINPO), China National Nuclear Corporation (CNNC), China; - Bhabha Atomic Research Centre (BARC), India; - The Korea Electric Power Research Institute (KEPRI), Republic of Korea; - The Korea Atomic Energy

  20. 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

  1. Choice of thermal reactor systems: a report

    Energy Technology Data Exchange (ETDEWEB)

    1977-09-01

    This is a report by the UK National Nuclear Corporation published by the UK Secretary of State for Energy (Mr. Benn) on 29th July 1977. It is concerned with the advantages and disadvantages of three thermal reactor systems -the AGR (advanced gas cooled reactor), the PWR (pressurised water reactor), and the SGHWR (steam generating heavy water reactor). The object was to help in the future choice of a thermal system for the UK to cover the next 25 years. The matter of export potential is also considered. A programme of four stations of 1100 to 1300 MW each over six years starting from 1979 was assumed. It is emphasised that a decision must be taken now both about reactor systems and actual orders. Headings are as follows: Extract from conclusions reached; Summary of main features of assessment; General conclusions regarding the following - safety, security of the investment, operational characteristics, development and launching requirements, effect on industry, and capital and generation costs. It is stated that in order to make an overall judgement on reactor choice the technical, commercial and social issues involved must be weighed in conjunction with cost differentials.

  2. Implementing artificial neural networks in nuclear power plants diagnostic systems: issues and challenges

    International Nuclear Information System (INIS)

    Boger, Z.

    1998-01-01

    A recent review of artificial intelligence applications in nuclear power plants (NPP) diagnostics and fault detection finds that mostly expert systems (ES) and artificial neural networks (ANN) techniques were researched and proposed, but the number of actual implementations in NPP diagnostics systems is very small. It lists the perceived obstacles to the ANN-based system acceptance and implementation. This paper analyses this list. Some of ANN limitations relate to 'quantitative' difficulties of designing and training large-scale ANNs. The availability of an efficient large-scale ANN training algorithm may alleviate most of these concerns. Other perceived drawbacks refer to the 'qualitative' aspects of ANN acceptance - how and when can we rely on the quality of the advice given by the ANN model. Several techniques are available that help to brighten the 'black box' image of the ANN. Analysis of the trained ANN can identify the significant inputs. Calculation of the Causal Indices may reveal the magnitude and sign of the influence of each input on each output. Both these techniques increase the confidence of the users when they conform to known knowledge, or point to plausible relationships. Analysis of the behavior of the neurons in the hidden layer can identify false ANN classification when presented with noisy or corrupt data. Auto-associative NN can identify faulty sensors or data. Two examples of the ANN capabilities as possible diagnostic tools are given, using NPP data, one classifying internal reactor disturbances by neutron noise spectra analysis, the other identifying the faults causes of several transients. To use these techniques the ANN developers need large amount of training data of as many transients as possible. Such data is routinely generated in NPP simulators during the periodic qualification of NPP operators. The IAEA can help by encouraging the saving and distributing the transient data to developers of ANN diagnostic system, to serve as

  3. Enhancing the functionality of reactor protection systems to provide diagnostic and monitoring information: The ISAT{sup TM} approach

    Energy Technology Data Exchange (ETDEWEB)

    Baldwin, J A; Rowe, B J [AEA Technology, Winfrith (United Kingdom); Jones, C D [Nuclear Electric Ltd., Kent (United Kingdom). Dungeness ` B` Power Station

    1997-12-31

    The ISAT{sup TM} architecture has been successfully implemented as the Single Channel Trip System (SCTS), part of the primary protection system of Nuclear Electric`s Dungeness `B` Advanced Gas-Cooled Reactors. The system is the first computer-based protection system licensed on a UK civil nuclear reactor. The system provides protection against single channel faults resulting in high coolant gas outlet temperature. The SCTS was designed to output data at several points in the system to an Ethernet to allow checks to be made on the operation of parts of the protection system and the system as a whole. In order to monitor the performance of this shutdown system a PC based monitoring system was developed to take input as data from the Ethernet, check its integrity and then analyze the data to provide information of the state of the system and subsystems. The SCTS monitor was basically intended to alert the operator to any fault on the safety system and indicate its source, provide a diagnosis of the cause of any trip initiated by the safety system, and log the occurrences of these incidents for later inspection. The intention was also to provide accurate real-time information on the thermocouple readings and to decrease the effort required to maintain the safety system. This paper will describe briefly the development of the ISAT{sup TM} monitoring system: how its requirements were arrived at, and how the design, code and testing were carried out to ensure approval for this application. It will then go on to report how the ISAT{sup TM} monitor has performed during its time in service: how more functionality has been added over and above its original requirements. Features of additional monitors for the SCTS and other ISAT{sup TM} systems will also be described. (author). 2 refs, 5 figs.

  4. Wavelet and Spectral Analysis of Some Selected Problems in Reactor Diagnostics

    Energy Technology Data Exchange (ETDEWEB)

    Sunde, Carl

    2004-12-01

    Both spectral and wavelet analysis were successfully used in various diagnostic problems involving non-stationary core processes in nuclear power reactors. Three different problems were treated: two-phase flow identification, detector tube impacting and core-barrel vibrations. The first two problems are of non-stationary nature, whereas the last one is not. In the first problem, neutron radiographic and visible light images of four different vertical two-phase flow regimes, bubbly, slug, chum and annular flow, were analysed and classified with a neuro-wavelet algorithm. The algorithm consists of a wavelet part, using the 2-D discrete wavelet transform and of an artificial neural network. It classifies the different flow regimes with up to 99% efficiency. Detector tubes in a Boiling Water Reactor may execute vibrations and may also impact on nearby fuel-assemblies. Signals from in-core neutron detectors in Ringhals-1 were analysed, for detection of impacting, with both a classical spectral method and wavelet-based methods. The wavelet methods include both the discrete and the continuous 1-D wavelet transform. It was found that there is agreement between the different methods as well as with visual inspections made during the outage at the plant. However, the wavelet technique has the advantage that it does not require expert judgement for the interpretation of the analysis. In the last part two analytical calculations of the neutron noise, induced by shell-mode core-barrel vibrations, were carried out. The results are in good agreement with calculations from a numerical simulator. An out-of-phase behaviour between in-core and ex-core positions was found, which is in agreement with earlier measurements from the Pressurised Water Reactor Ringhals-3. The results from these calculations are planned to be used when diagnosing the shell-mode core-barrel vibrations in an operating plant.

  5. Wavelet and Spectral Analysis of Some Selected Problems in Reactor Diagnostics

    International Nuclear Information System (INIS)

    Sunde, Carl

    2004-12-01

    Both spectral and wavelet analysis were successfully used in various diagnostic problems involving non-stationary core processes in nuclear power reactors. Three different problems were treated: two-phase flow identification, detector tube impacting and core-barrel vibrations. The first two problems are of non-stationary nature, whereas the last one is not. In the first problem, neutron radiographic and visible light images of four different vertical two-phase flow regimes, bubbly, slug, chum and annular flow, were analysed and classified with a neuro-wavelet algorithm. The algorithm consists of a wavelet part, using the 2-D discrete wavelet transform and of an artificial neural network. It classifies the different flow regimes with up to 99% efficiency. Detector tubes in a Boiling Water Reactor may execute vibrations and may also impact on nearby fuel-assemblies. Signals from in-core neutron detectors in Ringhals-1 were analysed, for detection of impacting, with both a classical spectral method and wavelet-based methods. The wavelet methods include both the discrete and the continuous 1-D wavelet transform. It was found that there is agreement between the different methods as well as with visual inspections made during the outage at the plant. However, the wavelet technique has the advantage that it does not require expert judgement for the interpretation of the analysis. In the last part two analytical calculations of the neutron noise, induced by shell-mode core-barrel vibrations, were carried out. The results are in good agreement with calculations from a numerical simulator. An out-of-phase behaviour between in-core and ex-core positions was found, which is in agreement with earlier measurements from the Pressurised Water Reactor Ringhals-3. The results from these calculations are planned to be used when diagnosing the shell-mode core-barrel vibrations in an operating plant

  6. 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

  7. RELAP/SCDAPSIM Reactor System Simulator Development and Training for University and Reactor Applications

    International Nuclear Information System (INIS)

    Hohorst, J.K.; Allison, C.M.

    2010-01-01

    The RELAP/SCDAPSIM code, designed to predict the behaviour of reactor systems during normal and accident conditions, is being developed as part of an international nuclear technology development program called SDTP (SCDAP Development and Training Program). SDTP involves more than 60 organizations in 28 countries. One of the important applications of the code is for simulator training of university faculty and students, reactor analysts, and reactor operations and technical support staff. Examples of RELAP/SCDAPSIM-based system thermal hydraulic and severe accident simulator packages include the SAFSIM simulator developed by NECSA for the SAFARI research reactor in South Africa, university-developed simulators at the University of Mexico and Shanghai Jiao Tong University in China, and commercial VISA and RELSIM packages used for analyst and reactor operations staff training. This paper will briefly describe the different packages/facilities. (authors)

  8. RELAP/SCDAPSIM Reactor System Simulator Development and Training for University and Reactor Applications

    Energy Technology Data Exchange (ETDEWEB)

    Hohorst, J.K.; Allison, C.M. [Innovative Systems Software, 1242 South Woodruff Avenue, Idaho Falls, Idaho 83404 (United States)

    2010-07-01

    The RELAP/SCDAPSIM code, designed to predict the behaviour of reactor systems during normal and accident conditions, is being developed as part of an international nuclear technology development program called SDTP (SCDAP Development and Training Program). SDTP involves more than 60 organizations in 28 countries. One of the important applications of the code is for simulator training of university faculty and students, reactor analysts, and reactor operations and technical support staff. Examples of RELAP/SCDAPSIM-based system thermal hydraulic and severe accident simulator packages include the SAFSIM simulator developed by NECSA for the SAFARI research reactor in South Africa, university-developed simulators at the University of Mexico and Shanghai Jiao Tong University in China, and commercial VISA and RELSIM packages used for analyst and reactor operations staff training. This paper will briefly describe the different packages/facilities. (authors)

  9. Pattern recognition application for surveillance of abnormal conditions in a nuclear reactor

    International Nuclear Information System (INIS)

    Pepelyshev, Yu.N.; Dzwinel, W.

    1990-01-01

    The system to monitor abnormal conditions in a nuclear reactor, based on the noise analysis of the reactor basic parameters such as power, temperature and coolant flow rate, has been developed. The pattern recognition techniques such as clustering, cluster analysis, feature selection and clusters visualization methods form the basis of the software. Apart from non-hierarchical clustering procedures applied earlier, the hierarchical one is recommended. The system application for IBR-2 Dubna reactor diagnostics is shown. 10 refs.; 6 figs

  10. Nuclear Energy Research Initiative (NERI): On-Line Intelligent Self-Diagnostic Monitoring for Next Generation Nuclear Plants - Phase I Annual Report

    International Nuclear Information System (INIS)

    Bond, L.G.; Doctor, S.R.; Gilbert, R.W.; Jarrell, D.B.; Greitzer, F.L.; Meador, R.J.

    2000-01-01

    OAK-B135 This OSTI ID belongs to an IWO and is being released out of the system. The Program Manager Rebecca Richardson has confirmed that all reports have been received. The objective of this project is to design and demonstrate the operation of the real-time intelligent self-diagnostic and prognostic system for next generation nuclear power plant systems. This new self-diagnostic technology is titled, ''On-Line Intelligent Self-Diagnostic Monitoring System'' (SDMS). This project provides a proof-of-principle technology demonstration for SDMS on a pilot plant scale service water system, where a distributed array of sensors is integrated with active components and passive structures typical of next generation nuclear power reactor and plant systems. This project employs state-of-the-art sensors, instrumentation, and computer processing to improve the monitoring and assessment of the power reactor system and to provide diagnostic and automated prognostics capabilities

  11. 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)

  12. 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

  13. Intelligent neural network diagnostic system

    International Nuclear Information System (INIS)

    Mohamed, A.H.

    2010-01-01

    Recently, artificial neural network (ANN) has made a significant mark in the domain of diagnostic applications. Neural networks are used to implement complex non-linear mappings (functions) using simple elementary units interrelated through connections with adaptive weights. The performance of the ANN is mainly depending on their topology structure and weights. Some systems have been developed using genetic algorithm (GA) to optimize the topology of the ANN. But, they suffer from some limitations. They are : (1) The computation time requires for training the ANN several time reaching for the average weight required, (2) Slowness of GA for optimization process and (3) Fitness noise appeared in the optimization of ANN. This research suggests new issues to overcome these limitations for finding optimal neural network architectures to learn particular problems. This proposed methodology is used to develop a diagnostic neural network system. It has been applied for a 600 MW turbo-generator as a case of real complex systems. The proposed system has proved its significant performance compared to two common methods used in the diagnostic applications.

  14. Plasma driving system requirements for commercial tokamak fusion reactors

    International Nuclear Information System (INIS)

    Brooks, J.N.; Kustom, R.C.; Stacey, W.M. Jr.

    1978-01-01

    The plasma driving system for a tokamak reactor is composed of an ohmic heating (OH) coil, equilibrium field (EF) coil, and their respective power supplies. Conceptual designs of an Experimental Power Reactor (EPR) and scoping studies of a Demonstration Power Reactor have shown that the driving system constitutes a significant part of the overall reactor cost. The capabilities of the driving system also set or help set important parameters of the burn cycle, such as the startup time, and the net power output. Previous detailed studies on driving system dynamics have helped to define the required characteristics for fast-pulsed superconducting magnets, homopolar generators, and very high power (GVA) power supplies for an EPR. This paper summarizes results for a single reactor configuration together with several design concepts for the driving system. Both the reactor configuration and the driving system concepts are natural extensions from the EPR. Thus, the new results presented in this paper can be compared with the previous EPR results to obtain a consistent picture of how the driving system requirements will evolve--for one particular design configuration

  15. Plasma driving system requirements for commercial tokamak fusion reactors

    International Nuclear Information System (INIS)

    Brooks, J.N.; Kustom, R.C.; Stacey, W.M. Jr.

    1977-01-01

    The plasma driving system for a tokamak reactor is composed of an ohmic heating (OH) coil, equilibrium field (EF) coil, and their respective power supplies. Conceptual designs of an Experimental Power Reactor (EPR) and scoping studies of a Demonstration Power Reactor have shown that the driving system constitutes a significant part of the overall reactor cost. The capabilities of the driving system also set or help set important parameters of the burn cycle, such as the startup time, and the net power output. Previous detailed studies on driving system dynamics have helped to define the required characteristics for fast-pulsed superconducting magnets, homopolar generators, and very high power (GVA) power supplies for an EPR. This paper summarizes results for a single reactor configuration together with several design concepts for the driving system. Both the reactor configuration and the driving system concepts are natural extensions from the EPR. Thus, the new results can be compared with the previous EPR results to obtain a consistent picture of how the driving system requirements will evolve--for one particular design configuration

  16. SRAC: JAERI thermal reactor standard code system for reactor design and analysis

    International Nuclear Information System (INIS)

    Tsuchihashi, Keichiro; Takano, Hideki; Horikami, Kunihiko; Ishiguro, Yukio; Kaneko, Kunio; Hara, Toshiharu.

    1983-01-01

    The SRAC (Standard Reactor Analysis Code) is a code system for nuclear reactor analysis and design. It is composed of neutron cross section libraries and auxiliary processing codes, neutron spectrum routines, a variety of transport, 1-, 2- and 3-D diffusion routines, dynamic parameters and cell burn-up routines. By making the best use of the individual code function in the SRAC system, the user can select either the exact method for an accurate estimate of reactor characteristics or the economical method aiming at a shorter computer time, depending on the purpose of study. The user can select cell or core calculation; fixed source or eigenvalue problem; transport (collision probability or Sn) theory or diffusion theory. Moreover, smearing and collapsing of macroscopic cross sections are separately done by the user's selection. And a special attention is paid for double heterogeneity. Various techniques are employed to access the data storage and to optimize the internal data transfer. Benchmark calculations using the SRAC system have been made extensively for the Keff values of various types of critical assemblies (light water, heavy water and graphite moderated systems, and fast reactor systems). The calculated results show good prediction for the experimental Keff values. (author)

  17. A knowledge based on-line diagnostic system for the fast breeder reactor KNKII

    International Nuclear Information System (INIS)

    Eggert, H.; Scherer, K.P.; Stiller, P.

    1989-01-01

    In the nuclear research center at Karlsruhe, a diagnostic expert system is developed to supervise a fast breeder process (KNKII). The problem is to detect critical phases in the beginning state before fault propagation. The expert system itself is integrated in a computer network (realized by a local area network), where different computers are involved as special detection systems (for example acoustic noise, temperature noise, covergas monitoring and so on), which produce partial diagnoses, based on intelligent signal processing techniques like pattern recognition. Additional to the detection systems a process computer is integrated as well as a test computer, which simulates hypothetical and real fault data. On the logical top level the expert system manages the partial diagnoses of the detection systems with the operating data of the process computer and to produce a final diagnosis including the explanation part for operator support. The knowledge base is developed by typical Artificial Intelligence tools. Both fact based and rule based knowledge representations are stored in form of flavors and predications. The inference engine operates on a rule based approach. Specific detail knowledge, based on experience about any years, is available to influence the decision process by increasing or decreasing of the generated hypotheses. In a meta knowledge base, a rule master triggers the special domain experts and contributes the tasks to the specific rule complexes. Such a system management guarantees a problem solving strategy, which operates event triggered and situation specific in a local inference domain. (author). 3 refs, 6 figs, 2 tabs

  18. Radiolytic production of chemical fuels in fusion reactor systems

    Energy Technology Data Exchange (ETDEWEB)

    Fish, J D

    1977-06-01

    Miley's energy flow diagram for fusion reactor systems is extended to include radiolytic production of chemical fuel. Systematic study of the economics and the overall efficiencies of fusion reactor systems leads to a criterion for evaluating the potential of radiolytic production of chemical fuel as a means of enhancing the performance of a fusion reactor system. The ecumenicity of the schema is demonstrated by application to (1) tokamaks, (2) mirror machines, (3) theta-pinch reactors, (4) laser-heated solenoids, and (5) inertially confined, laser-pellet devices. Pure fusion reactors as well as fusion-fission hybrids are considered.

  19. Radiolytic production of chemical fuels in fusion reactor systems

    International Nuclear Information System (INIS)

    Fish, J.D.

    1977-06-01

    Miley's energy flow diagram for fusion reactor systems is extended to include radiolytic production of chemical fuel. Systematic study of the economics and the overall efficiencies of fusion reactor systems leads to a criterion for evaluating the potential of radiolytic production of chemical fuel as a means of enhancing the performance of a fusion reactor system. The ecumenicity of the schema is demonstrated by application to (1) tokamaks, (2) mirror machines, (3) theta-pinch reactors, (4) laser-heated solenoids, and (5) inertially confined, laser-pellet devices. Pure fusion reactors as well as fusion-fission hybrids are considered

  20. Cable handling system for use in a nuclear reactor

    International Nuclear Information System (INIS)

    Crosgrove, R.O.; Larson, E.M.; Moody, E.

    1982-01-01

    A cable handling system for use in an installation such as a nuclear reactor is disclosed herein along with relevant portions of the reactor which, in a preferred embodiment, is a liquid metal fast breeder reactor. The cable handling system provides a specific way of interconnecting certain internal reactor components with certain external components, through an assembly of rotatable plugs. Moreover, this is done without having to disconnect these components from one another during rotation of the plugs and yet without interfering with other reactor components in the vicinity of the rotating plugs and cable handling system

  1. Reactor noise in critical and accelerator driven sub-critical systems

    International Nuclear Information System (INIS)

    Degweker, S.B.; Rana, Y.S.

    2007-01-01

    Noise methods have long been used for reactor kinetics parameters measurement and as diagnostic tools for monitoring the health of a nuclear power plant. It is conceivable that noise techniques would find similar applications in ADS. Measurement/monitoring the degree of sub-criticality of an ADS is one such application for which noise based methods are being considered, among others such as the pulsed source method. For this reason, theoretical studies on ADS noise have appeared since the late nineties. The principal difference between critical reactor noise and ADS noise is due to the statistical properties of the source. Unlike the source due to radioactive decay present in ordinary reactors, the machine produced ADS source cannot be assumed to be a Poisson process. In addition the source is pulsed. All this requires a new theoretical approach to the subject. In a number of papers (beginning in 2000) such a theoretical approach has been developed in BARC. Over the years, our approach has received general acceptance. The paper gives a description of the subject of reactor noise and its applications in critical reactors. The theory of noise in ADS is then outlined, highlighting the differences in approach and results from that of critical reactors. (author)

  2. 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.)

  3. Development of multi-functional telerobotic systems for reactor dismantlement

    International Nuclear Information System (INIS)

    Fujii, Yoshio; Usui, Hozumi; Shinohara, Yoshikuni

    1992-01-01

    This report summarizes technological features of advanced telerobotic systems for reactor dismantling application developed at the Japan Atomic Energy Research Institute. Taking into consideration the special environmental conditions in reactor dismantling, major effort was made to develop multifunctional telerobotic system of high reliability which can be used to perform various complex tasks in an unstructured environment and operated in an easy and flexible manner. The system development was carried out through constructing three systems in seccession; a light-duty and a heavy-duty system as a prototype system for engineering test in cold environment, and a demonstration system for practical on-site application to dismantling highly radioactive reactor internals of an experimental boiling water reactor JPDR (Japan Power Demonstration Reactor). Each system was equipped with one or two amphibious manipulators which can be operated in either a push-button manual, a bilateral master-slave, a teach-and-playback or a programmed control mode. Different scheme was adopted in each system at designing the manipulator, transporter and man-machine interface so as to compare their advantages and disadvantages. According to the JPDR decommissioning program, the demonstration system was successfully operated to dismantle a portion of the radioactive reactor internals of the JPDR, which used underwater plasma arc cutting method and proved the usefulness of the multi-functional telerobotic system for reducing the occupational hazards and enhancing the work efficiency in the course of dismantling highly radioactive reactor components. (author)

  4. Diagnostic and prognostic system for identification of accident scenarios and prediction of 'source term' in nuclear power plants under accident conditions

    International Nuclear Information System (INIS)

    Santhosh; Gera, B.; Kumar, Mithilesh

    2014-01-01

    Nuclear power plant experiences a number of transients during its operations. These transients may be due to equipment failure, malfunctioning of process support systems etc. In such a situation, the plant may result in an abnormal state which is undesired. In case of such an undesired plant condition, the operator has to carry out diagnostic and corrective actions. When an event occurs starting from the steady state operation, instruments' readings develop a time dependent pattern and these patterns are unique with respect to the type of the particular event. Therefore, by properly selecting the plant process parameters, the transients can be distinguished. In this connection, a computer based tool known as Diagnostic and Prognostic System has been developed for identification of large pipe break scenarios in 220 MWe Pressurised Heavy Water Reactors (PHWRs) and for prediction of expected 'Source Term' and consequence for a situation where Emergency Core Cooling System (ECCS) is not available or partially available. Diagnostic and Prognostic System is essentially a transient identification and expected source term forecasting system. The system is based on Artificial Neural Networks (ANNs) that continuously monitors the plant conditions and identifies a Loss Of Coolant Accident (LOCA) scenario quickly based on the reactor process parameter values. The system further identifies the availability of injection of ECCS and in case non-availability of ECCS, it can forecast expected 'Source Term'. The system is a support to plant operators as well as for emergency preparedness. The ANN is trained with a process parameter database pertaining to accident conditions and tested against blind exercises. In order to see the feasibility of implementing in the plant for real-time diagnosis, this system has been set up on a high speed computing facility and has been demonstrated successfully for LOCA scenarios. (author)

  5. IMPLEMENTATION OF TURNOUTS TECHNICAL DIAGNOSTICS SYSTEMS

    Directory of Open Access Journals (Sweden)

    S. YU. Buryak

    2015-06-01

    Full Text Available Purpose. In the paper it is necessary to: 1 find out the causes of turnouts faults to determine diagnostic features failures; 2 consider the requirements structure, purpose components of turnouts, work and technology of their maintenance to determine the construction of the economic activities related to system to the turnout’s maintenance; 3 substantiate the possibility, necessity and prospects of automated diagnostics turnout’s implementation; 4 elaborate a prototype of an automated hardware and software system for the turnouts control parameters and perform diagnostics on them. Methodology. In the paper possible turnouts faults were presented and manifestations and influence on its work were shown. According to the current technology works the process analyze of turnouts’ maintenance was conducted, were defined the basic performed operations during the examination of appearance, parameters and check the repair or replacement of parts and assemblies. Based on the analysis of reasons of turnouts malfunctioning and their fixes were systematized types of damages and ways to deal with them, an information scheme of troubleshooting were created, opportunities and limits of automating the process of diagnostics were identified and compared with the existing method of turnouts maintenance. A diagnostics system block diagram was created, an algorithm of its work was developed and established main basic principles of operation. Software and hardware to determine the turnout’s state considering diagnostic performance of points in use were applied. Findings. During the experiment was created a method of automated turnout’s diagnostics with AC electric drives, managed centrally. The results of automated hardware and software system make it possible to control turnout’s parameters and perform diagnostics on them. Originality. Authors created the method of turnout’s state determination by current curve and its spectral composition in the

  6. 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)

  7. Microchannel Reactor System for Catalytic Hydrogenation

    Energy Technology Data Exchange (ETDEWEB)

    Adeniyi Lawal; Woo Lee; Ron Besser; Donald Kientzler; Luke Achenie

    2010-12-22

    We successfully demonstrated a novel process intensification concept enabled by the development of microchannel reactors, for energy efficient catalytic hydrogenation reactions at moderate temperature, and pressure, and low solvent levels. We designed, fabricated, evaluated, and optimized a laboratory-scale microchannel reactor system for hydrogenation of onitroanisole and a proprietary BMS molecule. In the second phase of the program, as a prelude to full-scale commercialization, we designed and developed a fully-automated skid-mounted multichannel microreactor pilot plant system for multiphase reactions. The system is capable of processing 1 – 10 kg/h of liquid substrate, and an industrially relevant immiscible liquid-liquid was successfully demonstrated on the system. Our microreactor-based pilot plant is one-of-akind. We anticipate that this process intensification concept, if successfully demonstrated, will provide a paradigm-changing basis for replacing existing energy inefficient, cost ineffective, environmentally detrimental slurry semi-batch reactor-based manufacturing practiced in the pharmaceutical and fine chemicals industries.

  8. Automatic acoustic and vibration monitoring system for nuclear power plants

    International Nuclear Information System (INIS)

    Tothmatyas, Istvan; Illenyi, Andras; Kiss, Jozsef; Komaromi, Tibor; Nagy, Istvan; Olchvary, Geza

    1990-01-01

    A diagnostic system for nuclear power plant monitoring is described. Acoustic and vibration diagnostics can be applied to monitor various reactor components and auxiliary equipment including primary circuit machinery, leak detection, integrity of reactor vessel, loose parts monitoring. A noise diagnostic system has been developed for the Paks Nuclear Power Plant, to supervise the vibration state of primary circuit machinery. An automatic data acquisition and processing system is described for digitalizing and analysing diagnostic signals. (R.P.) 3 figs

  9. ATA diagnostic data handling system: an overview

    International Nuclear Information System (INIS)

    Chambers, F.W.; Kallman, J.; McDonald, J.; Slominski, M.

    1984-01-01

    The functions to be performed by the ATA diagnostic data handling system are discussed. The capabilities of the present data acquisition system (System 0) are presented. The goals for the next generation acquisition system (System 1), currently under design, are discussed. Facilities on the Octopus system for data handling are reviewed. Finally, we discuss what has been learned about diagnostics and computer based data handling during the past year

  10. Water chemistry data acquisition, processing, evaluation and diagnostic systems in Light Water Reactors: Future improvement of plant reliability and safety

    International Nuclear Information System (INIS)

    Uchida, S.; Takiguchi, H.; Ishigure, K.

    2006-01-01

    Data acquisition, processing and evaluation systems have been applied in major Japanese PWRs and BWRs to provide (1) reliable and quick data acquisition with manpower savings in plant chemical laboratories and (2) smooth and reliable information transfer among chemists, plant operators, and supervisors. Data acquisition systems in plants consist of automatic and semi-automatic instruments for chemical analyses, e. g., X-ray fluorescence analysis and ion chromatography, while data processing systems consist of PC base-sub-systems, e.g., data storage, reliability evaluation, clear display, and document preparation for understanding the plant own water chemistry trends. Precise and reliable evaluations of water chemistry data are required in order to improve plant reliability and safety. For this, quality assurance of the water chemistry data acquisition system is needed. At the same time, theoretical models are being applied to bridge the gaps between measured water chemistry data and the information desired to understand the interaction of materials and cooling water in plants. Major models which have already been applied for plant evaluation are: (1) water radiolysis models for BWRs and PWRs; (2) crevice radiolysis model for SCC in BWRs; and (3) crevice pH model for SG tubing in PWRs. High temperature water chemistry sensors and automatic plant diagnostic systems have been applied in only restricted areas. ECP sensors are gaining popularity as tools to determine the effects of hydrogen injection in BWR systems. Automatic plant diagnostic systems based on artificial intelligence will be more popular after having sufficient experience with off line diagnostic systems. (author)

  11. Creation of reactor's reliable system of emergency energy supply

    International Nuclear Information System (INIS)

    Batyrbekov, G.A.; Brovkin, A.Yu.; Petukhov, V.K.; Chekushin, A.I.; Chernyaev, V.P.; Yagotinets, N.A.

    1998-01-01

    System of reliable power supply of the WWR-K reactor complex is described, which completely provides safety operation of reactor equipment in the case of total voltage loss from external power transmission lines as well as under destruction of accumulation batteries by earthquake more than 6 balls. Switching on in operation of diesel-generators and system of constant current supply from accumulator batteries is occurred automatically under cessation of voltage supply from centralized power system. Reliable reactor dampening in case it work on capacity has been ensured. Reactor cooling under its emergency shutdown during both the partial or the total loss of coolant in first counter has been carried out. Under full coolant loss the system of emergency reactor cooling has been switched on in operation

  12. The combined hybrid system: A symbiotic thermal reactor/fast reactor system for power generation and radioactive waste toxicity reduction

    International Nuclear Information System (INIS)

    Hollaway, W.R.

    1991-08-01

    If there is to be a next generation of nuclear power in the United States, then the four fundamental obstacles confronting nuclear power technology must be overcome: safety, cost, waste management, and proliferation resistance. The Combined Hybrid System (CHS) is proposed as a possible solution to the problems preventing a vigorous resurgence of nuclear power. The CHS combines Thermal Reactors (for operability, safety, and cost) and Integral Fast Reactors (for waste treatment and actinide burning) in a symbiotic large scale system. The CHS addresses the safety and cost issues through the use of advanced reactor designs, the waste management issue through the use of actinide burning, and the proliferation resistance issue through the use of an integral fuel cycle with co-located components. There are nine major components in the Combined Hybrid System linked by nineteen nuclear material mass flow streams. A computer code, CHASM, is used to analyze the mass flow rates CHS, and the reactor support ratio (the ratio of thermal/fast reactors), IFR of the system. The primary advantages of the CHS are its essentially actinide-free high-level radioactive waste, plus improved reactor safety, uranium utilization, and widening of the option base. The primary disadvantages of the CHS are the large capacity of IFRs required (approximately one MW e IFR capacity for every three MW e Thermal Reactor) and the novel radioactive waste streams produced by the CHS. The capability of the IFR to burn pure transuranic fuel, a primary assumption of this study, has yet to be proven. The Combined Hybrid System represents an attractive option for future nuclear power development; that disposal of the essentially actinide-free radioactive waste produced by the CHS provides an excellent alternative to the disposal of intact actinide-bearing Light Water Reactor spent fuel (reducing the toxicity based lifetime of the waste from roughly 360,000 years to about 510 years)

  13. Computer System Analysis for Decommissioning Management of Nuclear Reactor

    International Nuclear Information System (INIS)

    Nurokhim; Sumarbagiono

    2008-01-01

    Nuclear reactor decommissioning is a complex activity that should be planed and implemented carefully. A system based on computer need to be developed to support nuclear reactor decommissioning. Some computer systems have been studied for management of nuclear power reactor. Software system COSMARD and DEXUS that have been developed in Japan and IDMT in Italy used as models for analysis and discussion. Its can be concluded that a computer system for nuclear reactor decommissioning management is quite complex that involved some computer code for radioactive inventory database calculation, calculation module on the stages of decommissioning phase, and spatial data system development for virtual reality. (author)

  14. Safety Analysis for Power Reactor Protection System

    International Nuclear Information System (INIS)

    Eisawy, E.A.; Sallam, H.

    2012-01-01

    The main function of a Reactor Protection System (RPS) is to safely shutdown the reactor and prevents the release of radioactive materials. The purpose of this paper is to present a technique and its application for used in the analysis of safety system of the Nuclear Power Plant (NPP). A more advanced technique has been presented to accurately study such problems as the plant availability assessments and Technical Specifications evaluations that are becoming increasingly important. The paper provides the Markov model for the Reactor Protection System of the NPP and presents results of model evaluations for two testing policies in technical specifications. The quantification of the Markov model provides the probability values that the system will occupy each of the possible states as a function of time.

  15. 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)

  16. Irradiation effects on plasma diagnostic components

    International Nuclear Information System (INIS)

    Nishitani, Takeo; Iida, Toshiyuki; Ikeda, Yujiro

    1998-10-01

    One of the most important issues to develop the diagnostics for the experimental thermonuclear reactor such as ITER is the irradiation effects on the diagnostics components. Typical neutron flux and fluence on the first wall are 1 MW/m 2 and 1 MWa/m 2 , respectively for ITER. In such radiation condition, most of the present diagnostics could not survive so that those will be planed to be installed far from the vacuum vessel. However, some diagnostics sensors such as bolometers and magnetic probes still have to be install inside vessel. And many transmission components for lights, wave and electric signals are inevitable even inside vessel. As a part of this R and D program of the ITER Engineering Design Activities (EDA), we carried out the irradiation tests on the basic materials of the transmission components and in-vessel diagnostics sensors in order to identify radiation hardened materials that can be used for diagnostic systems. (J.P.N.)

  17. Irradiation effects on plasma diagnostic components

    Energy Technology Data Exchange (ETDEWEB)

    Nishitani, Takeo [ed.] [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Iida, Toshiyuki; Ikeda, Yujiro [and others

    1998-10-01

    One of the most important issues to develop the diagnostics for the experimental thermonuclear reactor such as ITER is the irradiation effects on the diagnostics components. Typical neutron flux and fluence on the first wall are 1 MW/m{sup 2} and 1 MWa/m{sup 2}, respectively for ITER. In such radiation condition, most of the present diagnostics could not survive so that those will be planed to be installed far from the vacuum vessel. However, some diagnostics sensors such as bolometers and magnetic probes still have to be install inside vessel. And many transmission components for lights, wave and electric signals are inevitable even inside vessel. As a part of this R and D program of the ITER Engineering Design Activities (EDA), we carried out the irradiation tests on the basic materials of the transmission components and in-vessel diagnostics sensors in order to identify radiation hardened materials that can be used for diagnostic systems. (J.P.N.)

  18. Sampling system for a boiling reactor NPP

    International Nuclear Information System (INIS)

    Zabelin, A.I.; Yakovleva, E.D.; Solov'ev, Yu.A.

    1976-01-01

    Investigations and pilot running of the nuclear power plant with a VK-50 boiling reactor reveal the necessity of normalizing the design system of water sampling and of mandatory replacement of the needle-type throttle device by a helical one. A method for designing a helical throttle device has been worked out. The quantitative characteristics of depositions of corrosion products along the line of reactor water sampling are presented. Recommendations are given on the organizaton of the sampling system of a nuclear power plant with BWR type reactors

  19. Nuclear Energy Research Initiative (NERI): On-Line Intelligent Self-Diagnostic Monitoring for Next Generation Nuclear Plants - Phase I Annual Report

    Energy Technology Data Exchange (ETDEWEB)

    L. J. Bond; S. R. Doctor; R. W. Gilbert; D. B. Jarrell; F. L. Greitzer; R. J. Meador

    2000-09-01

    OAK-B135 This OSTI ID belongs to an IWO and is being released out of the system. The Program Manager Rebecca Richardson has confirmed that all reports have been received. The objective of this project is to design and demonstrate the operation of the real-time intelligent self-diagnostic and prognostic system for next generation nuclear power plant systems. This new self-diagnostic technology is titled, ''On-Line Intelligent Self-Diagnostic Monitoring System'' (SDMS). This project provides a proof-of-principle technology demonstration for SDMS on a pilot plant scale service water system, where a distributed array of sensors is integrated with active components and passive structures typical of next generation nuclear power reactor and plant systems. This project employs state-of-the-art sensors, instrumentation, and computer processing to improve the monitoring and assessment of the power reactor system and to provide diagnostic and automated prognostics capabilities.

  20. A Self-Diagnostic System for the M6 Accelerometer

    Science.gov (United States)

    Flanagan, Patrick M.; Lekki, John

    2001-01-01

    The design of a Self-Diagnostic (SD) accelerometer system for the Space Shuttle Main Engine is presented. This retrofit system connects diagnostic electronic hardware and software to the current M6 accelerometer system. This paper discusses the general operation of the M6 accelerometer SD system and procedures for developing and evaluating the SD system. Signal processing techniques using M6 accelerometer diagnostic data are explained. Test results include diagnostic data responding to changing ambient temperature, mounting torque and base mounting impedance.

  1. The software-defined fast post-processing for GEM soft x-ray diagnostics in the Tungsten Environment in Steady-state Tokamak thermal fusion reactor

    Science.gov (United States)

    Krawczyk, Rafał Dominik; Czarski, Tomasz; Linczuk, Paweł; Wojeński, Andrzej; Kolasiński, Piotr; GÄ ska, Michał; Chernyshova, Maryna; Mazon, Didier; Jardin, Axel; Malard, Philippe; Poźniak, Krzysztof; Kasprowicz, Grzegorz; Zabołotny, Wojciech; Kowalska-Strzeciwilk, Ewa; Malinowski, Karol

    2018-06-01

    This article presents a novel software-defined server-based solutions that were introduced in the fast, real-time computation systems for soft X-ray diagnostics for the WEST (Tungsten Environment in Steady-state Tokamak) reactor in Cadarache, France. The objective of the research was to provide a fast processing of data at high throughput and with low latencies for investigating the interplay between the particle transport and magnetohydrodynamic activity. The long-term objective is to implement in the future a fast feedback signal in the reactor control mechanisms to sustain the fusion reaction. The implemented electronic measurement device is anticipated to be deployed in the WEST. A standalone software-defined computation engine was designed to handle data collected at high rates in the server back-end of the system. Signals are obtained from the front-end field-programmable gate array mezzanine cards that acquire and perform a selection from the gas electron multiplier detector. A fast, authorial library for plasma diagnostics was written in C++. It originated from reference offline MATLAB implementations. They were redesigned for runtime analysis during the experiment in the novel online modes of operation. The implementation allowed the benchmarking, evaluation, and optimization of plasma processing algorithms with the possibility to check the consistency with reference computations written in MATLAB. The back-end software and hardware architecture are presented with data evaluation mechanisms. The online modes of operation for the WEST are discussed. The results concerning the performance of the processing and the introduced functionality are presented.

  2. Optical system for the Protein Crystallisation Diagnostics Facility (PCDF) on board the ISS

    Science.gov (United States)

    Joannes, Luc; Dupont, Olivier; Dewandel, Jean-Luc; Ligot, Renaud; Algrain, Hervé

    2004-06-01

    The Protein Crystallisation Diagnostic Facility (PCDF) is a multi-user facility to study the protein crystallisation under the conditions of micro-gravity onboard the International Space Station (ISS) Columbus facility. Large size protein crystals will growth under reduced gravity in thermally controlled reactors. A combination of diagnostic tools like video system, microscope, interferometer, and light scattering device shall help to understand the growth phenomena. Common methods of protein crystallisation shall be performed in PCDF: Dialysis where the protein solution and the salt solution are separated by a semi-permeable membrane. Extended Length Dialysis Batch where the saturation to get crystals is achieved by changing the concentration of the protein in the sample liquid. The overall ESA project is leaded by EADS Space Transportation, Friedrichshafen, Germany. Lambda-X is responsible for the Optical System (OS), with Verhaert Design and Development as sub-contractor for the mechanical design. The OS includes different compact parts: Original illumination systems based on LEDs of difference colours; Quantitative Mach-Zehnder interferometers to measure the concentration distribution around crystals; Imaging assemblies to visualize the protein volume with different field of views. The paper concentrates on the description of each part, and in particular on the imaging assembly which allow switching from one field of view to another by passive elements only.

  3. Fault-tolerant reactor protection system

    International Nuclear Information System (INIS)

    Gaubatz, D.C.

    1997-01-01

    A reactor protection system is disclosed having four divisions, with quad redundant sensors for each scram parameter providing input to four independent microprocessor-based electronic chassis. Each electronic chassis acquires the scram parameter data from its own sensor, digitizes the information, and then transmits the sensor reading to the other three electronic chassis via optical fibers. To increase system availability and reduce false scrams, the reactor protection system employs two levels of voting on a need for reactor scram. The electronic chassis perform software divisional data processing, vote 2/3 with spare based upon information from all four sensors, and send the divisional scram signals to the hardware logic panel, which performs a 2/4 division vote on whether or not to initiate a reactor scram. Each chassis makes a divisional scram decision based on data from all sensors. Each division performs independently of the others (asynchronous operation). All communications between the divisions are asynchronous. Each chassis substitutes its own spare sensor reading in the 2/3 vote if a sensor reading from one of the other chassis is faulty or missing. Therefore the presence of at least two valid sensor readings in excess of a set point is required before terminating the output to the hardware logic of a scram inhibition signal even when one of the four sensors is faulty or when one of the divisions is out of service. 16 figs

  4. Design and construction of reactor containment systems of the prototype fast breeder reactor MONJU

    International Nuclear Information System (INIS)

    Ikeda, Makinori; Kawata, Koji; Sato, Masaki; Ito, Masashi; Hayashi, Kazutoshi; Kunishima, Shigeru.

    1991-01-01

    The MONJU reactor containment systems consist of a reactor containment vessel, reactor cavity walls and cell liners. The reactor containment vessel is strengthened by ring stiffeners for earthquake stresses. To verify its earthquake-resistant strength, vibration and buckling tests were carried out by using 1/19 scale models. The reactor cavity walls, which form biological shield and support the reactor vessel, are constructed of steel plate frames filled with concrete. The cell liner consists of liner plates and thermal insulation to moderate the effects of sodium spills, and forms a gastight cell to maintain a nitrogen atmosphere. (author)

  5. Passive Decay Heat Removal System for Micro Modular Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Moon, Jangsik; Lee, Jeong Ik; Jeong, Yong Hoon [KAIST, Daejeon (Korea, Republic of)

    2015-10-15

    Dry cooling system is applied as waste heat removal system therefore it is able to consider wide construction site. Schematic figure of the reactor is shown in Fig. 1. In safety features, the reactor has double containment and passive decay heat removal (PDHR) system. The double containment prevents leakage from reactor coolant system to be emitted into environment. The passive decay heat removal system copes with design basis accidents (DBAs). Micros Modular Reactor (MMR) which has been being developed in KAIST is S-CO{sub 2} gas cooled reactor and shows many advantages. The S-CO{sub 2} power cycle reduces size of compressor, and it makes small size of power plant enough to be transported by trailer.The passive residual heat removal system is designed and thermal hydraulic (TH) analysis on coolant system is accomplished. In this research, the design process and TH analysis results are presented. PDHR system is designed for MMR and coolant system with the PDHR system is analyzed by MARS-KS code. Conservative assumptions are applied and the results show that PDHR system keeps coolant system under the design limitation.

  6. Reliability and diagnostic of modular systems

    Directory of Open Access Journals (Sweden)

    J. Kohlas

    2014-01-01

    Full Text Available Reliability and diagnostic are in general two problems discussed separately. Yet the two problems are in fact closely related to each other. Here, this relation is considered in the simple case of modular systems. We show, how the computation of reliability and diagnostic can efficiently be done within the same Bayesian network induced by the modularity of the structure function of the system.

  7. Hybrid case-neural network (CNN) diagnostic system

    International Nuclear Information System (INIS)

    Mohamed, A.H.

    2010-01-01

    recently, the mobile health care has a great attention for the researcher and people all over the world. Case based reasoning (CBR) systems have proved their performance as world wide web (WWW) medical diagnostic systems. They were preferred rather than different reasoning approaches due to their high performance and results' explanation. But, their operations require a complex knowledge acquisition and management processes. On the other hand, it is found that, artificial neural network (ANN) has a great acceptance as a classifier methodology using a little amount of knowledge. But, ANN lacks of an explanation capability .The present research introduces a new web-based hybrid diagnostic system that can use the ANN inside the CBR , cycle.It can provide higher performance for the web diagnostic systems. Besides, the proposed system can be used as a web diagnostic system. It can be applied for diagnosis different types of systems in several domains. It has been applied in diagnosis of the cancer diseases that has a great spreading in recent years as a case of study . However, the suggested system has proved its acceptance in the manner.

  8. An approach to build a knowledge base for reactor accident diagnostic expert system

    International Nuclear Information System (INIS)

    Yoshida, K.; Fujii, M.; Fujiki, K.; Yokobayashi, M.; Kohsaka, A.; Aoyagi, T.; Hirota, Y.

    1987-01-01

    In the development of a rule based expert system, one of the key issues is how to acquire knowledge and to build knowledge base (KB). On building the KB of DISKET, which is an expert system for nuclear reactor accident diagnosis developed in JAERI, several problems have been experienced as follows. To write rules is a time consuming task, and it is difficult to keep the objectivity and consistency of rules as the number of rules increase. Further, certainty factors (CFs) must be often determined according to engineering judgment, i.e., empirically or intuitively. A systematic approach was attempted to handle these difficulties and to build an objective KB efficiently. The approach described in this paper is based on the concept that a prototype KB, colloquially speaking an initial guess, should first be generated in a systematic way and then is to be modified and/or improved by human experts for practical use. Statistical methods, principally Factor Analysis, were used as the systematic way to build a prototype KB for the DISKET using a PWR plant simulator data. The source information is a number of data obtained from the simulation of transients, such as the status of components and annunciator etc., and major process parameters like pressures, temperatures and so on

  9. Emergency core cooling system in BWR type reactors

    International Nuclear Information System (INIS)

    Takizawa, Yoji

    1981-01-01

    Purpose: To rapidly recover the water level in the reactor upon occurrence of slight leakages in the reactor coolant pressure boundary, by promoting the depressurization in the reactor to thereby rapidly increase the high pressure core spray flow rate. Constitution: Upon occurrence of reactor water level reduction, a reactor isolation cooling system and a high pressure core spray system are actuated to start the injection of coolants into a reactor pressure vessel. In this case, if the isolation cooling system is failed to decrease the flow rate in a return pipeway, flow rate indicators show a lower value as compared with a predetermined value. The control device detects it and further confirms the rotation of a high pressure spray pump to open a valve. By the above operation, coolants pumped by the high pressure spray pump is flown by way of a communication pipeway to the return pipeway and sprayed from the top of the pressure vessel. This allows the vapors on the water surface in the pressure vessel to be cooled rapidly and increases the depressurization effects. (Horiuchi, T.)

  10. New measuring and protection system at VR-1 training reactor

    International Nuclear Information System (INIS)

    Kropik, M.; Jurickova, M.

    2006-01-01

    The contribution describes the new measuring and protection system of the VR-1 training reactor. The measuring and protection system upgrade is an integral part of the reactor I and C upgrade. The new measuring and protection system of the VR-1 reactor consists of the operational power measuring and the independent power protection systems. Both systems measure the reactor power and power rate, initiate safety action if safety limits are exceeded and send data (power, power rate, status, etc.) to the reactor control system. The operational power measuring system is a full power range system that receives signal from a fission chamber. The signal is evaluated according to the reactor power either in the pulse or current mode. The current mode utilizes the DC current and Campbell techniques. The new independent power protection system operates in the two highest reactor power decades. It receives signals from a boron chamber and evaluates it in the pulse mode. Both systems are computer based. The operational power measuring and independent power protection systems are diverse - different types and location of chambers, completely different hardware, software algorithms for the power and power rate calculations, software development tools and teems for the software manufacturing. (author)

  11. Radiation effects in IFMIF Li target diagnostic systems

    International Nuclear Information System (INIS)

    Molla, J.; Vila, R.; Shikama, T.; Horiike, H.; Simakov, S.; Ciotti, M.; Ibarra, A.

    2009-01-01

    Diagnostics for the lithium target will be crucial for the operation of IFMIF. Several parameters as the lithium temperature, target thickness or wave pattern must be monitored during operation. Radiation effects may produce malfunctioning in any of these diagnostics due to the exposure to high radiation fields. The main diagnostic systems proposed for the operation of IFMIF are reviewed in this paper from the point of view of radiation damage. The main tools for the assessment of the performance of these diagnostics are the neutronics calculations by using specialised codes and the information accumulated during the last decades on the radiation effects in functional materials, components and diagnostics for ITER. This analysis allows to conclude that the design of some of the diagnostic systems must be revised to assure the high availability required for the target system.

  12. Emergency reactor cooling systems for the experimental VHTR

    International Nuclear Information System (INIS)

    Mitake, Susumu; Suzuki, Katsuo; Miyamoto, Yoshiaki; Tamura, Kazuo; Ezaki, Masahiro.

    1983-03-01

    Performances and design of the panel cooling system which has been proposed to be equipped as an emergency reactor cooling system for the experimental multi purpose very high temperature gas-cooled reactor are explained. Effects of natural circulation flow which would develop in the core and temperature transients of the panel in starting have been precisely investigated. Conditions and procedures for settling accidents with the proposed panel cooling system have been also studied. Based on these studies, it has been shown that the panel cooling system is effective and useful for the emergency reactor cooling of the experimental VHTR. (author)

  13. 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)

  14. Fuel management of mixed reactor type power plant systems

    International Nuclear Information System (INIS)

    Csom, Gyula

    1988-01-01

    In equilibrium symbiotic power plant system containing both thermal reactors and fast breeders, excess plutonium produced by the fast breeders is used to enrich the fuel of the thermal reactors. In plutonium deficient symbiotic power plant system plutonium is supplied both by thermal plants and fast breeders. Mathematical models were constructed and different equations solved to characterize the fuel utilization of both systems if they contain only a single thermal type and a single fast type reactor. The more plutonium is produced in the system, the higher output ratio of thermal to fast reactors is achieved in equilibrium symbiotic power plant system. Mathematical equations were derived to calculate the doubling time and the breeding gain of the equilibrium symbiotic system. (V.N.) 2 figs.; 2 tabs

  15. MFTF plasma diagnostics data acquisition system

    International Nuclear Information System (INIS)

    Davis, G.E.; Coffield, F.E.

    1979-01-01

    The initial goal of the Data Acquisition System (DAS) is to control 11 instruments chosen as the startup diagnostic set and to collect, process, and display the data that these instruments produce. These instruments are described in a paper by Stan Thomas, et. al. entitled ''MFTF Plasma Diagnostics System.'' The DAS must be modular and flexible enough to allow upgrades in the quantity of data taken by an instrument, and also to allow new instruments to be added to the system. This is particularly necessary to support a research project where needs and requirements may change rapidly as a result of experimental findings. Typically, the startup configuration of the diagnostic instruments will contain only a fraction of the planned detectors, and produce approximately one half the data that the expanded version is designed to generate. Expansion of the system will occur in fiscal year 1982

  16. Residual heat removal system diagnostic advisor

    International Nuclear Information System (INIS)

    Tripp, L.

    1991-01-01

    This paper reports on the Residual Heat Removal System (RHRS) Diagnostic Advisor which is an expert system designed to alert the operators to abnormal conditions that exits in the RHRS and offer advice about the cause of the abnormal conditions. The Advisor uses a combination of rule-based and model-based diagnostic techniques to perform its functions. This diagnostic approach leads to a deeper understanding of the RHRS by the Advisor and consequently makes it more robust to unexpected conditions. The main window of the interactive graphic display is a schematic diagram of the RHRS piping system. When a conclusion about a failed component can be reached, the operator can bring up windows that describe the failure mode of the component and a brief explanation about how the Advisor arrived at its conclusion

  17. Steam generating system in LMFBR type reactors

    International Nuclear Information System (INIS)

    Kurosawa, Katsutoshi.

    1984-01-01

    Purpose: To suppress the thermal shock loads to the structures of reactor system and secondary coolant system, for instance, upon plant trip accompanying turbine trip in the steam generation system of LMFBR type reactors. Constitution: Additional feedwater heater is disposed to the pipeway at the inlet of a steam generator in a steam generation system equipped with a closed loop extended from a steam generator by way of a gas-liquid separator, a turbine and a condensator to the steam generator. The separated water at high temperature and high pressure from a gas-liquid separator is heat exchanged with coolants flowing through the closed loop of the steam generation system in non-contact manner and, thereafter, introduced to a water reservoir tank. This can avoid the water to be fed at low temperature as it is to the steam generator, whereby the thermal shock loads to the structures of the reactor system and the secondary coolant system can be suppressed. (Moriyama, K.)

  18. 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)

  19. Event tree analysis for the system of hybrid reactor

    International Nuclear Information System (INIS)

    Yang Yongwei; Qiu Lijian

    1993-01-01

    The application of probabilistic risk assessment for fusion-fission hybrid reactor is introduced. A hybrid reactor system has been analysed using event trees. According to the character of the conceptual design of Hefei Fusion-fission Experimental Hybrid Breeding Reactor, the probabilities of the event tree series induced by 4 typical initiating events were calculated. The results showed that the conceptual design is safe and reasonable. through this paper, the safety character of hybrid reactor system has been understood more deeply. Some suggestions valuable to safety design for hybrid reactor have been proposed

  20. NPP Mochovce units 1 and 2 diagnostic systems

    International Nuclear Information System (INIS)

    Heidenreich, S.

    1997-01-01

    In this paper the diagnostic systems (leak detection monitoring, vibration monitoring, lose parts monitoring, fatigue monitoring) of NPP Mochovce units 1 and 2 are presented. All of the designed diagnostic systems are personal computer based systems

  1. Mechatronics in design of monitoring and diagnostic systems

    Energy Technology Data Exchange (ETDEWEB)

    Uhl, T.; Barszcz, T. [Univ. of Mining and Metallurgy, Krakow (Poland); Hanc, A. [Energocontrol Ltd., Krakow (Poland)

    2003-07-01

    Nowadays development of computer engineering in area of hardware and software gives new possibilities of monitoring and diagnostics system design. The paper presents analysis of new possible solutions for design of monitoring and diagnostic systems including; smart sensor design, modular software design and communication modules. New concept of monitoring system based on home page server solution (nano-server) is presented. Smart sensor design concept with embedded hardware for diagnostic application is shown. New software concept for monitoring and diagnostics automation and examples of applications of new design for condition monitoring based on proposed solution are carefully discussed. (orig.)

  2. Fuel assembly transfer and storage system for nuclear reactors

    International Nuclear Information System (INIS)

    Allain, Albert; Thomas, Claude.

    1981-01-01

    Transfer and storage system on a site comprising several reactors and at least one building housing the installations common to all these reactors. The system includes: transfer and storage modules for the fuel assemblies comprising a containment capable of containing several assemblies carried on a transport vehicle, a set of tracks for the modules between the reactors and the common installations, handling facilities associated with each reactor for moving the irradiated assemblies from the reactor to a transfer module placed in loading position on a track serving the reactor and conversely to move the new assemblies from the transfer module to the reactor, and at least one handling facility located in the common installation building for loading the modules with new assemblies [fr

  3. Imaging systems for medical diagnostics

    International Nuclear Information System (INIS)

    Krestel, E.

    1990-01-01

    This book provides physicians and clinical physicists with detailed information on today's imaging modalities and assists them in selecting the optimal system for each clinical application. Physicists, engineers and computer specialists engaged in research and development and sales departments will also find this book to be of considerable use. It may also be employed at universities, training centers and in technical seminars. The physiological and physical fundamentals are explained in part 1. The technical solutions contained in part 2 illustrate the numerous possibilities available in X-ray diagnostics, computed tomography, nuclear medical diagnostics, magnetic resonance imaging, sonography and biomagnetic diagnostics. (orig.)

  4. Part I. Fuel-motion diagnostics in support of fast-reactor safety experiments. Part II. Fission product detection system in support of fast reactor safety experiments

    International Nuclear Information System (INIS)

    Devolpi, A.; Doerner, R.C.; Fink, C.L.; Regis, J.P.; Rhodes, E.A.; Stanford, G.S.; Braid, T.H.; Boyar, R.E.

    1986-05-01

    In all destructive fast-reactor safety experiments at TREAT, fuel motion and cladding failure have been monitored by the fast-neutron/gamma-ray hodoscope, providing experimental results that are directly applicable to design, modeling, and validation in fast-reactor safety. Hodoscope contributions to the safety program can be considered to fall into several groupings: pre-failure fuel motion, cladding failure, post-failure fuel motion, steel blockages, pretest and posttest radiography, axial-power-profile variations, and power-coupling monitoring. High-quality results in fuel motion have been achieved, and motion sequences have been reconstructed in qualitative and quantitative visual forms. A collimated detection system has been used to observe fission products in the upper regions of a test loop in the TREAT reactor. Particular regions of the loop are targeted through any of five channels in a rotatable assembly in a horizontal hole through the biological shield. A well-type neutron detector, optimized for delayed neutrons, and two GeLi gamma ray spectrometers have been used in several experiments. Data are presented showing a time history of the transport of Dn emitters, of gamma spectra identifying volatile fission products deposited as aerosols, and of fission gas isotopes released from the coolant

  5. Methodology to monitor and diagnostic vibrations of the motor-pumps used in the primary cooling system of IEAR-1 nuclear research reactor

    International Nuclear Information System (INIS)

    Benevenuti, Erion de Lima

    2004-01-01

    The objectives of this study are to establish a strategy to monitor and diagnose vibrations of the motor pumps used in the primary reactor cooling system of the IEA-R1 nuclear research reactor, to verify the possibility of using the existing installed monitoring vibration system and to implement such strategy in a continuous way. Four types of mechanical problems were considered: unbalancing, misalignment, gaps and faults in bearings. An adequate set of analysis tools, well established by the industry, was selected. These are: global measurements of vibration, velocity spectrum and acceleration envelope spectrum. Three sources of data and information were used; the data measured from the primary pumps, experimental results obtained with a Spectra Quest machine used to simulate mechanical defects and data from the literature. The results show that, for the specific case of the motor-pumps of IEA-R1 nuclear research reactor, although the technique using the envelope of acceleration, which is not available in the current system used to monitor the vibration of the motor pumps, is the one with best performance, the other techniques available in the system are sufficient to monitor the four types of mechanical problems mentioned. The proposed strategy is shown and detailed in this work. (author)

  6. An easy-to-use diagnostic system development shell

    Science.gov (United States)

    Tsai, L. C.; Ross, J. B.; Han, C. Y.; Wee, W. G.

    1987-01-01

    The Diagnostic System Development Shell (DSDS), an expert system development shell for diagnostic systems, is described. The major objective of building the DSDS is to create a very easy to use and friendly environment for knowledge engineers and end-users. The DSDS is written in OPS5 and CommonLisp. It runs on a VAX/VMS system. A set of domain independent, generalized rules is built in the DSDS, so the users need not be concerned about building the rules. The facts are explicitly represented in a unified format. A powerful check facility which helps the user to check the errors in the created knowledge bases is provided. A judgement facility and other useful facilities are also available. A diagnostic system based on the DSDS system is question driven and can call or be called by other knowledge based systems written in OPS5 and CommonLisp. A prototype diagnostic system for diagnosing a Philips constant potential X-ray system has been built using the DSDS.

  7. A Diagnostic Ultrasound Imaging System

    International Nuclear Information System (INIS)

    Lee, Seong Woo

    1999-01-01

    The ability to see the internal organs of the human body in a noninvasive way is a powerful diagnostic tool of modern medicine. Among these imaging modalities such as X-ray, MRI, and ultrasound. MRI and ultrasound are presenting much less risk of undesirable damage of both patient and examiner. In fact, no deleterious effects have been reported as a result of clinical examination by using MRI and ultrasound diagnostic equipment. As a result, their market volume has been rapidly increased. MRI has a good resolution. but there are a few disadvantages such as high price. non-real-time imaging capability. and expensive diagnostic cost. On the other hand, the ultrasound imaging system has inherently poor resolution as compared with X-ray and MRI. In spite of its poor resolution, the ultrasound diagnostic equipment is lower in price and has an ability of real-time imaging as compared with the others. As a result, the ultrasound imaging system has become general and essential modality for imaging the internal organs of human body. In this review various researches and developments to enhance the resolution of the ultrasound images are explained and future trends of the ultrasound imaging technology are described

  8. Cooling system for auxiliary reactor component

    International Nuclear Information System (INIS)

    Fujihira, Tomoko.

    1991-01-01

    A cooling system for auxiliary reactor components comprises three systems, that is, two systems of reactor component cooling water systems (RCCW systems) and a high pressure component cooling water system (HPCCW system). Connecting pipelines having partition valves are intervened each in a cooling water supply pipeline to an emmergency component of each of the RCCW systems, a cooling water return pipeline from the emmergency component of each of the RCCW systems, a cooling water supply pipeline to each of the emmergency components of one of the RCCW system and the HPCCW system and a cooling water return pipeline from each of the emmergency components of one of the RCCW system and the HPCCW system. With such constitution, cooling water can be supplied also to the emmergency components in the stand-by system upon periodical inspection or ISI, thereby enabling to improve the backup performance of the emmergency cooling system. (I.N.)

  9. MTX [Microwave Tokamak Experiment] plasma diagnostic system

    International Nuclear Information System (INIS)

    Rice, B.W.; Hooper, E.B.; Brooksby, C.A.

    1987-01-01

    In this paper, a general overview of the MTX plasma diagnostics system is given. This includes a description of the MTX machine configuration and the overall facility layout. The data acquisition system and techniques for diagnostic signal transmission are also discussed. In addition, the diagnostic instruments planned for both an initial ohmic-heating set and a second FEL-heating set are described. The expected range of plasma parameters along with the planned plasma measurements will be reviewed. 7 refs., 5 figs

  10. The IAEA power reactor information system - PRIS

    International Nuclear Information System (INIS)

    Laue, H.J.; Qureshi, A.; Skjoeldebrand, R.; White, D.

    1983-01-01

    The IAEA Power Reactor Information System, PRIS, is based on a collection of basic design data and operating experience data which the IAEA started in 1970. PRIS is used for annual publications on 'Power Reactors in Member States', 'Operating Experience with Nuclear Power Stations in Member States', which gives annual operating information for individual plants, and a 'Performance Analysis Report' summarizing each year's and earlier experience. Since 1973 information has been collected in a systematic manner on significant plant outages (= more than 10 full power hours). There is now information on more than 10,000 outages in the system which permits some conclusions to be drawn both in regard to individual plants and to categories of plants on the significance of different outage reasons and different types of equipment failures. PRIS has not been intended to be a component reliability information system as an international data collection must stop short of the level of detail which would be needed for that purpose. The objectives of PRIS have been to provide a factual background for assumptions on parameters which are essential for economic evaluations and for systems operation planning (load factor and availability). The outage information does, however, lend itself to conclusions about generic problems in different categories of plants and it can be used by an individual operator to find other plants where information about particular problems can be obtained. It would also now be possible to use PRIS for setting availability goals based on experience and not only on theoretical design considerations. The paper demonstrates the conclusions which can be drawn from 662 reactor years of operation of light and heavy water pressurized reactors and 390 reactor years of boiling water reactors and, in particular, the role that the main heat removal system and its components have played in the equipment failure category

  11. Spatially resolved soft x-ray diagnostics in fusion energy research

    International Nuclear Information System (INIS)

    Mlynar, J.; Weinzettl, V.; Imrisek, M.; Loeffelmann, V.

    2013-01-01

    With construction of ITER, the fusion community has progressed into a new stage of research with increased focus on reactor technologies. Corresponding development of diagnostic systems for fusion is required, including research of novel diagnostic methods, validation of radiation hard detectors, and tests of sensors for real-time operation and control, which comprise development of tools for fast data analyses. In parallel, diagnostic systems on running fusion experiments substantially contribute to better understanding of reactor-relevant plasma physics, in particular of energy confinement, plasma stability and transport of impurities. In this respect, spatially resolved Soft X-ray (SXR) diagnostic systems present an interesting case study of development towards reactor-relevant systems. In magnetic confinement fusion research, spatial distribution of SXR radiation with spectral range typically 1 keV - 15 keV is mostly measured by a photosensitive single-row semiconductor elements in a pinhole camera shielded by a beryllium foil. The SXR intensity strongly depends on plasma density, temperature and effective charge, which carry a valuable information on the plasma core physics. Data from SXR diagnostic can be also used for the operation control, among others due to their sensitivity to heavy impurity concentration or to the position of the peak temperature. In order to reconstruct the spatial distribution of SXR plasma emission from the measured line integrated signals, several tomographic methods have been developed and validated. However, the semiconductor elements cannot survive in harsh conditions of future fusion reactors due to radiation damage, which calls for development of radiation hard SXR cameras. In this contribution, role of the SXR diagnostics will be presented in experience and future plans of the Czech tokamak COMPASS (IPP Prague) and the French tokamak TORE SUPRA (CEA Cadarache). In IPP Prague, data from SXR cameras recently contributed to

  12. IAEA data base system for nuclear research reactors (RRDB)

    International Nuclear Information System (INIS)

    Lipscher, P.

    1986-01-01

    The IAEA Data Base System for Nuclear Research Reactors (RRDB) User's Guide is intended for the user who wishes to understand the concepts and operation of the RRDB system. The RRDB is a computerized system recording administrative, operational and technical data on all the nuclear research reactors currently operating, under construction, planned or shut down in IAEA Member States. The data is received by the IAEA from reactor centres on magnetic tapes or as responses to questionnaires. All the data on research, training, test and radioactive isotope production reactors and critical assemblies is stored on the RRDB system. A full set of RRDB programs (in NATURAL) are contained at the back of this Guide

  13. Ageing investigation and upgrading of components/systems of Kartini research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Syarip,; Setiawan, Widi [Yogyakarta Nuclear Research Centre, Yogyakarta (Indonesia)

    1998-10-01

    Kartini research reactor has been operated in good condition and has demonstrated successful operation for the past 18 years, utilized for: reactor kinetic and control studies, instrumentation tests, neutronic and thermohydraulic studies, routine neutron activation analysis, reactor safety studies, training for research reactor operators and supervisors, and reactor physics experiments. Several components of Kartini reactor use components from the abandoned IRT-2000 Project at Serpong and from Bandung Reactor Centre such as: reactor tank, reactor core, heat exchanger, motor blower for ventilation system, fuel elements, etc. To maintain a good operating performance and also for aging investigation purposes, the component failure data collection has been done. The method used is based on the Manual on Reliability Data Collection For Research Reactor PSAs, IAEA TECDOC 636, and analyzed by using Data Entry System (DES) computer code. Analysis result shows that the components/systems failure rate of Kartini reactor is around 1,5.10{sup -4} up to 2,8.10{sup -4} per hour, these values are within the ranges of the values indicated in IAEA TECDOC 478. Whereas from the analysis of irradiation history shows that the neutron fluence of fuel element with highest burn-up (2,05 gram U-235 in average) is around 1.04.10{sup 16} n Cm{sup -2} and this value is still far below its limiting value. Some reactor components/systems have been replaced and upgraded such as heat exchanger, instrumentation and control system (ICS), etc. The new reactor ICS was installed in 1994 which is designed as a distributed structure by using microprocessor based systems and bus system technology. The characteristic and operating performance of the new reactor ICS, as well as the operation history and improvement of the Kartini research reactor is presented. (J.P.N.)

  14. Survey of thorium utilization in power reactor systems

    International Nuclear Information System (INIS)

    Schwartz, M.H.; Schleifer, P.; Dahlberg, R.C.

    1976-01-01

    It is clear that thorium-fueled thermal power reactor systems based on current technology can play a vital role in serving present and long-term energy needs. Advanced thorium converters and thermal breeders can provide an expanded resource base from which the world's growing energy demands can be met. Utilization of a symbiotic system of fast breeders and thorium-fueled thermal reactors can be particularly effective in providing low cost power while conserving uranium resources. Breeder reactors are characterized by high capital costs and very low fuel costs since they produce more fuel than they consume. This excess fuel can be used to fuel thermal converter reactors whose capital costs are low. This symbiosis is optimized when 233 U is bred in the fast breeders and then used to fuel high-conversion-ratio thermal converter reactors operating on the thorium-uranium fuel cycle. The thorium-cycle HTGR, after undergoing more than fifteen years of development in both the United States and Europe, provides for the optimum utilization of our limited uranium resources. Other thermal reactor systems, previously operating on the uranium cycle, also show potential in their capability to utilize the thorium cycle effectively

  15. Reactor inventory monitoring system for Angra-1 reactor

    International Nuclear Information System (INIS)

    S Neto, Joaquim A.; Silva, Marcos C.; Pinheiro, Ronaldo F.M.; Soares, Milton; Martinez, Aquilino; Comerlato, Cesar A.; Oliveira, Eugenio A.

    1996-01-01

    This work describes the project of Reactor Inventory Monitoring System, which will be installed in Angra I Nuclear Power Plant. The inventory information is important to the operators take corrective actions in case of an incident that may cause a failure in the core cooling. (author)

  16. Molten-salt reactor information system

    International Nuclear Information System (INIS)

    Haubenreich, P.N.; Cardwell, D.W.; Engel, J.R.

    1975-06-01

    The Molten-Salt Reactor Information System (MSRIS) is a computer-based file of abstracts of documents dealing with the technology of molten-salt reactors. The file is stored in the IBM-360 system at ORNL, and may be searched through the use of established interactive computer programs from remote terminals connected to the computer via telephone lines. The system currently contains 373 entries and is subject to updating and expansion as additional information is developed. The nature and general content of the data file, a general approach for obtaining information from it, and the manner in which material is added to the file are described. Appendixes provide the list of keywords currently in use, the subject categories under which information is filed, and simplified procedures for searching the file from remote terminals. (U.S.)

  17. Electrical system regulations of the IEA-R1 reactor

    International Nuclear Information System (INIS)

    Mello, Jose Roberto de; Madi Filho, Tufic

    2013-01-01

    The IEA-R1 reactor of the Nuclear and Energy Research Institute (IPEN-CNEN/SP), is a research reactor open pool type, designed and built by the U.S. firm Babcock and Wilcox, having, as coolant and moderator, deionized light water and beryllium and graphite, as reflectors. Until about 1988, the reactor safety systems received power from only one source of energy. As an example, it may be cited the control desk that was powered only by the vital electrical system 220V, which, in case the electricity fails, is powered by the generator group: no-break 220V. In the years 1989 and 1990, a reform of the electrical system upgrading to increase the reactor power and, also, to meet the technical standards of the ABNT (Associacao Brasileira de Normas Tecnicas) was carried out. This work has the objective of showing the relationship between the electric power system and the IEA-R1 reactor security. Also, it demonstrates that, should some electrical power interruption occur, during the reactor operation, this occurrence would not start an accident event. (author)

  18. 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

  19. A methodology for performing virtual measurements in a nuclear reactor system

    International Nuclear Information System (INIS)

    Ikonomopoulos, A.; Uhrig, R.E.; Tsoukalas, L.H.

    1992-01-01

    A novel methodology is presented for monitoring nonphysically measurable variables in an experimental nuclear reactor. It is based on the employment of artificial neural networks to generate fuzzy values. Neural networks map spatiotemporal information (in the form of time series) to algebraically defined membership functions. The entire process can be thought of as a virtual measurement. Through such virtual measurements the values of nondirectly monitored parameters with operational significance, e.g., transient-type, valve-position, or performance, can be determined. Generating membership functions is a crucial step in the development and practical utilization of fuzzy reasoning, a computational approach that offers the advantage of describing the state of the system in a condensed, linguistic form, convenient for monitoring, diagnostics, and control algorithms

  20. Development of a system model for advanced small modular reactors.

    Energy Technology Data Exchange (ETDEWEB)

    Lewis, Tom Goslee,; Holschuh, Thomas Vernon,

    2014-01-01

    This report describes a system model that can be used to analyze three advance small modular reactor (SMR) designs through their lifetime. Neutronics of these reactor designs were evaluated using Monte Carlo N-Particle eXtended (MCNPX/6). The system models were developed in Matlab and Simulink. A major thrust of this research was the initial scoping analysis of Sandias concept of a long-life fast reactor (LLFR). The inherent characteristic of this conceptual design is to minimize the change in reactivity over the lifetime of the reactor. This allows the reactor to operate substantially longer at full power than traditional light water reactors (LWRs) or other SMR designs (e.g. high temperature gas reactor (HTGR)). The system model has subroutines for lifetime reactor feedback and operation calculations, thermal hydraulic effects, load demand changes and a simplified SCO2 Brayton cycle for power conversion.

  1. Supervisory Control and Diagnostics System Distributed Operating System

    International Nuclear Information System (INIS)

    McGoldrick, P.R.

    1979-01-01

    This paper contains a description of the Supervisory Control and Diagnostics System (SCDS) Distributed Operating System. The SCDS consists of nine 32-bit minicomputers with shared memory. The system's main purpose is to control a large Mirror Fusion Test Facility

  2. Reactor protection system including engineered features actuation system

    International Nuclear Information System (INIS)

    Palmaers, W.

    1982-01-01

    The safety concept requires to ensure that - the reactor protection system - the active engineered safeguard - and the necessary auxiliary systems are so designed and interfaced in respect of design and mode of action that, in the event of single component failure reliable control of the consequences of accidents remains ensured at all times and that the availability of the power plant is not limited unnecessarily. In order to satisfy these requirements due, importance was attached to a consistent spacial separation of the mutually redundant subsystems of the active safety equipment. The design and layout of the reactor protection system, of the power supply (emergency power supply), and of the auxiliary systems important from the safety engineering point of view, are such that their subsystems also largely satisfy the requirements of independence and spacial separation. (orig./RW)

  3. Stack Monitoring System At PUSPATI TRIGA Reactor

    International Nuclear Information System (INIS)

    Zamrul Faizad Omar; Mohd Sabri Minhat; Zareen Khan Abdul Jalil Khan; Ridzuan Abdul Mutalib; Khairulezwan Abdul Manan; Nurfarhana Ayuni Joha; Izhar Abu Hussin

    2014-01-01

    This paper describes the current Stack Monitoring System at PUSPATI TRIGA Reactor (RTP) building. A stack monitoring system is a continuous air monitor placed at the reactor top for monitoring the presence of radioactive gaseous in the effluent air from the RTP building. The system consists of four detectors that provide the reading for background, particulate, Iodine and Noble gas. There is a plan to replace the current system due to frequent fault of the system, thus thorough understanding of the current system is required. Overview of the whole system will be explained in this paper. Some current results would be displayed and moving forward brief plan would be mentioned. (author)

  4. 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

  5. Experimental evaluation of an expert system for nuclear reactor operators

    International Nuclear Information System (INIS)

    Nelson, W.R.

    1984-10-01

    The United States Nuclear Regulatory Commission (USNRC) is supporting a program for the experimental evaluation of an expert system for nuclear reactor operators. A prototype expert system, called the Response Tree System, has been developed and implemented at INEL. The Response Tree System is designed to assess the status of a reactor system following an accident and recommend corrective actions to reactor operators. The system is implemented using color graphic displays and is driven by a computer simulation of the reactor system. Control of the system is accomplished using a transparent touch panel. Controlled experiments are being conducted to measure performance differences between operators using the Response Tree System and those not using it to respond to simulated accident situations. This paper summarizes the methodology and results of the evaluation of the Response Tree System, including the quantitative results obtained in the experiments thus far. Design features of the Response Tree System are discussed, and general conclusions regarding the applicability of expert systems in reactor control rooms are presented

  6. Systemization of Design and Analysis Technology for Advanced Reactor

    International Nuclear Information System (INIS)

    Kim, Keung Koo; Lee, J.; Zee, S. K.

    2009-01-01

    The present study is performed to establish the base for the license application of the original technology by systemization and enhancement of the technology that is indispensable for the design and analysis of the advanced reactors including integral reactors. Technical reports and topical reports are prepared for this purpose on some important design/analysis methodology; design and analysis computer programs, structural integrity evaluation of main components and structures, digital I and C systems and man-machine interface design. PPS design concept is complemented reflecting typical safety analysis results. And test plans and requirements are developed for the verification of the advanced reactor technology. Moreover, studies are performed to draw up plans to apply to current or advanced power reactors the original technologies or base technologies such as patents, computer programs, test results, design concepts of the systems and components of the advanced reactors. Finally, pending issues are studied of the advanced reactors to improve the economics and technology realization

  7. Tandem Mirror Reactor Systems Code (Version I)

    International Nuclear Information System (INIS)

    Reid, R.L.; Finn, P.A.; Gohar, M.Y.

    1985-09-01

    A computer code was developed to model a Tandem Mirror Reactor. Ths is the first Tandem Mirror Reactor model to couple, in detail, the highly linked physics, magnetics, and neutronic analysis into a single code. This report describes the code architecture, provides a summary description of the modules comprising the code, and includes an example execution of the Tandem Mirror Reactor Systems Code. Results from this code for two sensitivity studies are also included. These studies are: (1) to determine the impact of center cell plasma radius, length, and ion temperature on reactor cost and performance at constant fusion power; and (2) to determine the impact of reactor power level on cost

  8. Additional reactor protection system of RBMK-1500

    International Nuclear Information System (INIS)

    1999-01-01

    Analysis of anticipated transients without scram of RBMK-1500 reactor showed that additional reactor protection system is required. Data of accident analysis in the case of loose of external electric power and loose of vacuum in condensers of turbines are provided

  9. ADAM: An Accident Diagnostic,Analysis and Management System - Applications to Severe Accident Simulation and Management

    International Nuclear Information System (INIS)

    Zavisca, M.J.; Khatib-Rahbar, M.; Esmaili, H.; Schulz, R.

    2002-01-01

    The Accident Diagnostic, Analysis and Management (ADAM) computer code has been developed as a tool for on-line applications to accident diagnostics, simulation, management and training. ADAM's severe accident simulation capabilities incorporate a balance of mechanistic, phenomenologically based models with simple parametric approaches for elements including (but not limited to) thermal hydraulics; heat transfer; fuel heatup, meltdown, and relocation; fission product release and transport; combustible gas generation and combustion; and core-concrete interaction. The overall model is defined by a relatively coarse spatial nodalization of the reactor coolant and containment systems and is advanced explicitly in time. The result is to enable much faster than real time (i.e., 100 to 1000 times faster than real time on a personal computer) applications to on-line investigations and/or accident management training. Other features of the simulation module include provision for activation of water injection, including the Engineered Safety Features, as well as other mechanisms for the assessment of accident management and recovery strategies and the evaluation of PSA success criteria. The accident diagnostics module of ADAM uses on-line access to selected plant parameters (as measured by plant sensors) to compute the thermodynamic state of the plant, and to predict various margins to safety (e.g., times to pressure vessel saturation and steam generator dryout). Rule-based logic is employed to classify the measured data as belonging to one of a number of likely scenarios based on symptoms, and a number of 'alarms' are generated to signal the state of the reactor and containment. This paper will address the features and limitations of ADAM with particular focus on accident simulation and management. (authors)

  10. Metrology/viewing system for next generation fusion reactors

    International Nuclear Information System (INIS)

    Spampinato, P.T.; Barry, R.E.; Chesser, J.B.; Menon, M.M.; Dagher, M.A.

    1997-01-01

    Next generation fusion reactors require accurate measuring systems to verify sub-millimeter alignment of plasma-facing components in the reactor vessel. A metrology system capable of achieving such accuracy must be compatible with the vessel environment of high gamma radiation, high vacuum, elevated temperature, and magnetic field. This environment requires that the system must be remotely deployed. A coherent, frequency modulated laser radar system is being integrated with a remotely operated deployment system to meet these requirements. The metrology/viewing system consists of a compact laser transceiver optics module which is linked through fiber optics to the laser source and imaging units that are located outside of the harsh environment. The deployment mechanism is a telescopic-mast positioning system. This paper identifies the requirements for the International Thermonuclear Experimental Reactor metrology and viewing system, and describes a remotely operated precision ranging and surface mapping system

  11. Boiling water reactor liquid radioactive waste processing system

    International Nuclear Information System (INIS)

    Anon.

    1977-01-01

    The standard sets forth minimum design, construction and performance requirements with due consideration for operation of the liquid radioactive waste processing system for boiling water reactor plants for routine operation including design basis fuel leakage and design basis occurrences. For the purpose of this standard, the liquid radioactive waste processing system begins at the interfaces with the reactor coolant pressure boundary, at the interface valve(s) in lines from other systems and at those sumps and floor drains provided for liquid waste with the potential of containing radioactive material. The system terminates at the point of controlled discharge to the environment, at the point of interface with the waste solidification system and at the point of recycle back to storage for reuse. The standard does not include the reactor coolant clean-up system, fuel pool clean-up system, sanitary waste system, any nonaqueous liquid system or controlled area storm drains

  12. Development of toroid-type HTS DC reactor series for HVDC system

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kwangmin, E-mail: kwangmin81@gmail.com [Changwon National University, 55306 Sarim-dong, Changwon 641-773 (Korea, Republic of); Go, Byeong-Soo; Park, Hea-chul; Kim, Sung-kyu; Kim, Seokho [Changwon National University, 55306 Sarim-dong, Changwon 641-773 (Korea, Republic of); Lee, Sangjin [Uiduk University, Gyeongju 780-713 (Korea, Republic of); Oh, Yunsang [Vector Fields Korea Inc., Pohang 790-834 (Korea, Republic of); Park, Minwon; Yu, In-Keun [Changwon National University, 55306 Sarim-dong, Changwon 641-773 (Korea, Republic of)

    2015-11-15

    Highlights: • The authors developed the 400 mH, 400 A class toroid-type HTS DC reactor system. • The target temperature, inductance and operating current are under 20 K at magnet, 400 mH and 400 A, respectively. All target performances of the HTS DC reactor were achieved. • The HTS DC reactor was conducted through the interconnection operation with a LCC type HVDC system. • Now, the authors are studying the 400 mH, 1500 A class toroid-type HTS DC reactor for the next phase HTS DC reactor. - Abstract: This paper describes design specifications and performance of a toroid-type high-temperature superconducting (HTS) DC reactor. The first phase operation targets of the HTS DC reactor were 400 mH and 400 A. The authors have developed a real HTS DC reactor system during the last three years. The HTS DC reactor was designed using 2G GdBCO HTS wires. The HTS coils of the toroid-type DC reactor magnet were made in the form of a D-shape. The electromagnetic performance of the toroid-type HTS DC reactor magnet was analyzed using the finite element method program. A conduction cooling method was adopted for reactor magnet cooling. The total system has been successfully developed and tested in connection with LCC type HVDC system. Now, the authors are studying a 400 mH, kA class toroid-type HTS DC reactor for the next phase research. The 1500 A class DC reactor system was designed using layered 13 mm GdBCO 2G HTS wire. The expected operating temperature is under 30 K. These fundamental data obtained through both works will usefully be applied to design a real toroid-type HTS DC reactor for grid application.

  13. Development of toroid-type HTS DC reactor series for HVDC system

    International Nuclear Information System (INIS)

    Kim, Kwangmin; Go, Byeong-Soo; Park, Hea-chul; Kim, Sung-kyu; Kim, Seokho; Lee, Sangjin; Oh, Yunsang; Park, Minwon; Yu, In-Keun

    2015-01-01

    Highlights: • The authors developed the 400 mH, 400 A class toroid-type HTS DC reactor system. • The target temperature, inductance and operating current are under 20 K at magnet, 400 mH and 400 A, respectively. All target performances of the HTS DC reactor were achieved. • The HTS DC reactor was conducted through the interconnection operation with a LCC type HVDC system. • Now, the authors are studying the 400 mH, 1500 A class toroid-type HTS DC reactor for the next phase HTS DC reactor. - Abstract: This paper describes design specifications and performance of a toroid-type high-temperature superconducting (HTS) DC reactor. The first phase operation targets of the HTS DC reactor were 400 mH and 400 A. The authors have developed a real HTS DC reactor system during the last three years. The HTS DC reactor was designed using 2G GdBCO HTS wires. The HTS coils of the toroid-type DC reactor magnet were made in the form of a D-shape. The electromagnetic performance of the toroid-type HTS DC reactor magnet was analyzed using the finite element method program. A conduction cooling method was adopted for reactor magnet cooling. The total system has been successfully developed and tested in connection with LCC type HVDC system. Now, the authors are studying a 400 mH, kA class toroid-type HTS DC reactor for the next phase research. The 1500 A class DC reactor system was designed using layered 13 mm GdBCO 2G HTS wire. The expected operating temperature is under 30 K. These fundamental data obtained through both works will usefully be applied to design a real toroid-type HTS DC reactor for grid application.

  14. 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

  15. 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.)

  16. Advanced robotic remote handling system for reactor dismantlement

    International Nuclear Information System (INIS)

    Shinohara, Yoshikuni; Usui, Hozumi; Fujii, Yoshio

    1991-01-01

    An advanced robotic remote handling system equipped with a multi-functional amphibious manipulator has been developed and used to dismantle a portion of radioactive reactor internals of an experimental boiling water reactor in the program of reactor decommissioning technology development carried out by the Japan Atomic Energy Research Institute. (author)

  17. REACTOR - a Concept for establishing a System-of-Systems

    Science.gov (United States)

    Haener, Rainer; Hammitzsch, Martin; Wächter, Joachim

    2014-05-01

    REACTOR is a working title for activities implementing reliable, emergent, adaptive, and concurrent collaboration on the basis of transactional object repositories. It aims at establishing federations of autonomous yet interoperable systems (Systems-of-Systems), which are able to expose emergent behaviour. Following the principles of event-driven service-oriented architectures (SOA 2.0), REACTOR enables adaptive re-organisation by dynamic delegation of responsibilities and novel yet coherent monitoring strategies by combining information from different domains. Thus it allows collaborative decision-processes across system, discipline, and administrative boundaries. Interoperability is based on two approaches that implement interconnection and communication between existing heterogeneous infrastructures and information systems: Coordinated (orchestration-based) communication and publish/subscribe (choreography-based) communication. Choreography-based communication ensures the autonomy of the participating systems to the highest possible degree but requires the implementation of adapters, which provide functional access to information (publishing/consuming events) via a Message Oriented Middleware (MOM). Any interconnection of the systems (composition of service and message cascades) is established on the basis of global conversations that are enacted by choreographies specifying the expected behaviour of the participating systems with respect to agreed Service Level Agreements (SLA) required by e.g. national authorities. The specification of conversations, maintained in commonly available repositories also enables the utilisation of systems for purposes (evolving) other than initially intended. Orchestration-based communication additionally requires a central component that controls the information transfer via service requests or event processing and also takes responsibility of managing business processes. Commonly available transactional object repositories are

  18. Resilient actions in the diagnostic process and system performance.

    Science.gov (United States)

    Smith, Michael W; Davis Giardina, Traber; Murphy, Daniel R; Laxmisan, Archana; Singh, Hardeep

    2013-12-01

    Systemic issues can adversely affect the diagnostic process. Many system-related barriers can be masked by 'resilient' actions of frontline providers (ie, actions supporting the safe delivery of care in the presence of pressures that the system cannot readily adapt to). We explored system barriers and resilient actions of primary care providers (PCPs) in the diagnostic evaluation of cancer. We conducted a secondary data analysis of interviews of PCPs involved in diagnostic evaluation of 29 lung and colorectal cancer cases. Cases covered a range of diagnostic timeliness and were analysed to identify barriers for rapid diagnostic evaluation, and PCPs' actions involving elements of resilience addressing those barriers. We rated these actions according to whether they were usual or extraordinary for typical PCP work. Resilient actions and associated barriers were found in 59% of the cases, in all ranges of timeliness, with 40% involving actions rated as beyond typical. Most of the barriers were related to access to specialty services and coordination with patients. Many of the resilient actions involved using additional communication channels to solicit cooperation from other participants in the diagnostic process. Diagnostic evaluation of cancer involves several resilient actions by PCPs targeted at system deficiencies. PCPs' actions can sometimes mitigate system barriers to diagnosis, and thereby impact the sensitivity of 'downstream' measures (eg, delays) in detecting barriers. While resilient actions might enable providers to mitigate system deficiencies in the short run, they can be resource intensive and potentially unsustainable. They complement, rather than substitute for, structural remedies to improve system performance. Measures to detect and fix system performance issues targeted by these resilient actions could facilitate diagnostic safety.

  19. 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

  20. Reactor coolant purification system circulation pumps (CUW pumps)

    International Nuclear Information System (INIS)

    Tsutsui, Toshiaki

    1979-01-01

    Coolant purification equipments for BWRs have been improved, and the high pressure purifying system has become the main type. The quantity of purifying treatment also changed to 2% of the flow rate of reactor feed water. As for the circulation pumps, canned motor pumps are adopted recently, and the improvements of reliability and safety are attempted. The impurities carried in by reactor feed water and the corrosion products generated in reactors and auxiliary equipments are activated by neutron irradiation or affect heat transfer adversely, adhering to fuel claddings are core structures. Therefore, a part of reactor coolant is led to the purification equipments, and returned to reactors after the impurities are eliminated perfectly. At the time of starting and stopping reactors, excess reactor water and the contaminated water from reactors are transferred to main condenser hot wells or waste treatment systems. Thus the prescribed water quality is maintained. The operational modes of and the requirements for the CUW pumps, the construction and the features of the canned motor type CUW pumps are explained. Recently, a pump operated for 11 months without any maintenance has been disassembled and inspected, but the wear of bearings has not been observed, and the high reliability of the pump has been proved. (Kako, I.)

  1. Modeling and simulation of CANDU reactor and its regulating system

    Science.gov (United States)

    Javidnia, Hooman

    Analytical computer codes are indispensable tools in design, optimization, and control of nuclear power plants. Numerous codes have been developed to perform different types of analyses related to the nuclear power plants. A large number of these codes are designed to perform safety analyses. In the context of safety analyses, the control system is often neglected. Although there are good reasons for such a decision, that does not mean that the study of control systems in the nuclear power plants should be neglected altogether. In this thesis, a proof of concept code is developed as a tool that can be used in the design. optimization. and operation stages of the control system. The main objective in the design of this computer code is providing a tool that is easy to use by its target audience and is capable of producing high fidelity results that can be trusted to design the control system and optimize its performance. Since the overall plant control system covers a very wide range of processes, in this thesis the focus has been on one particular module of the the overall plant control system, namely, the reactor regulating system. The center of the reactor regulating system is the CANDU reactor. A nodal model for the reactor is used to represent the spatial neutronic kinetics of the core. The nodal model produces better results compared to the point kinetics model which is often used in the design and analysis of control system for nuclear reactors. The model can capture the spatial effects to some extent. although it is not as detailed as the finite difference methods. The criteria for choosing a nodal model of the core are: (1) the model should provide more detail than point kinetics and capture spatial effects, (2) it should not be too complex or overly detailed to slow down the simulation and provide details that are extraneous or unnecessary for a control engineer. Other than the reactor itself, there are auxiliary models that describe dynamics of different

  2. Implementation of MOAS II diagnosis system at the OECD Halden Reactor project

    International Nuclear Information System (INIS)

    Kim, I.S.; Grini, R.E.; Nilsen, S.

    1995-01-01

    MOAS II is a surveillance and diagnosis system that uses several techniques for knowledge acquisition and diagnostic reasoning, e.g., goal tree-success tree, simplified directed graphs, diagnosis trees, and detailed knowledge of the process, such as mass or energy balance. This new approach was used at the Halden Man-Machine Laboratory of the OECD Halden Reactor Project. The performance of MOAS II, developed in G2 real-time expert system shell for the high-pressure preheaters of the NORS process, was tested against a variety of transient scenarios, including failures of control valves and sensors, and leakage of tubes of the preheaters. These tests showed that MOAS II successfully carried out its intended functions, i.e., quickly recognizing an occurring disturbance, correctly diagnosing its cause, and presenting advice on its control to the operator. The insights gained during the implementation are discussed

  3. Diagnostic Neural Network Systems for the Electronic Circuits

    International Nuclear Information System (INIS)

    Mohamed, A.H.

    2014-01-01

    Neural Networks is one of the most important artificial intelligent approaches for solving the diagnostic processes. This research concerns with uses the neural networks for diagnosis of the electronic circuits. Modern electronic systems contain both the analog and digital circuits. But, diagnosis of the analog circuits suffers from great complexity due to their nonlinearity. To overcome this problem, the proposed system introduces a diagnostic system that uses the neural network to diagnose both the digital and analog circuits. So, it can face the new requirements for the modern electronic systems. A fault dictionary method was implemented in the system. Experimental results are presented on three electronic systems. They are: artificial kidney, wireless network and personal computer systems. The proposed system has improved the performance of the diagnostic systems when applied for these practical cases

  4. Simplified safety and containment systems for the iris reactor

    International Nuclear Information System (INIS)

    Conway, L.E.; Lombardi, C.; Ricotti, M.; Oriani, L.

    2001-01-01

    The IRIS (International Reactor Innovative and Secure) is a 100 - 300 MW modular type pressurized water reactor supported by the U.S. DOE NERI Program. IRIS features a long-life core to provide proliferation resistance and to reduce the volume of spent fuel, as well as reduce maintenance requirements. IRIS utilizes an integral reactor vessel that contains all major primary system components. This integral reactor vessel makes it possible to reduce containment size; making the IRIS more cost competitive. IRIS is being designed to enhance reactor safety, and therefore a key aspect of the IRIS program is the development of the safety and containment systems. These systems are being designed to maximize containment integrity, prevent core uncover following postulated accidents, minimize the probability and consequences of severe accidents, and provide a significant simplification over current safety system designs. The design of the IRIS containment and safety systems has been identified and preliminary analyses have been completed. The IRIS safety concept employs some unique features that minimize the consequences of postulated design basis events. This paper will provide a description of the containment design and safety systems, and will summarize the analysis results. (author)

  5. Applications of plasma core reactors to terrestrial energy systems

    International Nuclear Information System (INIS)

    Lantham, T.S.; Biancardi, F.R.; Rodgers, R.J.

    1974-01-01

    Plasma core reactors offer several new options for future energy needs in addition to space power and propulsion applications. Power extraction from plasma core reactors with gaseous nuclear fuel allows operation at temperatures higher than conventional reactors. Highly efficient thermodynamic cycles and applications employing direct coupling of radiant energy are possible. Conceptual configurations of plasma core reactors for terrestrail applications are described. Closed-cycle gas turbines, MHD systems, photo- and thermo-chemical hydrogen production processes, and laser systems using plasma core reactors as prime energy sources are considered. Cycle efficiencies in the range of 50 to 65 percent are calculated for closed-cycle gas turbine and MHD electrical generators. Reactor advantages include continuous fuel reprocessing which limits inventory of radioactive by-products and thorium-U-233 breeder configurations with about 5-year doubling times

  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; Dozimetarski sistem reaktora RB

    Energy Technology Data Exchange (ETDEWEB)

    Lolic, B; Vukadin, D [Boris Kidric Institute of nuclear sciences, Vinca, Belgrade (Yugoslavia)

    1962-07-01

    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. 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

  10. 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.)

  11. Naval application of battery optimized reactor integral system

    International Nuclear Information System (INIS)

    Kim, N. H.; Kim, T. W.; Son, H. M.; Suh, K. Y.

    2007-01-01

    Past civilian N.S. Savanna (80 MW t h), Otto-Hahn (38 MW t h) and Mutsu (36 MW t h) experienced stable operations under various sea conditions to prove that the reactors were stable and suitable for ship power source. Russian nuclear icebreakers such as Lenin (90 MW t h x2), Arukuchika (150 MW t h x2) showed stable operations under severe conditions during navigation on the Arctic Sea. These reactor systems, however, should be made even more efficient, compact, safe and long life, because adding support from the land may not be available on the sea. In order to meet these requirements, a compact, simple, safe and innovative integral system named Naval Application Vessel Integral System (NAVIS) is being designed with such novel concepts as a primary liquid metal coolant, a secondary supercritical carbon dioxide (SCO 2 ) coolant, emergency reactor cooling system, safety containment and so on. NAVIS is powered by Battery Optimized Reactor Integral System (BORIS). An ultra-small, ultra-long-life, versatile-purpose, fast-spectrum reactor named BORIS is being developed for a multi-purpose application such as naval power source, electric power generation in remote areas, seawater desalination, and district heating. NAVIS aims to satisfy special environment on the sea with BORIS using the lead (Pb) coolant in the primary system. NAVIS improves the economical efficiency resorting to the SCO 2 Brayton cycle for the secondary system. BORIS is operated by natural circulation of Pb without needing pumps. The reactor power is autonomously controlled by load-following operation without an active reactivity control system, whereas B 4 C based shutdown control rod is equipped for an emergency condition. SCO 2 promises a high power conversion efficiency of the recompression Brayton cycle due to its excellent compressibility reducing the compression work at the bottom of the cycle and to a higher density than helium or steam decreasing the component size. Therefore, the SCO 2 Brayton

  12. 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.)

  13. A gas-cooled reactor surface power system

    International Nuclear Information System (INIS)

    Lipinski, R.J.; Wright, S.A.; Lenard, R.X.; Harms, G.A.

    1999-01-01

    A human outpost on Mars requires plentiful power to assure survival of the astronauts. Anywhere from 50 to 500 kW of electric power (kWe) will be needed, depending on the number of astronauts, level of scientific activity, and life-cycle closure desired. This paper describes a 250-kWe power system based on a gas-cooled nuclear reactor with a recuperated closed Brayton cycle conversion system. The design draws upon the extensive data and engineering experience developed under the various high-temperature gas cooled reactor programs and under the SP-100 program. The reactor core is similar in power and size to the research reactors found on numerous university campuses. The fuel is uranium nitride clad in Nb1%Zr, which has been extensively tested under the SP-100 program. The fuel rods are arranged in a hexagonal array within a BeO block. The BeO softens the spectrum, allowing better use of the fuel and stabilizing the geometry against deformation during impact or other loadings. The system has a negative temperature feedback coefficient so that the power level will automatically follow a variable load without the need for continuous adjustment of control elements. Waste heat is removed by an air-cooled heat exchanger using cold Martian air. The amount of radioactivity in the reactor at launch is very small (less than a Curie, and about equal to a truckload of uranium ore). The system will need to be engineered so that criticality can not occur for any launch accident. This system is also adaptable for electric propulsion or life-support during transit to and from Mars. copyright 1999 American Institute of Physics

  14. A gas-cooled reactor surface power system

    International Nuclear Information System (INIS)

    Lipinski, Ronald J.; Wright, Steven A.; Lenard, Roger X.; Harms, Gary A.

    1999-01-01

    A human outpost on Mars requires plentiful power to assure survival of the astronauts. Anywhere from 50 to 500 kW of electric power (kWe) will be needed, depending on the number of astronauts, level of scientific activity, and life-cycle closure desired. This paper describes a 250-kWe power system based on a gas-cooled nuclear reactor with a recuperated closed Brayton cycle conversion system. The design draws upon the extensive data and engineering experience developed under the various high-temperature gas cooled reactor programs and under the SP-100 program. The reactor core is similar in power and size to the research reactors found on numerous university campuses. The fuel is uranium nitride clad in Nb1%Zr, which has been extensively tested under the SP-100 program. The fuel rods are arranged in a hexagonal array within a BeO block. The BeO softens the spectrum, allowing better use of the fuel and stabilizing the geometry against deformation during impact or other loadings. The system has a negative temperature feedback coefficient so that the power level will automatically follow a variable load without the need for continuous adjustment of control elements. Waste heat is removed by an air-cooled heat exchanger using cold Martian air. The amount of radioactivity in the reactor at launch is very small (less than a Curie, and about equal to a truckload of uranium ore). The system will need to be engineered so that criticality can not occur for any launch accident. This system is also adaptable for electric propulsion or life-support during transit to and from Mars

  15. A Gas-Cooled Reactor Surface Power System

    Energy Technology Data Exchange (ETDEWEB)

    Harms, G.A.; Lenard, R.X.; Lipinski, R.J.; Wright, S.A.

    1998-11-09

    A human outpost on Mars requires plentiful power to assure survival of the astronauts. Anywhere from 50 to 500 kW of electric power (kWe) will be needed, depending on the number of astronauts, level of scientific activity, and life- cycle closure desired. This paper describes a 250-kWe power system based on a gas-cooled nuclear reactor with a recuperated closed Brayton cycle conversion system. The design draws upon the extensive data and engineering experience developed under the various high-temperature gas cooled reactor programs and under the SP-100 program. The reactor core is similar in power and size to the research reactors found on numerous university campuses. The fuel is uranium nitide clad in Nb 1 %Zr, which has been extensively tested under the SP-I 00 program The fiel rods are arranged in a hexagonal array within a BeO block. The BeO softens the spectrum, allowing better use of the fbel and stabilizing the geometty against deformation during impact or other loadings. The system has a negative temperature feedback coefficient so that the power level will automatically follow a variable load without the need for continuous adjustment of control elements. Waste heat is removed by an air-cooled heat exchanger using cold Martian air. The amount of radioactivity in the reactor at launch is very small (less than a Curie, and about equal to a truckload of uranium ore). The system will need to be engineered so that criticality cannot occur for any launch accident. This system is also adaptable for electric propulsion or life-support during transit to and from Mars.

  16. Microprocessor tester for the treat upgrade reactor trip system

    International Nuclear Information System (INIS)

    Lenkszus, F.R.; Bucher, R.G.

    1984-01-01

    The upgrading of the Transient Reactor Test (TREAT) Facility at ANL-Idaho has been designed to provide additional experimental capabilities for the study of core disruptive accident (CDA) phenomena. In addition, a programmable Automated Reactor Control System (ARCS) will permit high-power transients up to 11,000 MW having a controlled reactor period of from 15 to 0.1 sec. These modifications to the core neutronics will improve simulation of LMFBR accident conditions. Finally, a sophisticated, multiply-redundant safety system, the Reactor Trip System (RTS), will provide safe operation for both steady state and transient production operating modes. To insure that this complex safety system is functioning properly, a Dedicated Microprocessor Tester (DMT) has been implemented to perform a thorough checkout of the RTS prior to all TREAT operations

  17. User's manual for the reactor burnup system, REBUS

    International Nuclear Information System (INIS)

    Olson, A.P.; Regis, J.P.; Meneley, D.A.; Hoover, L.J.

    1972-01-01

    A user's manual for the REBUS System (REactor BUrnup System) is presented. Its primary purpose is to provide sufficient information about the REBUS capability to the user to ensure its efficient utilization. The current REBUS System either solves for the infinite time (equilibrium) operating conditions of a recycle system under fixed conditions, or solves for operating conditions during a single time step (non-equilibrium). The capability of studying various in-reactor fuel management and ex-reactor fuel management schemes has been included. REBUS has been operated with one- and two-dimensional diffusion theory neutronics solutions up to the present time. The model was specifically designed for extension to other neutronics models such as three-dimensional diffusion or transport theory and direct or synthesis solutions

  18. Gas-cooled reactor power systems for space

    International Nuclear Information System (INIS)

    Walter, C.E.

    1987-01-01

    Efficiency and mass characteristics for four gas-cooled reactor power system configurations in the 2- to 20-MWe power range are modeled. The configurations use direct and indirect Brayton cycles with and without regeneration in the power conversion loop. The prismatic ceramic core of the reactor consists of several thousand pencil-shaped tubes made from a homogeneous mixture of moderator and fuel. The heat rejection system is found to be the major contributor to system mass, particularly at high power levels. A direct, regenerated Brayton cycle with helium working fluid permits high efficiency and low specific mass for a 10-MWe system

  19. Nuclear reactor cavity floor passive heat removal system

    Science.gov (United States)

    Edwards, Tyler A.; Neeley, Gary W.; Inman, James B.

    2018-03-06

    A nuclear reactor includes a reactor core disposed in a reactor pressure vessel. A radiological containment contains the nuclear reactor and includes a concrete floor located underneath the nuclear reactor. An ex vessel corium retention system includes flow channels embedded in the concrete floor located underneath the nuclear reactor, an inlet in fluid communication with first ends of the flow channels, and an outlet in fluid communication with second ends of the flow channels. In some embodiments the inlet is in fluid communication with the interior of the radiological containment at a first elevation and the outlet is in fluid communication with the interior of the radiological containment at a second elevation higher than the first elevation. The radiological containment may include a reactor cavity containing a lower portion of the pressure vessel, wherein the concrete floor located underneath the nuclear reactor is the reactor cavity floor.

  20. 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.)

  1. Knowledge acquisition for nuclear power plant unit diagnostic system

    International Nuclear Information System (INIS)

    Li Xiaodong; Xi Shuren

    2003-01-01

    The process of acquiring knowledge and building a knowledge base is critical to realize fault diagnostic system at unit level in a nuclear power plant. It directly determines whether the diagnostic system can be applied eventually in a commercial plant. A means to acquire knowledge and its procedures was presented in this paper for fault diagnostic system in a nuclear power plant. The work can be carried out step by step and it is feasible in a commercial nuclear power plant. The knowledge base of the fault diagnostic system for a nuclear power plant can be built after the staff finish the tasks according to the framework presented in this paper

  2. FISS: a computer program for reactor systems studies

    International Nuclear Information System (INIS)

    Tamm, H.; Sherman, G.R.; Wright, J.H.; Nieman, R.E.

    1979-08-01

    ΣFISSΣ is a computer code for use in investigating alternative fuel cycle strategies for Canadian and world nuclear programs. The code performs a system simulation accounting for dynamic effects of growing nuclear systems. Facilities in the model include storage for irradiated fuel, mines, plants for enrichment, fuel fabrication, fuel reprocessing and heavy water, and reactors. FISS is particularly useful for comparing various reactor strategies and studying sensitivities of resource consumption, capital investment and energy costs with changes in fuel cycle parameters, reactor parameters and financial variables. (author)

  3. Introduction to nuclear power reactors and their health physics systems

    International Nuclear Information System (INIS)

    Brtis, J.S.

    1982-01-01

    This paper provides an introduction to: (1) the major systems of Boiling Water Reactors (BWR's) and Pressurized Water Reactors (PWR's), (2) the production and distribution of radiation sources in BWR's and PWR's, (3) the regulatory and functional requirements for nuclear power reactor design from a health physics standpoint, (4) the health physics systems provided to meet such requirements, and (5) a bibliography of documents germane to power reactor health physics design

  4. Development of toroid-type HTS DC reactor series for HVDC system

    Science.gov (United States)

    Kim, Kwangmin; Go, Byeong-Soo; Park, Hea-chul; Kim, Sung-kyu; Kim, Seokho; Lee, Sangjin; Oh, Yunsang; Park, Minwon; Yu, In-Keun

    2015-11-01

    This paper describes design specifications and performance of a toroid-type high-temperature superconducting (HTS) DC reactor. The first phase operation targets of the HTS DC reactor were 400 mH and 400 A. The authors have developed a real HTS DC reactor system during the last three years. The HTS DC reactor was designed using 2G GdBCO HTS wires. The HTS coils of the toroid-type DC reactor magnet were made in the form of a D-shape. The electromagnetic performance of the toroid-type HTS DC reactor magnet was analyzed using the finite element method program. A conduction cooling method was adopted for reactor magnet cooling. The total system has been successfully developed and tested in connection with LCC type HVDC system. Now, the authors are studying a 400 mH, kA class toroid-type HTS DC reactor for the next phase research. The 1500 A class DC reactor system was designed using layered 13 mm GdBCO 2G HTS wire. The expected operating temperature is under 30 K. These fundamental data obtained through both works will usefully be applied to design a real toroid-type HTS DC reactor for grid application.

  5. 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

  6. Laboratory Information Systems in Molecular Diagnostics: Why Molecular Diagnostics Data are Different.

    Science.gov (United States)

    Lee, Roy E; Henricks, Walter H; Sirintrapun, Sahussapont J

    2016-03-01

    Molecular diagnostic testing presents new challenges to information management that are yet to be sufficiently addressed by currently available information systems for the molecular laboratory. These challenges relate to unique aspects of molecular genetic testing: molecular test ordering, informed consent issues, diverse specimen types that encompass the full breadth of specimens handled by traditional anatomic and clinical pathology information systems, data structures and data elements specific to molecular testing, varied testing workflows and protocols, diverse instrument outputs, unique needs and requirements of molecular test reporting, and nuances related to the dissemination of molecular pathology test reports. By satisfactorily addressing these needs in molecular test data management, a laboratory information system designed for the unique needs of molecular diagnostics presents a compelling reason to migrate away from the current paper and spreadsheet information management that many molecular laboratories currently use. This paper reviews the issues and challenges of information management in the molecular diagnostics laboratory.

  7. 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)

  8. Multivariate statistical pattern recognition system for reactor noise analysis

    International Nuclear Information System (INIS)

    Gonzalez, R.C.; Howington, L.C.; Sides, W.H. Jr.; Kryter, R.C.

    1976-01-01

    A multivariate statistical pattern recognition system for reactor noise analysis was developed. The basis of the system is a transformation for decoupling correlated variables and algorithms for inferring probability density functions. The system is adaptable to a variety of statistical properties of the data, and it has learning, tracking, and updating capabilities. System design emphasizes control of the false-alarm rate. The ability of the system to learn normal patterns of reactor behavior and to recognize deviations from these patterns was evaluated by experiments at the ORNL High-Flux Isotope Reactor (HFIR). Power perturbations of less than 0.1 percent of the mean value in selected frequency ranges were detected by the system

  9. Multivariate statistical pattern recognition system for reactor noise analysis

    International Nuclear Information System (INIS)

    Gonzalez, R.C.; Howington, L.C.; Sides, W.H. Jr.; Kryter, R.C.

    1975-01-01

    A multivariate statistical pattern recognition system for reactor noise analysis was developed. The basis of the system is a transformation for decoupling correlated variables and algorithms for inferring probability density functions. The system is adaptable to a variety of statistical properties of the data, and it has learning, tracking, and updating capabilities. System design emphasizes control of the false-alarm rate. The ability of the system to learn normal patterns of reactor behavior and to recognize deviations from these patterns was evaluated by experiments at the ORNL High-Flux Isotope Reactor (HFIR). Power perturbations of less than 0.1 percent of the mean value in selected frequency ranges were detected by the system. 19 references

  10. 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

  11. 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)

  12. Light and heavy water replacing system in reactor container

    International Nuclear Information System (INIS)

    Miyamoto, Keiji.

    1979-01-01

    Purpose: To enable to determine the strength of a reactor container while neglecting the outer atmospheric pressure upon evacuation, by evacuating the gap between the reactor container and a biological thermal shield, as well as the container simultaneously upon light water - heavy water replacement. Method: Upon replacing light water with heavy water by vacuum evaporation system in a nuclear reactor having a biological thermal shield surrounding the reactor container incorporating therein a reactor core by way of a heat expansion absorbing gap, the reactor container and the havy water recycling system, as well as the inside of heat expansion absorbing gap are evacuated simultaneously. This enables to neglect the outer atmospheric outer pressure upon evacuation in the determination of the container strength, and the thickness of the container can be decreased by so much as the external pressure neglected. (Moriyama, K.)

  13. BWR recirculation pump diagnostic expert system

    International Nuclear Information System (INIS)

    Chiang, S.C.; Morimoto, C.N.; Torres, M.R.

    2004-01-01

    At General Electric (GE), an on-line expert system to support maintenance decisions for BWR recirculation pumps for nuclear power plants has been developed. This diagnostic expert system is an interactive on-line system that furnishes diagnostic information concerning BWR recirculation pump operational problems. It effectively provides the recirculation pump diagnostic expertise in the plant control room continuously 24 hours a day. The expert system is interfaced to an on-line monitoring system, which uses existing plant sensors to acquire non-safety related data in real time. The expert system correlates and evaluates process data and vibration data by applying expert rules to determine the condition of a BWR recirculation pump system by applying knowledge based rules. Any diagnosis will be automatically displayed, indicating which pump may have a problem, the category of the problem, and the degree of concern expressed by the validity index and color hierarchy. The rules incorporate the expert knowledge from various technical sources such as plant experience, engineering principles, and published reports. These rules are installed in IF-THEN formats and the resulting truth values are also expressed in fuzzy terms and a certainty factor called a validity index. This GE Recirculation Pump Expert System uses industry-standard software, hardware, and network access to provide flexible interfaces with other possible data acquisition systems. Gensym G2 Real-Time Expert System is used for the expert shell and provides the graphical user interface, knowledge base, and inference engine capabilities. (author)

  14. Design study of plant system for the fusion experimental reactor (FER)

    International Nuclear Information System (INIS)

    Iida, Hiromasa; Kuroda, Hideo; Yamada, Masao; Suzuki, Tatsushi; Honda, Tsutomu; Ohmura, Hiroshi; Itoh, Shinichi.

    1986-11-01

    This report describes design study results of the FER plant system. The purpose of this study is to have an image of the FER plant system as a whole by designing major auxiliary systems, reactor building and maintenance and radwaste desposal systems. The major auxiliary systems include tritium, cooling, evacuation and fueling systems. For these each systems, flowdiagrams are studied and designs of devices and pipings are conducted. In the reactor building design, layout of the above auxiliary systems in the building is studied with careful zoning concept by the radiation level. Structural integrity of the reactor building is also studied including seismic analysis. In the design of the maintenance and radwaste system flowdiagram of failed reactor components is developed and transfer vehicles and buildings are designed. Finally assuming JAERI Naka site as the reactor site layout of the whole FER plant system is developed. (author)

  15. Reactor structure and superconducting magnet system of ITER

    International Nuclear Information System (INIS)

    Tada, Eisuke; Yoshida, Kiyoshi; Shibanuma, Kiyoshi; Okuno, Kiyoshi; Tsuji, Hiroshi; Shimamoto, Susumu

    1993-01-01

    Fusion Experimental Reactors are one of the major steps toward realization of the fusion energy and the key objective are to demonstrate the scientific and technological feasibility prior to the Demo Fusion Reactor. ITER (International Thermonuclear Experimental Reactor) is one of experimental reactors and the conceptual design has been completed by the united efforts of USA, USSR, EC and Japan. In parallel with the conceptual design, key technology development in various areas has being conducted. This paper describes the overall design concepts and the latest technological achievements of the ITER reactor structure and superconducting magnet system. (author)

  16. Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems

    Energy Technology Data Exchange (ETDEWEB)

    D. E. Shropshire

    2009-01-01

    The Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems, prepared to support the U.S. Advanced Fuel Cycle Initiative (AFCI) systems analysis, provides a technology-oriented baseline system cost comparison between the open fuel cycle and closed fuel cycle systems. The intent is to understand their overall cost trends, cost sensitivities, and trade-offs. This analysis also improves the AFCI Program’s understanding of the cost drivers that will determine nuclear power’s cost competitiveness vis-a-vis other baseload generation systems. The common reactor-related costs consist of capital, operating, and decontamination and decommissioning costs. Fuel cycle costs include front-end (pre-irradiation) and back-end (post-iradiation) costs, as well as costs specifically associated with fuel recycling. This analysis reveals that there are large cost uncertainties associated with all the fuel cycle strategies, and that overall systems (reactor plus fuel cycle) using a closed fuel cycle are about 10% more expensive in terms of electricity generation cost than open cycle systems. The study concludes that further U.S. and joint international-based design studies are needed to reduce the cost uncertainties with respect to fast reactor, fuel separation and fabrication, and waste disposition. The results of this work can help provide insight to the cost-related factors and conditions needed to keep nuclear energy (including closed fuel cycles) economically competitive in the U.S. and worldwide. These results may be updated over time based on new cost information, revised assumptions, and feedback received from additional reviews.

  17. Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems

    International Nuclear Information System (INIS)

    Shropshire, D.E.

    2009-01-01

    The Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems, prepared to support the U.S. Advanced Fuel Cycle Initiative (AFCI) systems analysis, provides a technology-oriented baseline system cost comparison between the open fuel cycle and closed fuel cycle systems. The intent is to understand their overall cost trends, cost sensitivities, and trade-offs. This analysis also improves the AFCI Program's understanding of the cost drivers that will determine nuclear power's cost competitiveness vis-a-vis other baseload generation systems. The common reactor-related costs consist of capital, operating, and decontamination and decommissioning costs. Fuel cycle costs include front-end (pre-irradiation) and back-end (post-irradiation) costs, as well as costs specifically associated with fuel recycling. This analysis reveals that there are large cost uncertainties associated with all the fuel cycle strategies, and that overall systems (reactor plus fuel cycle) using a closed fuel cycle are about 10% more expensive in terms of electricity generation cost than open cycle systems. The study concludes that further U.S. and joint international-based design studies are needed to reduce the cost uncertainties with respect to fast reactor, fuel separation and fabrication, and waste disposition. The results of this work can help provide insight to the cost-related factors and conditions needed to keep nuclear energy (including closed fuel cycles) economically competitive in the U.S. and worldwide. These results may be updated over time based on new cost information, revised assumptions, and feedback received from additional reviews.

  18. Experiences with 'on-line' diagnostic instrumentation in nuclear power plants

    International Nuclear Information System (INIS)

    Gopal, R.; Ciaramitaro, W.; Smith, J.R.

    1981-01-01

    Over the past several years, Westinghouse has developed a coordinated system of on-line diagnostic instrumentation for the acquisition and analysis of data for diagnostics and incipient failure detection of critical plant equipment and systems. Primary motivation for this work is to improve Nuclear Steam Supply System (NSSS) availability and maintainability through the detection of malfunctions at their inception. These systems include: 1) Acoustic leak monitoring for detection and location of leaks in the primary system pressure boundary and other piping systems in PWR's; 2) Metal impact monitoring for detection of loose debris in the reactor vessel and steam generators; 3) Nuclear noise monitoring for monitoring core barrel vibration. Summarized in this paper are some of the features of the systems and inplant experience. (author)

  19. Experience with diagnostic instrumentation in nuclear power plants

    International Nuclear Information System (INIS)

    Gopal, R.; Ciaramitaro, W.

    1977-01-01

    Over the past several years, Westinghouse has developed a coordinated system of on-line diagnostic instrumentation for the acquisition and analysis of data for diagnostics and incipient failure detection of critical plant equipment and systems. Primary motivation for this work is to improve NSSS availability and Maintainability through the detection of malfunctions at their inception. These systems encompass the following areas: (1) Vibration Monitoring System for detection of changes in vibrational characteristics of the major components of Nuclear Steam Supply System (NSSS) and Balance of Plant (BOP); (2) Acoustic Monitoring System for detection and location of leaks in the primary system pressure boundary and other piping systems in PWRs; (3) Metal Impact Monitoring for detection of loose debris in the reactor vessel and steam generators; (4) Nuclear Noise Monitoring System for monitoring core barrel vibration; (5) Sensor Response Time Measurement System for detecting any degradation of process sensors; and (6) Transit Time Flow Meter for determining primary coolant flow rate. Summarized in this paper are some of the features of the systems and in-plant experience. These experiences demonstrate that diagnostic systems in combination with analytical and laboratory work for data interpretation do improve plant availability. (author)

  20. Modeling and performance of the MHTGR [Modular High-Temperature Gas-Cooled Reactor] reactor cavity cooling system

    International Nuclear Information System (INIS)

    Conklin, J.C.

    1990-04-01

    The Reactor Cavity Cooling System (RCCS) of the Modular High- Temperature Gas-Cooled Reactor (MHTGR) proposed by the U.S. Department of Energy is designed to remove the nuclear afterheat passively in the event that neither the heat transport system nor the shutdown cooling circulator subsystem is available. A computer dynamic simulation for the physical and mathematical modeling of and RCCS is described here. Two conclusions can be made form computations performed under the assumption of a uniform reactor vessel temperature. First, the heat transferred across the annulus from the reactor vessel and then to ambient conditions is very dependent on the surface emissivities of the reactor vessel and RCCS panels. These emissivities should be periodically checked to ensure the safety function of the RCCS. Second, the heat transfer from the reactor vessel is reduced by a maximum of 10% by the presence of steam at 1 atm in the reactor cavity annulus for an assumed constant in the transmission of radiant energy across the annulus can be expected to result in an increase in the reactor vessel temperature for the MHTGR. Further investigation of participating radiation media, including small particles, in the reactor cavity annulus is warranted. 26 refs., 7 figs., 1 tab

  1. New reactor technology: safety improvements in nuclear power systems.

    Science.gov (United States)

    Corradini, M L

    2007-11-01

    Almost 450 nuclear power plants are currently operating throughout the world and supplying about 17% of the world's electricity. These plants perform safely, reliably, and have no free-release of byproducts to the environment. Given the current rate of growth in electricity demand and the ever growing concerns for the environment, nuclear power can only satisfy the need for electricity and other energy-intensive products if it can demonstrate (1) enhanced safety and system reliability, (2) minimal environmental impact via sustainable system designs, and (3) competitive economics. The U.S. Department of Energy with the international community has begun research on the next generation of nuclear energy systems that can be made available to the market by 2030 or earlier, and that can offer significant advances toward these challenging goals; in particular, six candidate reactor system designs have been identified. These future nuclear power systems will require advances in materials, reactor physics, as well as thermal-hydraulics to realize their full potential. However, all of these designs must demonstrate enhanced safety above and beyond current light water reactor systems if the next generation of nuclear power plants is to grow in number far beyond the current population. This paper reviews the advanced Generation-IV reactor systems and the key safety phenomena that must be considered to guarantee that enhanced safety can be assured in future nuclear reactor systems.

  2. Recycling systems for BWR type reactors

    International Nuclear Information System (INIS)

    Takagi, Akio; Yamamoto, Fumiaki; Fukumoto, Ryuji.

    1986-01-01

    Purpose: To stabilize the coolant flowing characteristics and reactor core reactivity. Constitution: The recycling system in a BWR type reactor comprises a recycling pump disposed to the outside of a reactor pressure vessel, a ring header connected to the recycling pump through main pipe ways, and a plurality of pipes branched from and connected with the ring header and connected to a plurality of jet pumps within the pressure vessel. Then, by making the diameter for the pipeways of each of the branched pipes different from each other, the effective cross-sectional area is varied to thereby average the coolant flow rate supplied to each of the jet pumps. (Seki, T.)

  3. Development of a diagnostic system for identifying accident conditions in a reactor

    International Nuclear Information System (INIS)

    Santhosh; Gera, B.; Kumar, Mithilesh; Thangamani, I.; Prasad, Hari; Srivastava, A.; Dutta, Anu; Sharma, Pavan K.; Majumdar, P.; Verma, V.; Mukhopadhyay, D.; Ganju, Sunil; Chatterjee, B.; Sanyasi Rao, V.V.S.; Lele, H.G.; Ghosh, A.K.

    2009-07-01

    This report describes a methodology for identification of accident conditions in a nuclear reactor from the signals available to the operator. A large database of such signals is generated through analyses - for core, containment, environmental dispersion and radiological dose to train a computer code based on an Artificial Neural Networks (ANNs). At present, in the prediction mode, information on LOCA (location and size of break), status of availability of ECCS, and expected doses can be predicted well for a 220 MWe PHWR. (author)

  4. Power Trip Set-points of Reactor Protection System for New Research Reactor

    International Nuclear Information System (INIS)

    Lee, Byeonghee; Yang, Soohyung

    2013-01-01

    This paper deals with the trip set-point related to the reactor power considering the reactivity induced accident (RIA) of new research reactor. The possible scenarios of reactivity induced accidents were simulated and the effects of trip set-point on the critical heat flux ratio (CHFR) were calculated. The proper trip set-points which meet the acceptance criterion and guarantee sufficient margins from normal operation were then determined. The three different trip set-points related to the reactor power are determined based on the RIA of new research reactor during FP condition, over 0.1%FP and under 0.1%FP. Under various reactivity insertion rates, the CHFR are calculated and checked whether they meet the acceptance criterion. For RIA at FP condition, the acceptance criterion can be satisfied even if high power set-point is only used for reactor trip. Since the design of the reactor is still progressing and need a safety margin for possible design changes, 18 MW is recommended as a high power set-point. For RIA at 0.1%FP, high power setpoint of 18 MW and high log rate of 10%pp/s works well and acceptance criterion is satisfied. For under 0.1% FP operations, the application of high log rate is necessary for satisfying the acceptance criterion. Considering possible decrease of CHFR margin due to design changes, the high log rate is suggested to be 8%pp/s. Suggested trip set-points have been identified based on preliminary design data for new research reactor; therefore, these trip set-points will be re-established by considering design progress of the reactor. The reactor protection system (RPS) of new research reactor is designed for safe shutdown of the reactor and preventing the release of radioactive material to environment. The trip set point of RPS is essential for reactor safety, therefore should be determined to mitigate the consequences from accidents. At the same time, the trip set-point should secure margins from normal operational condition to avoid

  5. TFTR diagnostic control and data acquisition system

    International Nuclear Information System (INIS)

    Sauthoff, N.R.; Daniels, R.E.; PPL Computer Division

    1985-01-01

    General computerized control and data-handling support for TFTR diagnostics is presented within the context of the Central Instrumentation, Control and Data Acquisition (CICADA) System. Procedures, hardware, the interactive man--machine interface, event-driven task scheduling, system-wide arming and data acquisition, and a hierarchical data base of raw data and results are described. Similarities in data structures involved in control, monitoring, and data acquisition afford a simplification of the system functions, based on ''groups'' of devices. Emphases and optimizations appropriate for fusion diagnostic system designs are provided. An off-line data reduction computer system is under development

  6. A recommender system for medical imaging diagnostic.

    Science.gov (United States)

    Monteiro, Eriksson; Valente, Frederico; Costa, Carlos; Oliveira, José Luís

    2015-01-01

    The large volume of data captured daily in healthcare institutions is opening new and great perspectives about the best ways to use it towards improving clinical practice. In this paper we present a context-based recommender system to support medical imaging diagnostic. The system relies on data mining and context-based retrieval techniques to automatically lookup for relevant information that may help physicians in the diagnostic decision.

  7. Power reactor noise studies and applications

    Energy Technology Data Exchange (ETDEWEB)

    Arzhanov, V

    2002-03-01

    The present thesis deals with the neutron noise arising in power reactor systems. Generally, it can be divided into two major parts: first, neutron noise diagnostics, or more specifically, novel methods and algorithms to monitor nuclear industrial reactors; and second, contributions to neutron noise theory as applied to power reactor systems. Neutron noise diagnostics is presented by two topics. The first one is a theoretical study on the possibility to use a newly proposed current-flux (C/F) detector in Pressurised Water Reactors (PWR) for the localisation of anomalies. The second topic concerns various methods to detect guide tube impacting in Boiling Water Reactors (BWR). The significance of these problems comes from the operational experience. The thesis describes a novel method to localise vibrating control rods in a PWR by using only one C/F detector. Another novel method, based on wavelet analysis, is put forward to detect impacting guide tubes in a BWR. Neutron noise theory is developed for both Accelerator Driven Systems (ADS) and traditional reactors. By design the accelerator-driven systems would operate in a subcritical mode with a strong external source. This calls for a revision of many concepts and methods that have been developed for traditional reactors and also it poses a number of new problems. As for the latter, the thesis investigates the space-dependent neutron noise caused by a fluctuating source. It is shown that the frequency-dependent spatial behaviour exhibits some new properties that are different from those known in traditional critical systems. On the other hand, various reactor physics approximations (point kinetic, adiabatic etc.) have not been defined yet for the subcritical systems. In this respect the thesis presents a systematic formulation of the above mentioned approximations as well as investigations of their properties. Another important problem in neutron noise theory is the treatment of moving boundaries. In this case one

  8. Power reactor noise studies and applications

    International Nuclear Information System (INIS)

    Arzhanov, V.

    2002-03-01

    The present thesis deals with the neutron noise arising in power reactor systems. Generally, it can be divided into two major parts: first, neutron noise diagnostics, or more specifically, novel methods and algorithms to monitor nuclear industrial reactors; and second, contributions to neutron noise theory as applied to power reactor systems. Neutron noise diagnostics is presented by two topics. The first one is a theoretical study on the possibility to use a newly proposed current-flux (C/F) detector in Pressurised Water Reactors (PWR) for the localisation of anomalies. The second topic concerns various methods to detect guide tube impacting in Boiling Water Reactors (BWR). The significance of these problems comes from the operational experience. The thesis describes a novel method to localise vibrating control rods in a PWR by using only one C/F detector. Another novel method, based on wavelet analysis, is put forward to detect impacting guide tubes in a BWR. Neutron noise theory is developed for both Accelerator Driven Systems (ADS) and traditional reactors. By design the accelerator-driven systems would operate in a subcritical mode with a strong external source. This calls for a revision of many concepts and methods that have been developed for traditional reactors and also it poses a number of new problems. As for the latter, the thesis investigates the space-dependent neutron noise caused by a fluctuating source. It is shown that the frequency-dependent spatial behaviour exhibits some new properties that are different from those known in traditional critical systems. On the other hand, various reactor physics approximations (point kinetic, adiabatic etc.) have not been defined yet for the subcritical systems. In this respect the thesis presents a systematic formulation of the above mentioned approximations as well as investigations of their properties. Another important problem in neutron noise theory is the treatment of moving boundaries. In this case one

  9. 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.

  10. 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

  11. Diagnostic of the spatial and velocity distribution of alpha particles in tokamak fusion reactor using beat-wave generated lower hybrid wave. Progress report, 1994-1995

    International Nuclear Information System (INIS)

    Hwang, D.Q.; Horton, R.D.; Evans, R.

    1995-01-01

    The alpha particle population from fusion reactions in a DT tokamak reactor can have dramatic effects on the pressure profiles, energetic particle confinement, and the overall stability of the plasma; thus leading to important design consideration of a fusion reactor based on the tokamak concept. In order to fully understand the effects of the alpha population, a non-invasive diagnostic technique suitable for use in a reacting plasma environment needs to be developed to map out both the spatial and velocity distribution of the alphas. The proposed experimental goals for the eventual demonstration of LH wave interaction with a fast ion population is given in the reduced 3 year plan in table 1. At present time the authors are approaching the 8th month in their first year of this project. Up to now, their main effort has been concentrated in the operation of the two beat wave sources in burst mode. The second priority in the experimental project is the probe diagnostics and computer aided data acquisition system. The progress made so far is given, and they are ready to perform the beat-wave generated lower hybrid wave experiment. Some theoretical calculation had been reported at APS meetings. More refined theoretical models are being constructed in collaboration with Drs. J. Rogers and E. Valeo at PPPL

  12. Modeling and diagnostic techniques applicable to the analysis of pressure noise in pressurized water reactors and pressure-sensing systems

    International Nuclear Information System (INIS)

    Mullens, J.A.; Thie, J.A.

    1984-01-01

    Pressure noise data from a PWR are interpreted by means of a computer-implemented model. The model's parameters, namely hydraulic impedances and noise sources, are either calculated or deduced from fits to data. Its accuracy is encouraging and raises the possibility of diagnostic assistance for nuclear plant monitoring. A number of specific applications of pressure noise in the primary system of a PWR and in a pressure sensing system are suggested

  13. Target diagnostic system for the National Ignition Facility (NIF)

    International Nuclear Information System (INIS)

    Leeper, R.J.; Chandler, G.A.; Cooper, G.W.; Derzon, M.S.

    1996-01-01

    A review of recent progress on the design of a diagnostic system proposed for ignition target experiments on the National Ignition Facility (NIF) will be presented. This diagnostic package contains an extensive suite of optical, x-ray, gamma-ray, and neutron diagnostics that enable measurements of the performance of both direct and indirect driven NIF targets. The philosophy used in designing all of the diagnostics in the set has emphasized redundant and independent measurement of fundamental physical quantities relevant to the operation of the NIF target. A unique feature of these diagnostics is that they are being designed to be capable of operating, in the high radiation, EMP, and debris backgrounds expected on the NIF facility. The diagnostic system proposed can be categorized into three broad areas: laser characterization, hohlraum characterization, and capsule performance diagnostics. The operating principles of a representative instrument from each class of diagnostic employed in this package will be summarized and illustrated with data obtained in recent prototype diagnostic tests

  14. System and prospect assesment of the small innovative reactor IRIS-50

    International Nuclear Information System (INIS)

    Lumbanraja, Sahala M.; Wibowo

    2002-01-01

    System and prospect of the small innovative reactor IRIS-50 in Indonesia have been studied. IRIS-50 (International Reactor Innovative and Secure) is an advanced light water cooled modular reactor being developed by an international consortium led by Westinghouse. This reactor is specifically developed to match market demands, or to developing country. This reactor is based on simplified operation and maintenance, enhanced and safety, easy to inspect, short construction time, small investment cost, competitive generating cost, and easily suited to the infrastructures. IRIS main characteristic is integral reactor concept, being all the major reactor coolant system components located inside the pressure vessel

  15. 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)

  16. Reliability analysis of reactor protection systems

    International Nuclear Information System (INIS)

    Alsan, S.

    1976-07-01

    A theoretical mathematical study of reliability is presented and the concepts subsequently defined applied to the study of nuclear reactor safety systems. The theory is applied to investigations of the operational reliability of the Siloe reactor from the point of view of rod drop. A statistical study conducted between 1964 and 1971 demonstrated that most rod drop incidents arose from circumstances associated with experimental equipment (new set-ups). The reliability of the most suitable safety system for some recently developed experimental equipment is discussed. Calculations indicate that if all experimental equipment were equipped with these new systems, only 1.75 rod drop accidents would be expected to occur per year on average. It is suggested that all experimental equipment should be equipped with these new safety systems and tested every 21 days. The reliability of the new safety system currently being studied for the Siloe reactor was also investigated. The following results were obtained: definite failures must be detected immediately as a result of the disturbances produced; the repair time must not exceed a few hours; the equipment must be tested every week. Under such conditions, the rate of accidental rod drops is about 0.013 on average per year. The level of nondefinite failures is less than 10 -6 per hour and the level of nonprotection 1 hour per year. (author)

  17. Thioaptamer Diagnostic System, Phase II

    Data.gov (United States)

    National Aeronautics and Space Administration — AM Biotechnologies (AM) in partnership with Sandia National Laboratories will develop a Thioaptamer Diagnostic System (TDS) in response to Topic X10.01 Reusable...

  18. A Design of Alarm System in a Research Reactor Facility

    International Nuclear Information System (INIS)

    Park, Jaekwan; Jang, Gwisook; Seo, Sangmun; Suh, Yongsuk

    2013-01-01

    The digital alarm system has become an indispensable design to process a large amount of alarms of power plants. Korean research reactor operated for decades maintains a hybrid alarm system with both an analog annunciator and a digital alarm display. In this design, several alarms are indicated on an analog panel and digital display, respectively, and it requires more attention and effort of the operators. As proven in power plants, a centralized alarm system design is necessary for a new research reactor. However, the number of alarms and operators in a research reactor is significantly lesser than power plants. Thus, simplification should be considered as an important factor for the operation efficiency. This paper introduces a simplified alarm system. As advances in information technology, fully digitalized alarm systems have been applied to power plants. In a new research reactor, it will be more useful than an analog or hybrid configuration installed in research reactors decades ago. However, the simplification feature should be considered as an important factor because the number of alarms and number of operators in a research reactor is significantly lesser than in power plants

  19. 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.)

  20. Design Requirements of an Advanced HANARO Reactor Core Cooling System

    International Nuclear Information System (INIS)

    Park, Yong Chul; Ryu, Jeong Soo

    2007-12-01

    An advanced HANARO Reactor (AHR) is an open-tank-type and generates thermal power of 20 MW and is under conceptual design phase for developing it. The thermal power is including a core fission heat, a temporary stored fuel heat in the pool, a pump heat and a neutron reflecting heat in the reflector vessel of the reactor. In order to remove the heat load, the reactor core cooling system is composed of a primary cooling system, a primary cooling water purification system and a reflector cooling system. The primary cooling system must remove the heat load including the core fission heat, the temporary stored fuel heat in the pool and the pump heat. The purification system must maintain the quality of the primary cooling water. And the reflector cooling system must remove the neutron reflecting heat in the reflector vessel of the reactor and maintain the quality of the reflector. In this study, the design requirement of each system has been carried out using a design methodology of the HANARO within a permissible range of safety. And those requirements are written by english intend to use design data for exporting the research reactor

  1. Plasma nitriding monitoring reactor: A model reactor for studying plasma nitriding processes using an active screen

    Science.gov (United States)

    Hamann, S.; Börner, K.; Burlacov, I.; Spies, H.-J.; Strämke, M.; Strämke, S.; Röpcke, J.

    2015-12-01

    A laboratory scale plasma nitriding monitoring reactor (PLANIMOR) has been designed to study the basics of active screen plasma nitriding (ASPN) processes. PLANIMOR consists of a tube reactor vessel, made of borosilicate glass, enabling optical emission spectroscopy (OES) and infrared absorption spectroscopy. The linear setup of the electrode system of the reactor has the advantages to apply the diagnostic approaches on each part of the plasma process, separately. Furthermore, possible changes of the electrical field and of the heat generation, as they could appear in down-scaled cylindrical ASPN reactors, are avoided. PLANIMOR has been used for the nitriding of steel samples, achieving similar results as in an industrial scale ASPN reactor. A compact spectrometer using an external cavity quantum cascade laser combined with an optical multi-pass cell has been applied for the detection of molecular reaction products. This allowed the determination of the concentrations of four stable molecular species (CH4, C2H2, HCN, and NH3). With the help of OES, the rotational temperature of the screen plasma could be determined.

  2. Plasma nitriding monitoring reactor: A model reactor for studying plasma nitriding processes using an active screen

    International Nuclear Information System (INIS)

    Hamann, S.; Röpcke, J.; Börner, K.; Burlacov, I.; Spies, H.-J.; Strämke, M.; Strämke, S.

    2015-01-01

    A laboratory scale plasma nitriding monitoring reactor (PLANIMOR) has been designed to study the basics of active screen plasma nitriding (ASPN) processes. PLANIMOR consists of a tube reactor vessel, made of borosilicate glass, enabling optical emission spectroscopy (OES) and infrared absorption spectroscopy. The linear setup of the electrode system of the reactor has the advantages to apply the diagnostic approaches on each part of the plasma process, separately. Furthermore, possible changes of the electrical field and of the heat generation, as they could appear in down-scaled cylindrical ASPN reactors, are avoided. PLANIMOR has been used for the nitriding of steel samples, achieving similar results as in an industrial scale ASPN reactor. A compact spectrometer using an external cavity quantum cascade laser combined with an optical multi-pass cell has been applied for the detection of molecular reaction products. This allowed the determination of the concentrations of four stable molecular species (CH 4 , C 2 H 2 , HCN, and NH 3 ). With the help of OES, the rotational temperature of the screen plasma could be determined

  3. Plasma nitriding monitoring reactor: A model reactor for studying plasma nitriding processes using an active screen

    Energy Technology Data Exchange (ETDEWEB)

    Hamann, S., E-mail: hamann@inp-greifswald.de; Röpcke, J. [INP-Greifswald, Felix-Hausdorff-Str. 2, 17489 Greifswald (Germany); Börner, K.; Burlacov, I.; Spies, H.-J. [TU Bergakademie Freiberg, Institute of Materials Engineering, Gustav-Zeuner-Str. 5, 09599 Freiberg (Germany); Strämke, M.; Strämke, S. [ELTRO GmbH, Arnold-Sommerfeld-Ring 3, 52499 Baesweiler (Germany)

    2015-12-15

    A laboratory scale plasma nitriding monitoring reactor (PLANIMOR) has been designed to study the basics of active screen plasma nitriding (ASPN) processes. PLANIMOR consists of a tube reactor vessel, made of borosilicate glass, enabling optical emission spectroscopy (OES) and infrared absorption spectroscopy. The linear setup of the electrode system of the reactor has the advantages to apply the diagnostic approaches on each part of the plasma process, separately. Furthermore, possible changes of the electrical field and of the heat generation, as they could appear in down-scaled cylindrical ASPN reactors, are avoided. PLANIMOR has been used for the nitriding of steel samples, achieving similar results as in an industrial scale ASPN reactor. A compact spectrometer using an external cavity quantum cascade laser combined with an optical multi-pass cell has been applied for the detection of molecular reaction products. This allowed the determination of the concentrations of four stable molecular species (CH{sub 4}, C{sub 2}H{sub 2}, HCN, and NH{sub 3}). With the help of OES, the rotational temperature of the screen plasma could be determined.

  4. Advanced High-Temperature Reactor Dynamic System Model Development: April 2012 Status

    Energy Technology Data Exchange (ETDEWEB)

    Qualls, A L; Cetiner, M S; Wilson, Jr, T L

    2012-04-30

    The Advanced High-Temperature Reactor (AHTR) is a large-output fluoride-salt-cooled high-temperature reactor (FHR). An early-phase preconceptual design of a 1500 MW(e) power plant was developed in 2011 [Refs. 1 and 2]. An updated version of this plant is shown as Fig. 1. FHRs feature low-pressure liquid fluoride salt cooling, coated-particle fuel, a high-temperature power cycle, and fully passive decay heat rejection. The AHTR is designed to be a “walk away” reactor that requires no action to prevent large off-site releases following even severe reactor accidents. This report describes the development of dynamic system models used to further the AHTR design toward that goal. These models predict system response during warmup, startup, normal operation, and limited off-normal operating conditions. Severe accidents that include a loss-of-fluid inventory are not currently modeled. The scope of the models is limited to the plant power system, including the reactor, the primary and intermediate heat transport systems, the power conversion system, and safety-related or auxiliary heat removal systems. The primary coolant system, the intermediate heat transport system and the reactor building structure surrounding them are shown in Fig. 2. These systems are modeled in the most detail because the passive interaction of the primary system with the surrounding structure and heat removal systems, and ultimately the environment, protects the reactor fuel and the vessel from damage during severe reactor transients. The reactor silo also plays an important role during system warmup. The dynamic system modeling tools predict system performance and response. The goal is to accurately predict temperatures and pressures within the primary, intermediate, and power conversion systems and to study the impacts of design changes on those responses. The models are design tools and are not intended to be used in reactor qualification. The important details to capture in the primary

  5. Light water reactor safeguards system evaluation

    International Nuclear Information System (INIS)

    Varnado, G.B.; Ericson, D.M. Jr.; Bennett, H.A.; Hulme, B.L.; Daniel, S.L.

    1978-01-01

    A methodology for assessing the effectiveness of safeguards systems was developed in this study and was applied to a typical light water reactor plant. The relative importance of detection systems, barriers, response forces and other safeguards system components was examined in extensive parameter variation studies. (author)

  6. Diapo, applying advanced AI methods to diagnosis of PWR reactor coolant pump

    International Nuclear Information System (INIS)

    Porcheron, M.; Ricard, B.

    1993-01-01

    Electricite de France has decided to increase the capabilities of its monitoring and diagnostic architecture with the development of an AI system for reactor coolant pump diagnostic support. This development is carried out with the cooperation of the equipment constructor Jeumont Schneider Industries. This diagnostic system will eventually be included in an integrated surveillance architecture. We present the architecture of the system and the basics of the knowledge model used. Main data for diagnosis are provided by sensor data issued by the pump monitoring system. Diagnostic reasoning is based on the cooperation of two main activities : a heuristic search among typical symptomatic situations that leads to the formulation of hypotheses and a ''deep'' causal analysis that consists in backtracking from identified situations up to initial faults or causes. This approach is well fitted to field expert reasoning, and provides powerful diagnostic capabilities that help to overcome conventional limitations of expert systems entirely based on heuristic knowledge. (authors). 9 figs., 11 refs

  7. Ventilation system in the RA reactor building - design specifications

    International Nuclear Information System (INIS)

    Badrljica, R.

    1984-09-01

    Protective role of the ventilation system of nuclear facilities involve construction of ventilation barriers which prevent release of radioactive particulates or gases, elimination od radioactive particulates and gases from the air which is released from contaminated zones into the reactor environment. Ventilation barriers are created by dividing the building into a number of ventilation zones with different sub pressure compared to the atmospheric pressure. The RA reactor building is divided into four ventilation zones. First zone is the zone of highest risk. It includes reactor core with horizontal experimental channels, underground rooms of the primary coolant system (D 2 O), helium system, hot cells and the space above the the reactor core. Second zone is the reactor hall and the room for irradiated fuel storage. The third zone includes corridors in the basement, ground floor and first floor where the probability of contamination is small. The fourth zone includes the annex where the contamination risk is low. There is no have natural air circulation in the reactor building. Ventilators for air input and outlet maintain the sub pressure in the building (pressure lower than the atmospheric pressure). This prevents release of radioactivity into the atmosphere [sr

  8. Self-adapted thermocouple-diagnostic complex

    International Nuclear Information System (INIS)

    Alekseev, S.V.; Grankovskij, K.Eh.; Olejnikov, P.P.; Prijmak, S.V.; Shikalov, V.F.

    2003-01-01

    A self-adapted thermocouple-diagnostic complex (STDC) for obtaining the reliable data on the coolant temperature in the reactors of NPP is described. The STDC in based on the thermal pulse monitoring of a thermocouple in the measuring channel of a reactor. Measurement method and STDC composition are substantiated. It is shown that introduction of the developed STDC ensures realization of precise and reliable temperature monitoring in the reactors of all types [ru

  9. Fast breeder reactor reference system classification for the ENEA data bank

    International Nuclear Information System (INIS)

    Righini, R.

    1988-01-01

    This report contains the Reference System Classification (RSC) of fast breeder reactors: it provides a functional system breakdown of the reactor. For each system the following important characteristics are reported: the main function, the mode of operation, its location in the reactor, the main interface system, its main components and the component working environment (fluid and/or atmosphere type). The RSC represent a basic step in organizing the ENEA data bank for the registration and processing of reliability data on typical fast reactor components; it provides a functional component breakdown and represent a plant-unique identification in the process of omogenization of event-data coming from different reactors. In this report it was tried to take into account different generations of nuclear power plants, different plant layouts and solutions: in particular loop and pool reactors are separately treated

  10. Performance of the prism reactor's passive decay heat removal system

    International Nuclear Information System (INIS)

    Magee, P.M.; Hunsbedt, A.

    1989-01-01

    The PRISM modular reactor concept has a totally passive safety-grade decay heat removal system referred to as the Reactor Vessel Auxiliary Cooling System (RVACS) that rejects heat from the reactor by radiation and natural convection of air. The system is inherently reliable and is not subject to the failure modes commonly associated with active cooling systems. The thermal performance of RVACS exceeds requirements and significant thermal margins exist. RVACS has been shown to perform its function under many postulated accident conditions. The PRISM power plant is equipped with three methods for shutdown: condenser cooling in conjunction with intermediate sodium and steam generator systems, and auxiliary cooling system (ACS) which removes heat from the steam generator by natural convection of air and transport of heat from the core by natural convection in the primary and intermediate systems, and a safety- grade reactor vessel auxiliary cooling system (RVACS) which removes heat passively from the reactor containment vessel by natural convection of air. The combination of one active and two passive systems provides a highly reliable and economical shutdown heat removal system. This paper provides a summary of the RVACS thermal performance for expected operating conditions and postulated accident events. The supporting experimental work, which substantiates the performance predictions, is also summarized

  11. Beam profile diagnostics system for SDUV-FEL

    International Nuclear Information System (INIS)

    Xu Yichao; Han Lifeng; Chen Yongzhong

    2010-01-01

    A new beam profile diagnostics system for Shanghai Deep Ultraviolet Free Electron Laser (SDUV-FEL) has been developed based on industrial Ethernet, with good versatility and scalability. The system includes three major subsystems for image acquisition,pneumatic control and stepper motor control, respectively. Virtual instrument technology is adopted to drive the devices, and to develop the measurement software. In this paper,we describe the system structure, and its hardware and software design. The results of system commissioning are given as well. As an important diagnostic tool and data acquisition method, the system has been successfully applied to the measurement and control of the SDUV-FEL.(authors)

  12. SP-100 Program: space reactor system and subsystem investigations

    International Nuclear Information System (INIS)

    Harty, R.B.

    1983-01-01

    For a space reactor power system, a comprehensive safety program will be required to assure that no undue risk is present. This report summarizes the nuclear safety review/approval process that will be required for a space reactor system. The documentation requirements are presented along with a summary of the required contents of key documents. Finally, the aerospace safety program conducted for the SNAP-10A reactor system is summarized. The results of this program are presented to show the type of program that can be expected and to provide information that could be usable in future programs

  13. Development trends for diagnostic systems in nuclear power plants

    International Nuclear Information System (INIS)

    Kunze, U.; Pohl, U.

    1998-01-01

    Monitoring systems used in nuclear power plants have made remarkable progress over the past four or five years. Development has followed the trends and changes in philosophy for the purpose of monitoring systems in nuclear power plants: They are no longer expected to fulfill only safety tasks, the plant personnel require information on which to base condition-oriented maintenance. A new generation of monitoring and diagnostic systems has been developed by Siemens recently. This new generation, called Series '95, is PC-based. An overview is given for the KUeS '95 loose parts diagnostic system, the SUeS '95 vibration monitoring system, the FLUeS leak detection system and the SIPLUG valve diagnostics system. The objectives behind the development of these new systems are both safety-related and economic. The new systems improve the reliability and quality of monitoring techniques and incorporate better detection and diagnostic capabilities. Progress has also been made in automation of the systems so as to reduce routine work, give higher sensitivity for the monitoring task and reduce the scope of maintenance. (author)

  14. Hybrid Plasma Reactor/Filter for Transportable Collective Protection Systems

    Energy Technology Data Exchange (ETDEWEB)

    Josephson, Gary B.; Tonkyn, Russell G.; Frye, J. G.; Riley, Brian J.; Rappe, Kenneth G.

    2011-04-06

    Pacific Northwest National Laboratory (PNNL) has performed an assessment of a Hybrid Plasma/Filter system as an alternative to conventional methods for collective protection. The key premise of the hybrid system is to couple a nonthermal plasma (NTP) reactor with reactive adsorption to provide a broader envelope of protection than can be provided through a single-solution approach. The first step uses highly reactive species (e.g. oxygen radicals, hydroxyl radicals, etc.) created in a nonthermal plasma (NTP) reactor to destroy the majority (~75% - 90%) of an incoming threat. Following the NTP reactor an O3 reactor/filter uses the O3 created in the NTP reactor to further destroy the remaining organic materials. This report summarizes the laboratory development of the Hybrid Plasma Reactor/Filter to protect against a ‘worst-case’ simulant, methyl bromide (CH3Br), and presents a preliminary engineering assessment of the technology to Joint Expeditionary Collective Protection performance specifications for chemical vapor air purification technologies.

  15. EBR-2 [Experimental Breeder Reactor-2] test programs

    International Nuclear Information System (INIS)

    Sackett, J.I.; Lehto, W.K.; Lindsay, R.W.; Planchon, H.P.; Lambert, J.D.B.; Hill, D.J.

    1990-01-01

    The Experimental Breeder Reactor-2 (EBR-2) is a sodium cooled power reactor supplying about 20 MWe to the Idaho National Engineering Laboratory (INEL) grid and, in addition, is the key component in the development of the Integral Fast Reactor (IFR). EBR-2's testing capability is extensive and has seen four major phases: (1) demonstration of LMFBR power plant feasibility, (2) irradiation testing for fuel and material development, (3) testing the off-normal performance of fuel and plant systems and (4) operation as the IFR prototype, developing and demonstrating the IFR technology associated with fuel and plant design. Specific programs being carried out in support of the IFR include advanced fuels and materials development, advanced control system development, plant diagnostics development and component testing. This paper discusses EBR-2 as the IFR prototype and the associated testing programs. 29 refs

  16. 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)

  17. Small reactor power systems for manned planetary surface bases

    Energy Technology Data Exchange (ETDEWEB)

    Bloomfield, H.S.

    1987-12-01

    A preliminary feasibility study of the potential application of small nuclear reactor space power systems to manned planetary surface base missions was conducted. The purpose of the study was to identify and assess the technology, performance, and safety issues associated with integration of reactor power systems with an evolutionary manned planetary surface exploration scenario. The requirements and characteristics of a variety of human-rated modular reactor power system configurations selected for a range of power levels from 25 kWe to hundreds of kilowatts is described. Trade-off analyses for reactor power systems utilizing both man-made and indigenous shielding materials are provided to examine performance, installation and operational safety feasibility issues. The results of this study have confirmed the preliminary feasibility of a wide variety of small reactor power plant configurations for growth oriented manned planetary surface exploration missions. The capability for power level growth with increasing manned presence, while maintaining safe radiation levels, was favorably assessed for nominal 25 to 100 kWe modular configurations. No feasibility limitations or technical barriers were identified and the use of both distance and indigenous planetary soil material for human rated radiation shielding were shown to be viable and attractive options.

  18. Small reactor power systems for manned planetary surface bases

    International Nuclear Information System (INIS)

    Bloomfield, H.S.

    1987-12-01

    A preliminary feasibility study of the potential application of small nuclear reactor space power systems to manned planetary surface base missions was conducted. The purpose of the study was to identify and assess the technology, performance, and safety issues associated with integration of reactor power systems with an evolutionary manned planetary surface exploration scenario. The requirements and characteristics of a variety of human-rated modular reactor power system configurations selected for a range of power levels from 25 kWe to hundreds of kilowatts is described. Trade-off analyses for reactor power systems utilizing both man-made and indigenous shielding materials are provided to examine performance, installation and operational safety feasibility issues. The results of this study have confirmed the preliminary feasibility of a wide variety of small reactor power plant configurations for growth oriented manned planetary surface exploration missions. The capability for power level growth with increasing manned presence, while maintaining safe radiation levels, was favorably assessed for nominal 25 to 100 kWe modular configurations. No feasibility limitations or technical barriers were identified and the use of both distance and indigenous planetary soil material for human rated radiation shielding were shown to be viable and attractive options

  19. Proton storage ring (PSR) diagnostics and control system

    International Nuclear Information System (INIS)

    Clout, P.

    1983-01-01

    When any new accelerator or storage ring is built that advances the state of the art, the diagnostic system becomes extremely important in tuning the facility to full specification. This paper will discuss the various diagnostic devices planned or under construction for the PSR and their connection into the control system

  20. 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.)

  1. Benchmarking Diagnostic Algorithms on an Electrical Power System Testbed

    Science.gov (United States)

    Kurtoglu, Tolga; Narasimhan, Sriram; Poll, Scott; Garcia, David; Wright, Stephanie

    2009-01-01

    Diagnostic algorithms (DAs) are key to enabling automated health management. These algorithms are designed to detect and isolate anomalies of either a component or the whole system based on observations received from sensors. In recent years a wide range of algorithms, both model-based and data-driven, have been developed to increase autonomy and improve system reliability and affordability. However, the lack of support to perform systematic benchmarking of these algorithms continues to create barriers for effective development and deployment of diagnostic technologies. In this paper, we present our efforts to benchmark a set of DAs on a common platform using a framework that was developed to evaluate and compare various performance metrics for diagnostic technologies. The diagnosed system is an electrical power system, namely the Advanced Diagnostics and Prognostics Testbed (ADAPT) developed and located at the NASA Ames Research Center. The paper presents the fundamentals of the benchmarking framework, the ADAPT system, description of faults and data sets, the metrics used for evaluation, and an in-depth analysis of benchmarking results obtained from testing ten diagnostic algorithms on the ADAPT electrical power system testbed.

  2. Systematic Benchmarking of Diagnostic Technologies for an Electrical Power System

    Science.gov (United States)

    Kurtoglu, Tolga; Jensen, David; Poll, Scott

    2009-01-01

    Automated health management is a critical functionality for complex aerospace systems. A wide variety of diagnostic algorithms have been developed to address this technical challenge. Unfortunately, the lack of support to perform large-scale V&V (verification and validation) of diagnostic technologies continues to create barriers to effective development and deployment of such algorithms for aerospace vehicles. In this paper, we describe a formal framework developed for benchmarking of diagnostic technologies. The diagnosed system is the Advanced Diagnostics and Prognostics Testbed (ADAPT), a real-world electrical power system (EPS), developed and maintained at the NASA Ames Research Center. The benchmarking approach provides a systematic, empirical basis to the testing of diagnostic software and is used to provide performance assessment for different diagnostic algorithms.

  3. An intelligent safety system concept for future CANDU reactors

    International Nuclear Information System (INIS)

    Hinds, H.W.

    1980-01-01

    A review of the current Regional Over-power Trip (ROPT) system employed on the Bruce NGS-A reactors confirmed the belief that future reactors should have an improved ROPT system. We are developing such an 'intelligent' safety system. It uses more of the available information on reactor status and employs modern computer technology. Fast triplicated safety computers compute maps of fuel channel power, based on readings from prompt-responding flux detectors. The coefficients for this calculation are downloaded periodically from a fourth supervisor computer. These coefficients are based on a detailed 3-D flux shape derived from physics data and other plant information. A demonstration of one of three safety channels of such a system is planned. (auth)

  4. 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.)

  5. 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.)

  6. Diagnostics and camera strobe timers for hydrogen pellet injectors

    International Nuclear Information System (INIS)

    Bauer, M.L.; Fisher, P.W.; Qualls, A.L.

    1993-01-01

    Hydrogen pellet injectors have been used to fuel fusion experimental devices for the last decade. As part of developments to improve pellet production and velocity, various diagnostic devices were implemented, ranging from witness plates to microwave mass meters to high speed photography. This paper will discuss details of the various implementations of light sources, cameras, synchronizing electronics and other diagnostic systems developed at Oak Ridge for the Tritium Proof-of-Principle (TPOP) experiment at the Los Alamos National Laboratory's Tritium System Test Assembly (TSTA), a system built for the Oak Ridge Advanced Toroidal Facility (ATF), and the Tritium Pellet Injector (TPI) built for the Princeton Tokamak Fusion Test Reactor (TFTR). Although a number of diagnostic systems were implemented on each pellet injector, the emphasis here will be on the development of a synchronization system for high-speed photography using pulsed light sources, standard video cameras, and video recorders. This system enabled near real-time visualization of the pellet shape, size and flight trajectory over a wide range of pellet speeds and at one or two positions along the flight path. Additionally, the system provides synchronization pulses to the data system for pseudo points along the flight path, such as the estimated plasma edge. This was accomplished using an electronic system that took the time measured between sets of light gates, and generated proportionally delayed triggers for light source strobes and pseudo points. Systems were built with two camera stations, one located after the end of the barrel, and a second camera located closer to the main reactor vessel wall. Two or three light gates were used to sense pellet velocity and various spacings were implemented on the three experiments. Both analog and digital schemes were examined for implementing the delay system. A digital technique was chosen

  7. Beam diagnostic system for SSC on HIRFL central console

    International Nuclear Information System (INIS)

    Zhang Guixu; Wang Zhen; Huang Tuanhua

    1998-01-01

    The SSC ion beam diagnostic system on the console of HIRFL in institute of modern physics is presented. The information between console and diagnostic system can be transferred via DECnet communication. The central computer for HIRFL console is VAX-8350, the working computer of diagnostic system is changed from IBM PC/XT to COMPAQ 486, and the operating program is rewritten from FORTRAN to C. In order to communicate information, DECnet TTT function is put into both programs on the VAX and PC

  8. The flow measurement methods for the primary system of integral reactors

    International Nuclear Information System (INIS)

    Lee, J.; Seo, J. K.; Lee, D. J.

    2001-01-01

    It is the common features of the integral reactors that the main components of the primary system are installed within the reactor vessel, and so there are no any flow pipes connecting the reactor coolant pumps or steam generators. Due to no any flow pipes, it is impossible to measure the differential pressure at the primary system of the integral reactors, and it also makes impossible measure the primary coolant flow rate. The objective of the study is to draw up the flow measurement methods for the primary system of integral reactors. As a result of the review, we have made a selection of the flow measurement method by pump speed, bt HBM, and by pump motor power as the flow measurement methods for the primary system of integral reactors. Peculiarly, we did not found out a precedent which the direct pump motor power-flow rate curve is used as the flow measurement method in the existing commercial nuclear power reactors. Therefore, to use this method for integral reactors, it is needed to bear the follow-up measures in mind. The follow-up measures is included in this report

  9. The roles of EBR-II and TREAT [Transient Reactor Test] in establishing liquid metal reactor safety

    International Nuclear Information System (INIS)

    Sackett, J.I.; Lehto, W.K.; Solbrig, C.W.

    1990-01-01

    This paper examines the role of the Experimental Breeder Reactor II (EBR-II) and Transient Reactor Test (TREAT) facilities in contributing to the understanding and resolution of key safety issues in liquid metal reactor safety during the decade of the 80's. Fuels and materials testing has been carried out to address questions on fuels behavior during steady-state and upset conditions. In addition, EBR-II has conducted plant tests to demonstrate passive response to ATWS events and to develop control and diagnostic strategies for safe operation of advanced LMRs. TREAT and EBR-II complement each other and between them provide a transient testing capability that covers the whole range of concerns during overpower conditions. EBR-II, with use of the special Automatic Control Rod Drive System, can generate power change rates that overlap the lower end of the TREAT capability. 21 refs

  10. Automated ultrasonic examination of light water reactor systems

    International Nuclear Information System (INIS)

    Walter, J.H.

    1975-01-01

    An automated ultrasonic examination system has been developed to meet the pre- and inservice inspection requirements of light water reactors. This system features remotely-controlled travelling instrument carriers, computerized collection and storage or inspection data in a manner providing real time comparison against code standards, and computer control over the positioning of the instrument carriers to provide precise location data. The system is currently being utilized in the field for a variety of reactor inspections. The principal features of the system and the recent inspection experience are discussed. (author)

  11. In-service inspections of the reactor cooling system of pressurized water reactors

    International Nuclear Information System (INIS)

    Fuerste, W.; Hohnerlein, G.; Werden, B.

    1982-01-01

    In order to guarantee constant safety of the components of the reactor cooling system, regular in-service inspections are carried out after commissioning of the nuclear power plant. This contribution is concerned with the components of the reactor cooling system, referring to the legal requirements, safety-related purposes and scope of the in-service inspections during the entire period of operation of a nuclear power plant. Reports are made with respect to type, examination intervals, examination technique, results and future development. The functional tests which are carried out within the scope of the in-service inspections are not part of this contribution. (orig.) [de

  12. Scanning tunneling microscope assembly, reactor, and system

    Science.gov (United States)

    Tao, Feng; Salmeron, Miquel; Somorjai, Gabor A

    2014-11-18

    An embodiment of a scanning tunneling microscope (STM) reactor includes a pressure vessel, an STM assembly, and three spring coupling objects. The pressure vessel includes a sealable port, an interior, and an exterior. An embodiment of an STM system includes a vacuum chamber, an STM reactor, and three springs. The three springs couple the STM reactor to the vacuum chamber and are operable to suspend the scanning tunneling microscope reactor within the interior of the vacuum chamber during operation of the STM reactor. An embodiment of an STM assembly includes a coarse displacement arrangement, a piezoelectric fine displacement scanning tube coupled to the coarse displacement arrangement, and a receiver. The piezoelectric fine displacement scanning tube is coupled to the coarse displacement arrangement. The receiver is coupled to the piezoelectric scanning tube and is operable to receive a tip holder, and the tip holder is operable to receive a tip.

  13. Breeder reactor fuel fabrication system development

    International Nuclear Information System (INIS)

    Bennett, D.W.; Fritz, R.L.; McLemore, D.R.; Yatabe, J.M.

    1981-01-01

    Significant progress has been made in the design and development of remotely operated breeder reactor fuel fabrication and support systems (e.g., analytical chemistry). These activities are focused by the Secure Automated Fabrication (SAF) Program sponsored by the Department of Energy to provide: a reliable supply of fuel pins to support US liquid metal cooled breeder reactors and at the same time demonstrate the fabrication of mixed uranium/plutonium fuel by remotely operated and automated methods

  14. Evaluation of tritium production rate in a gas-cooled reactor with continuous tritium recovery system for fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Matsuura, Hideaki, E-mail: mat@nucl.kyushu-u.ac.jp [Department of Applied Quantum Physics and Nuclear Engineering, Kyushu University, 744 Motooka, Fukuoka 819-0395 (Japan); Nakaya, Hiroyuki; Nakao, Yasuyuki [Department of Applied Quantum Physics and Nuclear Engineering, Kyushu University, 744 Motooka, Fukuoka 819-0395 (Japan); Shimakawa, Satoshi; Goto, Minoru; Nakagawa, Shigeaki [Japan Atomic Energy Agency, 4002 Oarai, Ibaraki 311-1393 (Japan); Nishikawa, Masabumi [Graduate School of Engineering Science, Kyushu University, 6-10-1 Hakozaki, Fukuoka 812-8581 (Japan)

    2013-10-15

    Highlights: • The performance of a gas-cooled reactor as a tritium production system was studied. • A continuous tritium recovery using helium gas was considered. • Gas-cooled reactors with 3 GW output in all can produce ∼6 kg of tritium in a year • Performance of the system was examined for Li{sub 4}SiO{sub 4}, Li{sub 2}TiO{sub 3} and LiAlO{sub 2} compounds. -- Abstract: The performance of a high-temperature gas-cooled reactor as a tritium production with continuous tritium recovery system is examined. A gas turbine high-temperature reactor of 300-MWe (600 MW) nominal capacity (GTHTR300) is assumed as the calculation target, and using the continuous-energy Monte Carlo transport code MVP-BURN, burn-up simulations for the three-dimensional entire-core region of the GTHTR300 were performed. A Li loading pattern for the continuous tritium recovery system in the gas-cooled reactor is presented. It is shown that module gas-cooled reactors with a total thermal output power of 3 GW in all can produce ∼6 kg of tritium maximum in a year.

  15. 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

  16. BEACON TSM application system to the operation of PWR reactors

    International Nuclear Information System (INIS)

    Lozano, J. A.; Mildrum, C.; Serrano, J. F.

    2012-01-01

    BEACON-TSM is an advanced core monitoring system for PWR reactor cores, and also offers the possibility to perform a wide range of predictive calculation in support of reactor operation. BEACON-TSM is presently installed and licensed in the 5 Spanish PWR reactors of standard Westinghouse design. the purpose of this paper is to describe the features of this software system and to show the advantages obtainable by a nuclear power plant from its use. To illustrate the capabilities and benefits of BEACON-TSM two real case reactor operating situations are presented. (Author)

  17. PRISM reactor system design and analysis of postulated unscrammed events

    International Nuclear Information System (INIS)

    Van Tuyle, G.J.; Slovik, G.C.

    1991-01-01

    Key safety characteristics of the PRISM reactor system include the passive reactor shutdown characteristic and the passive shutdown heat removal system, RVACS. While these characteristics are simple in principle, the physical processes are fairly complex, particularly for the passive reactor shutdown. It has been possible to adapt independent safety analysis codes originally developed for the Clinch River Breeder Reactor review, although some limitations remain. In this paper, the analyses of postulated unscrammed events are discussed, along with limitations in the predictive capabilities and plans to correct the limitations in the near future. (author)

  18. 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)

  19. Overview of data acquisition system for SST-1 diagnostics

    International Nuclear Information System (INIS)

    Sharma, Manika; Mansuri, Imran; Raval, Tushar; Sharma, A.L; Pradhan, S.

    2016-01-01

    Highlights: • An account of architecture and data acquisition activities of SST-1 data acquisition system (DAS) for SST-1 diagnostics and subsystems. • PXI based Data acquisition system and CAMAC based Data acquisition system for slow and fast plasma diagnostics. • SST-1 DAS interface and its communication with SST-1 central control system. Integration of SST-1 DAS with timing system. • SST-1 DAS data archival and data analysis. - Abstract: The recent first phase operations of SST-1 in short pulse mode have provided an excellent opportunity for the essential initial tests and benchmark of the SST-1 Data Acquisition System. This paper describes the SST-1 Data Acquisition systems (DAS), which with its heterogeneous composition and distributed architecture, aims to cover a wide range of slow to fast channels interfaced with a large set of diagnostics. The DAS also provides the essential user interface for data acquisition to cater both on and off-line data usage. The central archiving and retrieval service is based on a dual step architecture involving a combination of Network Attached Server (NAS) and a Storage Area Network (SAN). SST-1 Data Acquisition Systems have been reliably operated in the SST-1 experimental campaigns. At present different distributed DAS caters the need of around 130 channels from different SST-1 diagnostics and its subsystems. PXI based DAS and CAMAC based DAS have been chosen to cater the need, with sampling rates varying from 10Ksamples/sec to 1Msamples/sec. For these large sets of channels acquiring from individual diagnostics and subsystems has been a combined setup, subjected to a gradual phase of optimization and tests resulting into a series of improvisations over the recent operations. In order to facilitate a reliable data acquisition, the model further integrates the objects of the systems with the Central Control System of SST-1 using the TCP/IP communication. The associated DAS software essentially addresses the

  20. Overview of data acquisition system for SST-1 diagnostics

    Energy Technology Data Exchange (ETDEWEB)

    Sharma, Manika, E-mail: bithi@ipr.res.in; Mansuri, Imran; Raval, Tushar; Sharma, A.L; Pradhan, S.

    2016-11-15

    Highlights: • An account of architecture and data acquisition activities of SST-1 data acquisition system (DAS) for SST-1 diagnostics and subsystems. • PXI based Data acquisition system and CAMAC based Data acquisition system for slow and fast plasma diagnostics. • SST-1 DAS interface and its communication with SST-1 central control system. Integration of SST-1 DAS with timing system. • SST-1 DAS data archival and data analysis. - Abstract: The recent first phase operations of SST-1 in short pulse mode have provided an excellent opportunity for the essential initial tests and benchmark of the SST-1 Data Acquisition System. This paper describes the SST-1 Data Acquisition systems (DAS), which with its heterogeneous composition and distributed architecture, aims to cover a wide range of slow to fast channels interfaced with a large set of diagnostics. The DAS also provides the essential user interface for data acquisition to cater both on and off-line data usage. The central archiving and retrieval service is based on a dual step architecture involving a combination of Network Attached Server (NAS) and a Storage Area Network (SAN). SST-1 Data Acquisition Systems have been reliably operated in the SST-1 experimental campaigns. At present different distributed DAS caters the need of around 130 channels from different SST-1 diagnostics and its subsystems. PXI based DAS and CAMAC based DAS have been chosen to cater the need, with sampling rates varying from 10Ksamples/sec to 1Msamples/sec. For these large sets of channels acquiring from individual diagnostics and subsystems has been a combined setup, subjected to a gradual phase of optimization and tests resulting into a series of improvisations over the recent operations. In order to facilitate a reliable data acquisition, the model further integrates the objects of the systems with the Central Control System of SST-1 using the TCP/IP communication. The associated DAS software essentially addresses the

  1. 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

  2. Approach to developing reliable space reactor power systems

    International Nuclear Information System (INIS)

    Mondt, J.F.; Shinbrot, C.H.

    1991-01-01

    The Space Reactor Power System Project is in the engineering development phase of a three-phase program. During Phase II, the Engineering Development Phase, the SP-100 Project has defined and is pursuing a new approach to developing reliable power systems. The approach to developing such a system during the early technology phase is described in this paper along with some preliminary examples to help explain the approach. Developing reliable components to meet space reactor power system requirements is based on a top down systems approach which includes a point design based on a detailed technical specification of a 100 kW power system

  3. Reactor coolant pressure boundary leakage detection system

    International Nuclear Information System (INIS)

    Dissing, E.; Svansson, L.

    1980-01-01

    This study deals with a system for monitoring the leakage of reactor coolant. This system is based primarily on the detection of the 13 N content in the containment atmosphere. 13 N is produced from the oxygen of the reactor water via the recoil proton nuclear process Hl+016/yields/ 13 N+ 4 He. The generation is therefore independent of fuel element leakage and of the corrosion product content in the water. It is solely related to the neutron flux level in the reactor core. Typical figures for the equilibrium 13 N concentration in the containment atmosphere following a 4 kg/minute coolant leakage are 5 kBq m/sup -3/ and 7 kBq m/sup -3/ for BWR and PWR respectively. These levels are readily measured with a 10 liter Ge(Li) flow detector assembly operated at elevated pressure. 8 refs

  4. Reactor coolant pressure boundary leakage detection system

    International Nuclear Information System (INIS)

    Dissing, E.; Svansson, L.

    1980-01-01

    This study deals with a system for monitoring the leakage of reactor coolant. This system is based primarily on the detection of the N13 content in the containment atmosphere. N13 is produced from the oxygen of the reactor water via the recoil proton nuclear process Hl+016/yields/Nl3+He4. The generation is therefore independent of fuel element leakage and of the corrosion product content in the water. It is solely related to the neutron flux level in the reactor core. Typical figures for the equilibrium N13 concentration in the containment atmosphere following a 4 kg/minute coolant leakage are 5 kBq m/sup -3/ and 7 kBq m/sup -3/ for BWR and PWR respectively. These levels are readily measured with a 10 liter Ge(Li) flow detector assembly operated at elevated pressure. 8 refs

  5. Reactor coolant pressure boundary leakage detection system

    International Nuclear Information System (INIS)

    Dissing, E.; Svansson, L.

    1979-08-01

    The present paper deals with a system for monitoring the leakage of reactor coolant. This system is based primarily on the detection of the N13 content in the containment atmosphere. N13 is produced from the oxygen of the reactor water via the recoil proton nuclear process H1+016 → N13+He4. The generation is therefore independent of fuel element leakage and of the corrosion product content in the water. It is solely related to the neutron flux level in the reactor core. Typical figures for the equilibrium N13 concentration in the containment atmosphere following a 4 kg/minute coolant leakage are 5 kBq m -3 and 7 kBq m -3 for BWR and PWR respectively. These levels are readily measured with a 10 liter Ge (Li) flow detector assembly operated at elevated pressure. (Auth.)

  6. 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

  7. Application of artificial intelligence to the detection and diagnosis of core faults in fast reactors

    International Nuclear Information System (INIS)

    Le Guillou, G.; Berlin, C.; Parcy, J.P.

    1983-01-01

    An automatic diagnostic system designed specially for fuel assembly accidents in fast reactors of the French type is described. It has an ''expert system'' structure and, at the present stage of the project, is operating on a laboratory scale in an experimental unit (Solar 16/65). The software aspects of the development of such systems (structure and formulation of rules, language, etc.) are presented in detail. Various simulation tests are outlined. ''Real-time'' application of the system will begin in 1983 on the Phenix reactor. (author)

  8. 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.)

  9. Performance Test for Neutron Detector and Associated System using Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Seongwoo; Park, Sung Jae; Cho, Man Soon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Oh, Se Hyun [USERS, Daejeon (Korea, Republic of); Shin, Ho Cheol [KHNP CRI, Daejeon (Korea, Republic of)

    2016-10-15

    SPND (Self-Powered Neutron Detector) has been developed to extend its lifespan. ENFMS (Ex-Core Flux Monitoring System) of pressurized water reactor has been also improved. After the development and improvement, their performance must be verified under the neutron irradiation environment. We used a research reactor for the performance verification of neutron detector and associated system because the research reactor can meet the neutron flux level of commercial nuclear reactor. In this paper, we report the performance verification method and result for the SPND and ENFMS using the research reactor. The performance tests for the SPND and ENFMS were conducted using UCI TRIGA reactor. The test environment of commercial reactor’s neutron flux level must be required. However, it is difficult to perform the test in the commercial rector due to the constraint of time and space. The research reactor can be good alternative neutron source for the test of neutron detectors and associated system.

  10. NEPTUNE: a modular system for light-water reactor calculation

    International Nuclear Information System (INIS)

    Bouchard, J.; Kanevoky, A.; Reuss, P.

    1975-01-01

    A complete modular system of light water reactor calculations has been designed. It includes basic nuclear data processing, the APOLLO phase: transport calculations for cells, multicells, fuel assemblies or reactors, the NEPTUNE phase: reactor calculations. A fuel management module, devoted to the automatic determination of the best shuffling strategy is included in NEPTUNE [fr

  11. A flexible simulator for training an early fault diagnostic system

    International Nuclear Information System (INIS)

    Marsiletti, M.; Santinelli, A.; Zuenkov, M.; Poletykin, A.

    1997-01-01

    An early fault diagnostic system has been developed addressed to timely trouble shooting in process plants during any operational modes. The theory of this diagnostic system is related with the usage of learning methods for automatic generation of knowledge bases. This approach enables the conversion of ''cause→effect'' relations into ''effect→possible-causes'' ones. The diagnostic rules are derived from the operation of a plant simulator according to a specific procedure. Flexibility, accuracy and high speed are the major characteristics of the training simulator, used to generate the diagnostic knowledge base. The simulator structure is very flexible, being based on LEGO code but allowing the use of practically any kind of FORTRAN routines (recently also ACSL macros has been introduced) as plant modules: this permits, when needed, a very accurate description of the malfunctions the diagnostic system should ''known''. The high speed is useful to shorten the ''learning'' phase of the diagnostic system. The feasibility of the overall system has been assessed, using as reference plant the conventional Sampierdarena (Italy) power station, that is a combined cycle plant dedicated to produce both electrical and heat power. The hardware configuration of this prototype system was made up of a network of a Hewlett-Packard workstation and a Digital VAX-Station. The paper illustrates the basic structure of the simulator used for this diagnostic system training purpose, as well as the theoretical background on which the diagnostic system is based. Some evidence of the effectiveness of the concept through the application to Sampierdarena 40 MW cogeneration plant is reported. Finally an outline of an ongoing application to a WWER-1000 plant is given; the operating system is, in this case, UNIX. (author)

  12. Ultrasonic thermometry system for measuring very high temperatures in reactor safety experiments

    International Nuclear Information System (INIS)

    Carlson, G.A.; Sullivan, W.H.; Plein, H.G.; Kerley, T.M.

    1979-06-01

    Ultrasonic thermometry has many potential applications in reactor safety experiments, where extremely high temperatures and lack of visual access may preclude the use of conventional diagnostics. This report details ultrasonic thermometry requirements for one such experiment, the molten fuel pool experiment. Sensors, transducers, and signal processing electronics are described in detail. Axial heat transfer in the sensors is modelled and found acceptably small. Measurement errors, calculations of their effect, and ways to minimize them are given. A rotating sensor concept is discussed which holds promise of alleviating sticking problems at high temperature. Applications of ultrasonic thermometry to three in-core experiments are described. In them, five 10-mm-length sensor elements were used to measure axial temperatures in a UO 2 or UO 2 -steel system fission-heated to about 2860 0 C

  13. System Design of a Supercritical CO_2 cooled Micro Modular Reactor

    International Nuclear Information System (INIS)

    Kim, Seong Gu; Cho, Seongkuk; Yu, Hwanyeal; Kim, Yonghee; Jeong, Yong Hoon; Lee, Jeong Ik

    2014-01-01

    Small modular reactor (SMR) systems that have advantages of little initial capital cost and small restriction on construction site are being developed by many research organizations around the world. Existing SMR concepts have the same objective: to achieve compact size and a long life core. Most of small modular reactors have much smaller size than the large nuclear power plant. However, existing SMR concepts are not fully modularized. This paper suggests a complete modular reactor with an innovative concept for reactor cooling by using a supercritical carbon dioxide. The authors propose the supercritical CO_2 Brayton cycle (S-CO_2 cycle) as a power conversion system to achieve small volume of power conversion unit (PCU) and to contain the reactor core and PCU in one vessel. A conceptual design of the proposed small modular reactor was developed, which is named as KAIST Micro Modular Reactor (MMR). The supercritical CO_2 Brayton cycle for the S-CO_2 cooled reactor core was optimized and the size of turbomachinery and heat exchanger were estimated preliminary. The nuclear fuel composed with UN was proposed and the core lifetime was obtained from a burnup versus reactivity calculation. Furthermore, a system layout with fully passive safety systems for both normal operation and emergency operation was proposed. (author)

  14. Plenum separator system for pool-type nuclear reactors

    International Nuclear Information System (INIS)

    Sharbaugh, J.E.

    1983-01-01

    This invention provides a plenum separator system for pool-type nuclear reactors which substantially lessens undesirable thermal effects on major components. A primary feature of the invention is the addition of one or more intermediate plena, containing substantially stagnant and stratified coolant, which separate the hot and cold plena and particularly the hot plena from critical reactor components. This plenum separator system also includes a plurality of components which together form a dual pass flow path annular region spaced from the reactor vessel wall by an annular gas space. The bypass flow through the flow path is relatively small and is drawn from the main coolant pumps and discharged to an intermediate plenum

  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. An expert system for turbogenerator diagnostics

    International Nuclear Information System (INIS)

    Bessenyei, Z.; Tomcsanyi, T.; Toth, Z.; Laczay, I.

    1992-01-01

    In 1990, an expert system for turbo-generator diagnostics (EST-D) was installed at the 3rd and 4th units of the Paks NPP (Hungary). The expert system is strongly integrated to the ARGUS II vibration monitoring and diagnostics system. The system works on IBM PC AT. The VEIKI's and the NPP's human experts were interviewed to fill up the knowledgebase. The system is able to identify 13 different faults of the parts of a turbogenerator. The knowledgebase consists of ca 200 rules. The rules were built in and the system was verified and validated using a model of the turbines and using the experiences gathered with ARGUS II during the last 3 years. The maintenance personnel is authorized to modify and/or extend the knowledgebase. The input data for evaluation come from measured vibration patterns produced by the ARGUS II system, database of events, and maintenance data input by the maintenance personnel. The expert system is based on the modified GENESYS 2.1 shell (developed by SZAMALK, Hungary). Some limitations from PC application were eliminated, and a new, independent explanation module and man-machine interface were developed. Using this man-machine interface, one of the basic goals of the expert system developments was achieved: the human experts contribution is not necessary for diagnoses. The operator of the diagnostics system is able to produce the reports of diagnoses. Of course the interface allows the human experts to see the diagnoses through. It should be mentioned, at the beginning of 1991, we installed a similar expert system at the 1st 1000 MW WWER type unit of the Kalinin NPP (Soviet Union). In this paper, the operation of the EST-D, the man-machine interface and the operational experiences of the first 4 months work are explained. 2 refs., 14 figs

  17. 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.)

  18. Clinical applications of SONIALVISION 100 digital diagnostic table system

    International Nuclear Information System (INIS)

    Shiomi, Takeshi; Shimizu, Tatsuya; Iinuma, Masao; Takemoto, Hajime; Tanaka, Shuji

    2003-01-01

    This report refers to the clinical applications of our newly developed SONIALVISION 100 fully digitalized X-ray diagnostic table system. The main design concept of the SONIALVISION 100 system is the improvement of workflow in various clinical fields. The development of digital imaging technologies has come to allow fully digitalized X-ray diagnostic table systems to be widely utilized in various clinical applications, including interventional radiology (IVR) and examinations using contrast medium. This report mainly refers to the clinical applications of the Shimadzu SONIALVISION 100 digitalized X-ray diagnostic table system, also presenting some typical image data demonstrating the high efficiency, made available through the use of this new system, in high-speed spot imaging and digital tomography. (author)

  19. System modeling and reactor design studies of the Advanced Thermionic Initiative space nuclear reactor

    International Nuclear Information System (INIS)

    Lee, H.H.; Abdul-Hamid, S.; Klein, A.C.

    1996-01-01

    In-core thermionic space reactor design concepts that operate at a nominal power output range of 20 to 50 kW(electric) are described. Details of the neutronic, thermionic, thermal hydraulics, and shielding performance are presented. Because of the strong absorption of thermal neutrons by natural tungsten and the large amount of natural tungsten within the reactor core, two designs are considered. An overall system design code has been developed at Oregon State University to model advanced in-core thermionic energy conversion-based nuclear reactor systems for space applications. The results show that the driverless single-cell Advanced Thermionic Initiative (ATI) configuration, which does not have driver fuel rods, proved to be more efficient than the driven core, which has driver rods. The results also show that the inclusion of the true axial and radial power distribution decrease the overall conversion efficiency. The flattening of the radial power distribution by three different methods would lead to a higher efficiency. The results show that only one TFE works at the optimum emitter temperature; all other TFEs are off the optimum performance and result in a 40% decrease of the efficiency of the overall system. The true axial profile is significantly different as there is a considerable amount of neutron leakage out of the top and bottom of the reactor. The analysis reveals that the axial power profile actually has a chopped cosine shape. For this axial profile, the reactor core overall efficiency for the driverless ATI reactor version is found to be 5.84% with a total electrical power of 21.92 kW(electric). By considering the true axial power profile instead of the uniform power profile, each TFE loses ∼80 W(electric)

  20. Modelling aerosol behavior in reactor cooling systems

    International Nuclear Information System (INIS)

    McDonald, B.H.

    1990-01-01

    This paper presents an overview of some of the areas of concern in using computer codes to model fission-product aerosol behavior in the reactor cooling system (RCS) of a water-cooled nuclear reactor during a loss-of-coolant accident. The basic physical processes that require modelling include: fission product release and aerosol formation in the reactor core, aerosol transport and deposition in the reactor core and throughout the rest of the RCS, and the interaction between aerosol transport processes and the thermalhydraulics. In addition to these basic physical processes, chemical reactions can have a large influence on the nature of the aerosol and its behavior in the RCS. The focus is on the physics and the implications of numerical methods used in the computer codes to model aerosol behavior in the RCS

  1. Assessment of core protection and monitoring systems for an advanced reactor SMART

    International Nuclear Information System (INIS)

    In, Wang Kee; Hwang, Dae Hyun; Yoo, Yeon Jong; Zee, Sung Qunn

    2002-01-01

    Analogue and digital core protection/monitoring systems were assessed for the implementation in an advanced reactor. The core thermal margins to nuclear fuel design limits (departure from nucleate boiling and fuel centerline melting) were estimated using the design data for a commercial pressurized water reactor and an advanced reactor. The digital protection system resulted in a greater power margin to the fuel centerline melting by at least 30% of rated power for both commercial and advanced reactors. The DNB margin with the digital system is also higher than that for the analogue system by 8 and 12.1% of rated power for commercial and advanced reactors, respectively. The margin gain with the digital system is largely due to the on-line calculations of DNB ratio and peak local power density from the live sensor signals. The digital core protection and monitoring systems are, therefore, believed to be more appropriate for the advanced reactor

  2. Nonlinear dynamic analysis of nuclear reactor primary coolant systems

    International Nuclear Information System (INIS)

    Saffell, B.F. Jr.; Macek, R.W.; Thompson, T.R.; Lippert, R.F.

    1979-01-01

    The ADINA computer code is utilized to perform mechanical response analysis of pressurized reactor primary coolant systems subjected to postulated loss-of-coolant accident (LOCA) loadings. Specifically, three plant analyses are performed utilizing the geometric and material nonlinear analysis capabilities of ADINA. Each reactor system finite element model represents the reactor vessel and internals, piping, major components, and component supports in a single coupled model. Material and geometric nonlinear capabilities of the beam and truss elements are employed in the formulation of each finite element model. Loadings applied to each plant for LOCA dynamic analysis include steady-state pressure, dead weight, strain energy release, transient piping hydraulic forces, and reactor vessel cavity pressurization. Representative results are presented with some suggestions for consideration in future ADINA code development

  3. Development of intellectual reactor design system IRDS

    International Nuclear Information System (INIS)

    Kugo, T.; Tsuchihashi, K.; Nakagawa, M.; Mori, T.

    1993-01-01

    An intellectual reactor design system IRDS has been developed to support feasibility study and conceptual design of new type reactors in the fields of reactor core design including neutronics, thermal-hydraulics and fuel design. IRDS is an integrated software system in which a variety of computer codes in the different fields are installed. An integration of simulation modules are performed by the information transfer between modules through design model in which the design information of the current design work is stored. An object oriented architecture is realized in frame representation of core configuration in a design data base. The knowledge relating to design tasks to be performed are encapsulated, to support the conceptual design work. The system is constructed on an engineering workstation, and supports efficiently design work through man-machine interface adopting the advanced information processing technologies. Optimization methods for design parameters with use of the artificial intelligence technique are now under study, to reduce the parametric study work. A function to search design window in which design is feasible is realized in the fuel pin design. (orig.)

  4. System for unattended surveillance of nuclear reactor behavior

    International Nuclear Information System (INIS)

    Gonzalez, R.C.; Howington, L.C.

    1977-01-01

    A multivariate statistical pattern recognition system for reactor noise analysis is presented. The basis of the system is a transformation for decoupling correlated variables and algorithms for inferring probability density functions. The system is adaptable to a variety of statistical properties of the data, and it has learning, tracking, updating, and dimensionality reduction capabilities. System design emphasizes control of the false-alarm rate. Its abilities to learn normal patterns and to recognize deviations from these patterns were evaluated by experiments at the ORNL High-Flux Isotope Reactor. Power perturbations of less than 0.1% of the mean value in selected frequency ranges were readily detected by the pattern recognition system

  5. Closed Brayton cycle power conversion systems for nuclear reactors :

    Energy Technology Data Exchange (ETDEWEB)

    Wright, Steven A. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Lipinski, Ronald J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Vernon, Milton E. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sanchez, Travis [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2006-04-01

    This report describes the results of a Sandia National Laboratories internally funded research program to study the coupling of nuclear reactors to gas dynamic Brayton power conversion systems. The research focused on developing integrated dynamic system models, fabricating a 10-30 kWe closed loop Brayton cycle, and validating these models by operating the Brayton test-loop. The work tasks were performed in three major areas. First, the system equations and dynamic models for reactors and Closed Brayton Cycle (CBC) systems were developed and implemented in SIMULINKTM. Within this effort, both steady state and dynamic system models for all the components (turbines, compressors, reactors, ducting, alternators, heat exchangers, and space based radiators) were developed and assembled into complete systems for gas cooled reactors, liquid metal reactors, and electrically heated simulators. Various control modules that use proportional-integral-differential (PID) feedback loops for the reactor and the power-conversion shaft speed were also developed and implemented. The simulation code is called RPCSIM (Reactor Power and Control Simulator). In the second task an open cycle commercially available Capstone C30 micro-turbine power generator was modified to provide a small inexpensive closed Brayton cycle test loop called the Sandia Brayton test-Loop (SBL-30). The Capstone gas-turbine unit housing was modified to permit the attachment of an electrical heater and a water cooled chiller to form a closed loop. The Capstone turbine, compressor, and alternator were used without modification. The Capstone systems nominal operating point is 1150 K turbine inlet temperature at 96,000 rpm. The annular recuperator and portions of the Capstone control system (inverter) and starter system also were reused. The rotational speed of the turbo-machinery is controlled by adjusting the alternator load by using the electrical grid as the load bank. The SBL-30 test loop was operated at

  6. Power noise spectrum classification in the problem of the IBR-2 reactor

    International Nuclear Information System (INIS)

    Bargel, M.; Kitowski, J.; Pepelyshev, Yu.N.

    1988-01-01

    The classification spectrum results of random fluctuations in the IBR-2 energy pulse are presented. The work is performed for the application of the obtained results to the reactor diagnostics and the study of its noise uncontrolled states. For classification of the spectra the method of pattern recognition based upon the ISODATA heuristic algorithm is used. It is shown that a set of noise uncontrolled reactor states, registered during the reactor operation period at power of 0.4-2 MVt with the first variant of moving reflector (1983-1986) is formed into 4(5) most typical states. Each of the states corresponds to the general conditions of the reactor core cooling and provides the normal work of the moving reflector. However, these states differ in coolant flow, power level and peculiarities of the moving reflector rotation regime. One type of anomal power noise, connected with some disorder in the moving reflctor work, is isolated. This work also presents the possibility of control over the state of moving reflectors according to the change in the amplitude of power oscillations at some frequences. The reactor noise classification results can be used as the data bank for the IBR-2 reactor diagnostic system

  7. 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)

  8. Prism reactor system design and analysis of postulated unscrammed events

    International Nuclear Information System (INIS)

    Van Tuyle, G.J.; Slovik, G.C.

    1991-08-01

    Key safety characteristics of the PRISM reactor system include the passive reactor shutdown characteristic and the passive shutdown heat removal system, RVACS. While these characteristics are simple in principle, the physical processes are fairly complex, particularly for the passive reactor shutdown. It has been possible to adapt independent safety analysis codes originally developed for the Clinch River Breeder Reactor review, although some limitations remain. In this paper, the analyses of postulated unscrammed events are discussed, along with limitations in the predictive capabilities and plans to correct the limitations in the near future. 6 refs., 4 figs

  9. PRISM reactor system design and analysis of postulated unscrammed events

    International Nuclear Information System (INIS)

    Van Tuyle, G.J.; Slovik, G.C.; Rosztoczy, Z.; Lane, J.

    1991-01-01

    Key safety characteristics of the PRISM reactor system include the passive reactor shutdown characteristics and the passive shutdown heat removal system, RVACS. While these characteristics are simple in principle, the physical processes are fairly complex, particularly for the passive reactor shutdown. It has been possible to adapt independent safety analysis codes originally developed for the Clinch River Breeder Reactor review, although some limitations remain. In this paper, the analyses of postulated unscrammed events are discussed, along with limitations in the predictive capabilities and plans to correct the limitations in the near future. 6 refs., 4 figs

  10. Advanced Light Source beam diagnostics systems

    International Nuclear Information System (INIS)

    Hinkson, J.

    1993-10-01

    The Advanced Light Source (ALS), a third-generation synchrotron light source, has been recently commissioned. Beam diagnostics were very important to the success of the operation. Each diagnostic system is described in this paper along with detailed discussion of its performance. Some of the systems have been in operation for two years. Others, in the storage ring, have not yet been fully commissioned. These systems were, however, working well enough to provide the essential information needed to store beam. The devices described in this paper include wall current monitors, a beam charge monitor, a 50 ohm Faraday cup, DC current transformers, broad-hand striplines, fluorescence screens, beam collimators and scrapers, and beam position monitors. Also, the means by which waveforms are digitized and displayed in the control room is discussed

  11. A protection system of low temperature thermo-supply nuclear reactor

    International Nuclear Information System (INIS)

    Jiang Binsen

    1988-09-01

    A Protection system of low temperature thermo-supply nuclear reactor is introduced. It is the first protection system, which is designed and manufactred on the basis of Chinese National Standard GB 4083-83 'General Safety Principle of Nuclear Reactor Protection System', to be considered under the circumstances of industry level in China. Advantages of the protection system are as follows: 1)The single failure criteria can fully be fulfilled by the protection system. 2) On-line testing system can be used for detecting all of failure components and quick identifying the failure points in the system. 3) It is convenience for maintenacnce of the system. To complete this project is very important and helpful in promoting the development of the protection system and safety operation of nuclear reactor in China

  12. System for chemical decontamination of nuclear reactor primary systems

    International Nuclear Information System (INIS)

    Schlonski, J.S.; McGiure, M.F.; Corpora, G.J.

    1992-01-01

    This patent describes a method of chemically decontaminating a nuclear reactor primary system, having a residual heat removal system with one or more residual heat removal heat exchangers, each having an upstream and a downstream side, at or above ambient pressure. It comprises: injecting decontamination chemicals using an injection means; circulating the injected decontamination chemicals throughout the primary system; directing the circulated decontamination chemicals and process fluids to a means for removing suspended solids and dissolved materials after the circulated chemicals and process fluids have passed through the residual heat removal heat exchanger; decontaminating the process fluids; and feeding the decontaminated process fluids to the injection means. This patent also describes a chemical decontamination system for use at, or above, ambient pressure in a nuclear reactor primary system having a residual heat removal system. It comprises: means for injecting decontamination chemicals into the primary system; means for removing dissolved and suspended materials and decontamination chemicals from the primary system; one or more residual heat removal pumps; means located downstream of one of the residual heat removal heat exchangers; and a return line connecting the means

  13. Multi-objective optimization of the reactor coolant system

    International Nuclear Information System (INIS)

    Chen Lei; Yan Changqi; Wang Jianjun

    2014-01-01

    Background: Weight and size are important criteria in evaluating the performance of a nuclear power plant. It is of great theoretical value and engineering significance to reduce the weight and volume of the components for a nuclear power plant by the optimization methodology. Purpose: In order to provide a new method for the optimization of nuclear power plant multi-objective, the concept of the non-dominated solution was introduced. Methods: Based on the parameters of Qinshan I nuclear power plant, the mathematical models of the reactor core, the reactor vessel, the main pipe, the pressurizer and the steam generator were built and verified. The sensitivity analyses were carried out to study the influences of the design variables on the objectives. A modified non-dominated sorting genetic algorithm was proposed and employed to optimize the weight and the volume of the reactor coolant system. Results: The results show that the component mathematical models are reliable, the modified non-dominated sorting generic algorithm is effective, and the reactor inlet temperature is the most important variable which influences the distribution of the non-dominated solutions. Conclusion: The optimization results could provide a reference to the design of such reactor coolant system. (authors)

  14. 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

  15. Vibration test on KMRR reactor structure and primary cooling system piping

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Seung Hoh; Kim, Tae Ryong; Park, Jin Hoh; Park, Jin Suk; Ryoo, Jung Soo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1994-10-01

    Most equipments, piping systems and reactor structures in nuclear power plants are subjected to flow induced vibration due to high temperature and high pressure coolant flowing inside or outside of the equipments, systems and structures. Because the flow induced vibration sometimes causes significant damage to reactor structures and piping systems, it is important and necessary to evaluate the vibration effect on them and to prove their structural integrity. Korea Multipurpose Research Reactor (KMRR) being constructed by KAERI is 30 MWt pool type research reactor. Since its main structures and piping systems were designed and manufactured in accordance with the standards and guidelines for commercial nuclear power plant, it was decided to evaluate their vibratory response in accordance with the standards and guidelines for commercial NPP. The objective of this vibration test is the assessment of vibration levels of KMRR reactor structure and primary cooling piping system for their structural integrity under the steady-state or transient operating condition. 38 figs, 14 tabs, 2 refs. (Author).

  16. Vibration test on KMRR reactor structure and primary cooling system piping

    International Nuclear Information System (INIS)

    Chung, Seung Hoh; Kim, Tae Ryong; Park, Jin Hoh; Park, Jin Suk; Ryoo, Jung Soo

    1994-10-01

    Most equipments, piping systems and reactor structures in nuclear power plants are subjected to flow induced vibration due to high temperature and high pressure coolant flowing inside or outside of the equipments, systems and structures. Because the flow induced vibration sometimes causes significant damage to reactor structures and piping systems, it is important and necessary to evaluate the vibration effect on them and to prove their structural integrity. Korea Multipurpose Research Reactor (KMRR) being constructed by KAERI is 30 MWt pool type research reactor. Since its main structures and piping systems were designed and manufactured in accordance with the standards and guidelines for commercial nuclear power plant, it was decided to evaluate their vibratory response in accordance with the standards and guidelines for commercial NPP. The objective of this vibration test is the assessment of vibration levels of KMRR reactor structure and primary cooling piping system for their structural integrity under the steady-state or transient operating condition. 38 figs, 14 tabs, 2 refs. (Author)

  17. Development of the next generation reactor analysis code system, MARBLE

    International Nuclear Information System (INIS)

    Yokoyama, Kenji; Hazama, Taira; Nagaya, Yasunobu; Chiba, Go; Kugo, Teruhiko; Ishikawa, Makoto; Tatsumi, Masahiro; Hirai, Yasushi; Hyoudou, Hideaki; Numata, Kazuyuki; Iwai, Takehiko; Jin, Tomoyuki

    2011-03-01

    A next generation reactor analysis code system, MARBLE, has been developed. MARBLE is a successor of the fast reactor neutronics analysis code systems, JOINT-FR and SAGEP-FR (conventional systems), which were developed for so-called JUPITER standard analysis methods. MARBLE has the equivalent analysis capability to the conventional system because MARBLE can utilize sub-codes included in the conventional system without any change. On the other hand, burnup analysis functionality for power reactors is improved compared with the conventional system by introducing models on fuel exchange treatment and control rod operation and so on. In addition, MARBLE has newly developed solvers and some new features of burnup calculation by the Krylov sub-space method and nuclear design accuracy evaluation by the extended bias factor method. In the development of MARBLE, the object oriented technology was adopted from the view-point of improvement of the software quality such as flexibility, expansibility, facilitation of the verification by the modularization and assistance of co-development. And, software structure called the two-layer system consisting of scripting language and system development language was applied. As a result, MARBLE is not an independent analysis code system which simply receives input and returns output, but an assembly of components for building an analysis code system (i.e. framework). Furthermore, MARBLE provides some pre-built analysis code systems such as the fast reactor neutronics analysis code system. SCHEME, which corresponds to the conventional code and the fast reactor burnup analysis code system, ORPHEUS. (author)

  18. Fuel management of mixed reactor type power plant systems

    International Nuclear Information System (INIS)

    Csom, Gyula

    1988-01-01

    Breeding gain in symbiotic nuclear power plant system consisting of both thermal and fast breeder reactors depends on the characteristics and the ratio of thermal and fast reactors. The composition of the symbiotic power plant systems was determined for equilibrium and plutonium deficient systems. According to natural uranium utilization, symbiotic power plant systems are not less efficient than the systems containing only fast breeders. Depleted uranium can be applied in both types of systems. Reprocessing demands of the symbiotic power plant sytems were determined. (V.N.) 23 figs.; 1 tab

  19. Lightning protection system analysis at Multipurpose Reactor G A. Siwabessy building

    International Nuclear Information System (INIS)

    Teguh-Sulistyo

    2003-01-01

    Analysis to the part of lightning protection system at Multi Purpose Reactor GA Siwabessy (RSG-GAS) have been done. Observation examined the damage of some part of the earthing system caused by human error of chemically system. The analysis performed some assumptions and simulations to the points of lightning stroke. From this analysis obtained that the reactor building do not have vertical finial which can protect effectively to the whole reactor building and auxiliary building. Installing some new finials at some places are needed to protect building therefore the reactor building and auxiliary building well safe from lighting stroke

  20. Real-time numerical simulation with high efficiency for an experimental reactor system

    International Nuclear Information System (INIS)

    Ding Shuling; Li Fu; Li Sifeng; Chu Xinyuan

    2006-01-01

    The paper presents a systematic and efficient method for numerical real-time simulation of an experimental reactor. The reactor models were built based on the physical characteristics of the experimental reactor, and several real-time simulation approaches were discussed and compared in the paper. How to implement the real-time reactor simulation system in Windows platform for the sake of hardware-in-loop experiment for the reactor power control system was discussed. (authors)