WorldWideScience

Sample records for safety control systems

  1. Safety implications of control systems

    International Nuclear Information System (INIS)

    Smith, O.L.

    1983-01-01

    The Safety Implications of Control Systems Program has three major activities in support of USI-A47. The first task is a failure mode and effects analysis of all plant systems which may potentially induce control system disturbance that have safety implications. This task has made a preliminary study of overfill events and recommended cases for further analysis on the hybrid simulator. Work continues on overcooling and undercooling. A detailed investigation of electric power network is in progress. LERs are providing guidance on important failure modes that will provide initial conditions for further simulator studies. The simulator taks is generating a detailed model of the control system supported by appropriate neutronics, hydraulics, and thermodynamics submodels of all other principal plant components. The simulator is in the last stages of development. Checkout calculations are in progress to establish model stability, robustness, and qualitative credibility. Verification against benchmark codes and plant data will follow

  2. FOOD SAFETY CONTROL SYSTEM IN CHINA

    Institute of Scientific and Technical Information of China (English)

    Liu Wei-jun; Wei Yi-min; Han Jun; Luo Dan; Pan Jia-rong

    2007-01-01

    Most countries have expended much effort to develop food safety control systems to ensure safe food supplies within their borders. China, as one of the world's largest food producers and consumers,pays a lot of attention to food safety issues. In recent years, China has taken actions and implemented a series of plans in respect to food safety. Food safety control systems including regulatory, supervisory,and science and technology systems, have begun to be established in China. Using, as a base, an analysis of the current Chinese food safety control system as measured against international standards, this paper discusses the need for China to standardize its food safety control system. We then suggest some policies and measures to improve the Chinese food safety control system.

  3. Safety-related control air systems

    International Nuclear Information System (INIS)

    Anon.

    1977-01-01

    This Standard applies to those portions of the control air system that furnish air required to support, control, or operate systems or portions of systems that are safety related in nuclear power plants. This Standard relates only to the air supply system(s) for safety-related air operated devices and does not apply to the safety-related air operated device or to air operated actuators for such devices. The objectives of this Standard are to provide (1) minimum system design requirements for equipment, piping, instruments, controls, and wiring that constitute the air supply system; and (2) the system and component testing and maintenance requirements

  4. Safety Management System in Croatia Control Ltd.

    OpenAIRE

    Pavlin, Stanislav; Sorić, Vedran; Bilać, Dragan; Dimnik, Igor; Galić, Daniel

    2009-01-01

    International Civil Aviation Organization and other international aviation organizations regulate the safety in civil aviation. In the recent years the International Civil Aviation Organization has introduced the concept of the safety management system through several documents among which the most important is the 2006 Safety Management Manual. It treats the safety management system in all the segments of civil aviation, from carriers, aerodromes and air traffic control to design, constructi...

  5. Assessment of Safety Standards for Automotive Electronic Control Systems

    Science.gov (United States)

    2016-06-01

    This report summarizes the results of a study that assessed and compared six industry and government safety standards relevant to the safety and reliability of automotive electronic control systems. These standards include ISO 26262 (Road Vehicles - ...

  6. Development of digital safety system logic and control

    International Nuclear Information System (INIS)

    Nishikawa, H.; Sakamoto, H.

    1995-01-01

    Advanced-BWR (ABWR) uses total digital control and instrumentation (C and I) system. In particular, ABWR adopts a newly developed safety system using advanced digital technology. In the presentation the digital safety system design, manufacturing and factory validation test method are shortly overviewed. The digital safety system consists of micro-processor based digital controllers, data and information transmission by optical fibers and human-machine interface using color flat displays. This new developed safety system meet the nuclear safety requirements such as high reliability, independence of divisions, operability and maintainability. (2 refs., 4 figs., 1 tab.)

  7. Safety-related control air systems - approved 1977

    International Nuclear Information System (INIS)

    Anon.

    1978-01-01

    This standard applies to those portions of the control air system that furnish air required to support, control, or operate systems or portions of systems that are safety related in nuclear power plants. This standard relates only to the air supply system(s) for safety-related air operated devices and does not apply to the safety-related air operated device or to air operated actuators for such devices. The objectives of this standard are to provide (1) minimum system design requirements for equipment, piping, instruments, controls, and wiring that constitute the air supply system; and (2) the system and component testing and maintenance requirements

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

    International Nuclear Information System (INIS)

    El-Shanshoury, A.I.

    2009-01-01

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

  9. Instrumentation and control systems important to safety in nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2005-01-01

    This Safety Guide was prepared under the IAEA programme for establishing safety standards for nuclear power plants. It supplements Safety Standards Series No. NS-R-1: Safety of Nuclear Power Plants: Design (the Requirements for Design), which establishes the design requirements for ensuring the safety of nuclear power plants. This Safety Guide describes how the requirements should be met for instrumentation and control (I and C) systems important to safety. This publication is a revision and combination of two previous Safety Guides: Safety Series Nos 50-SG-D3 and 50-SG-D8, which are superseded by this new Safety Guide. The revision takes account of developments in I and C systems important to safety since the earlier Safety Guides were published in 1980 and 1984, respectively. The objective of this Safety Guide is to provide guidance on the design of I and C systems important to safety in nuclear power plants, including all I and C components, from the sensors allocated to the mechanical systems to the actuated equipment, operator interfaces and auxiliary equipment. This Safety Guide deals mainly with design requirements for those I and C systems that are important to safety. It expands on paragraphs of Ref in the area of I and C systems important to safety. This publication is intended for use primarily by designers of nuclear power plants and also by owners and/or operators and regulators of nuclear power plants. This Safety Guide provides general guidance on I and C systems important to safety which is broadly applicable to many nuclear power plants. More detailed requirements and limitations for safe operation specific to a particular plant type should be established as part of the design process. The present guidance is focused on the design principles for systems important to safety that warrant particular attention, and should be applied to both the design of new I and C systems and the modernization of existing systems. Guidance is provided on how design

  10. A new radiation safety control system for Ganil

    International Nuclear Information System (INIS)

    Saint Jores, P. De; Luong, T.T.; Martina, L.; Vega, G.

    1991-01-01

    A second generation radiation safety control system has been installed to upgrade the initial system which was not flexible enough to support new ion beams and new experimental conditions required by the accelerator operation. The main reasons which necessitated the improvement of the safety control system are presented. The new system which controls the Ganil accelerator from the first quarter of 1990 is described. It uses a star structured architecture, VME standard processors and front-end modules activated by pDOS operating system and high level language (C and Fortran) tasks, associated with enhanced resolution color displays for real time synoptics. (R.P.) 4 refs., 4 figs

  11. Quantitative safety assessment of air traffic control systems through system control capacity

    Science.gov (United States)

    Guo, Jingjing

    Quantitative Safety Assessments (QSA) are essential to safety benefit verification and regulations of developmental changes in safety critical systems like the Air Traffic Control (ATC) systems. Effectiveness of the assessments is particularly desirable today in the safe implementations of revolutionary ATC overhauls like NextGen and SESAR. QSA of ATC systems are however challenged by system complexity and lack of accident data. Extending from the idea "safety is a control problem" in the literature, this research proposes to assess system safety from the control perspective, through quantifying a system's "control capacity". A system's safety performance correlates to this "control capacity" in the control of "safety critical processes". To examine this idea in QSA of the ATC systems, a Control-capacity Based Safety Assessment Framework (CBSAF) is developed which includes two control capacity metrics and a procedural method. The two metrics are Probabilistic System Control-capacity (PSC) and Temporal System Control-capacity (TSC); each addresses an aspect of a system's control capacity. And the procedural method consists three general stages: I) identification of safety critical processes, II) development of system control models and III) evaluation of system control capacity. The CBSAF was tested in two case studies. The first one assesses an en-route collision avoidance scenario and compares three hypothetical configurations. The CBSAF was able to capture the uncoordinated behavior between two means of control, as was observed in a historic midair collision accident. The second case study compares CBSAF with an existing risk based QSA method in assessing the safety benefits of introducing a runway incursion alert system. Similar conclusions are reached between the two methods, while the CBSAF has the advantage of simplicity and provides a new control-based perspective and interpretation to the assessments. The case studies are intended to investigate the

  12. Automated Systems for Road Safety control in a Developing World ...

    African Journals Online (AJOL)

    An Automated system was finally designed and developed for road safety control. This Automated system is believed to have the capacity to minimize or eliminate the problems identified in this study on traffic control in a developing world. Key words: drivers, traffic situation information, accident causation, FRSC ...

  13. Process Control Systems in the Chemical Industry: Safety vs. Security

    Energy Technology Data Exchange (ETDEWEB)

    Jeffrey Hahn; Thomas Anderson

    2005-04-01

    Traditionally, the primary focus of the chemical industry has been safety and productivity. However, recent threats to our nation’s critical infrastructure have prompted a tightening of security measures across many different industry sectors. Reducing vulnerabilities of control systems against physical and cyber attack is necessary to ensure the safety, security and effective functioning of these systems. The U.S. Department of Homeland Security has developed a strategy to secure these vulnerabilities. Crucial to this strategy is the Control Systems Security and Test Center (CSSTC) established to test and analyze control systems equipment. In addition, the CSSTC promotes a proactive, collaborative approach to increase industry's awareness of standards, products and processes that can enhance the security of control systems. This paper outlines measures that can be taken to enhance the cybersecurity of process control systems in the chemical sector.

  14. Safety Evaluation Approach with Security Controls for Safety I and C Systems on Nuclear Power Plants

    International Nuclear Information System (INIS)

    Kim, D. H.; Jeong, S. Y.; Kim, Y. M.; Park, H. S.; Lee, M. S.; Kim, T. H.

    2016-01-01

    This paper addresses concepts of safety and security and relations between them for assessing effects of security features in safety systems. Also, evaluation approach for avoiding confliction with safety requirements and cyber security features which may be adopted in safety-related digital I and C system will be described. In this paper, safety-security life cycle model based confliction avoidance method was proposed to evaluate the effects when the cyber security control features are implemented in the safety I and C system. Also, safety effect evaluation results using the proposed evaluation method were described. In case of technical security controls, many of them are expected to conflict with safety requirements, otherwise operational and managerial controls are not relatively. Safety measures and cyber security measures for nuclear power plants should be implemented not to conflict with one another. Where safety function and security features are both required within the systems, and also where security features are implemented within safety systems, they should be justified

  15. Safety Evaluation Approach with Security Controls for Safety I and C Systems on Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Kim, D. H.; Jeong, S. Y.; Kim, Y. M.; Park, H. S. [KINS, Daejeon (Korea, Republic of); Lee, M. S.; Kim, T. H. [Formal Works Inc., Seoul (Korea, Republic of)

    2016-05-15

    This paper addresses concepts of safety and security and relations between them for assessing effects of security features in safety systems. Also, evaluation approach for avoiding confliction with safety requirements and cyber security features which may be adopted in safety-related digital I and C system will be described. In this paper, safety-security life cycle model based confliction avoidance method was proposed to evaluate the effects when the cyber security control features are implemented in the safety I and C system. Also, safety effect evaluation results using the proposed evaluation method were described. In case of technical security controls, many of them are expected to conflict with safety requirements, otherwise operational and managerial controls are not relatively. Safety measures and cyber security measures for nuclear power plants should be implemented not to conflict with one another. Where safety function and security features are both required within the systems, and also where security features are implemented within safety systems, they should be justified.

  16. Operation safety of control systems. Principles and methods

    International Nuclear Information System (INIS)

    Aubry, J.F.; Chatelet, E.

    2008-01-01

    This article presents the main operation safety methods that can be implemented to design safe control systems taking into account the behaviour of the different components with each other (binary 'operation/failure' behaviours, non-consistent behaviours and 'hidden' failures, dynamical behaviours and temporal aspects etc). To take into account these different behaviours, advanced qualitative and quantitative methods have to be used which are described in this article: 1 - qualitative methods of analysis: functional analysis, preliminary risk analysis, failure mode and failure effects analyses; 2 - quantitative study of systems operation safety: binary representation models, state space-based methods, event space-based methods; 3 - application to the design of control systems: safe specifications of a control system, qualitative analysis of operation safety, quantitative analysis, example of application; 4 - conclusion. (J.S.)

  17. Continuous restraint control systems: safety improvement for various occupants

    NARCIS (Netherlands)

    Laan, E. van der; Jager, B. de; Veldpaus, F.; Steinbuch, M.; Nunen, E. van; Willemsen, D.

    2009-01-01

    Occupant safety can be significantly improved by continuous restraint control systems. These restraint systems adjust their configuration during the impact according to the actual operating conditions, such as occupant size, weight, occupant position, belt usage and crash severity. In this study,

  18. Safety Metrics for Human-Computer Controlled Systems

    Science.gov (United States)

    Leveson, Nancy G; Hatanaka, Iwao

    2000-01-01

    The rapid growth of computer technology and innovation has played a significant role in the rise of computer automation of human tasks in modem production systems across all industries. Although the rationale for automation has been to eliminate "human error" or to relieve humans from manual repetitive tasks, various computer-related hazards and accidents have emerged as a direct result of increased system complexity attributed to computer automation. The risk assessment techniques utilized for electromechanical systems are not suitable for today's software-intensive systems or complex human-computer controlled systems.This thesis will propose a new systemic model-based framework for analyzing risk in safety-critical systems where both computers and humans are controlling safety-critical functions. A new systems accident model will be developed based upon modem systems theory and human cognitive processes to better characterize system accidents, the role of human operators, and the influence of software in its direct control of significant system functions Better risk assessments will then be achievable through the application of this new framework to complex human-computer controlled systems.

  19. New technique for determining unavailability of computer controlled safety systems

    International Nuclear Information System (INIS)

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

    1984-04-01

    The availability of a safety system for a fusion reactor is determined. A fusion reactor processes tritium and requires an Emergency Tritium Cleanup (ETC) system for accidental tritium releases. The ETC is computer controlled and because of its complexity, is an excellent candidate for this analysis. The ETC system unavailability, for preliminary untested software, is calculated based on different assumptions about operator response. These assumptions are: (a) the operator shuts down the system after the first indication of plant failure; (b) the operator shuts down the system after following optimized failure verification procedures; or (c) the operator is taken out of the decision process, and the computer uses the optimized failure verification procedures

  20. Spallation Neutron Source Accelerator Facility Target Safety and Non-safety Control Systems

    International Nuclear Information System (INIS)

    Battle, Ronald E.; DeVan, B.; Munro, John K. Jr.

    2006-01-01

    The Spallation Neutron Source (SNS) is a proton accelerator facility that generates neutrons for scientific researchers by spallation of neutrons from a mercury target. The SNS became operational on April 28, 2006, with first beam on target at approximately 200 W. The SNS accelerator, target, and conventional facilities controls are integrated by standardized hardware and software throughout the facility and were designed and fabricated to SNS conventions to ensure compatibility of systems with Experimental Physics Integrated Control System (EPICS). ControlLogix Programmable Logic Controllers (PLCs) interface to instruments and actuators, and EPICS performs the high-level integration of the PLCs such that all operator control can be accomplished from the Central Control room using EPICS graphical screens that pass process variables to and from the PLCs. Three active safety systems were designed to industry standards ISA S84.01 and IEEE 603 to meet the desired reliability for these safety systems. The safety systems protect facility workers and the environment from mercury vapor, mercury radiation, and proton beam radiation. The facility operators operated many of the systems prior to beam on target and developed the operating procedures. The safety and non-safety control systems were tested extensively prior to beam on target. This testing was crucial to identify wiring and software errors and failed components, the result of which was few problems during operation with beam on target. The SNS has continued beam on target since April to increase beam power, check out the scientific instruments, and continue testing the operation of facility subsystems

  1. Developing and maintaining national food safety control systems ...

    African Journals Online (AJOL)

    The establishment of effective food safety systems is pivotal to ensuring the safety of the national food supply as well as food products for regional and international trade. The development, structure and implementation of modern food safety systems have been driven over the years by a number of developments.

  2. Safety problems in vehicles with adaptive cruise control system

    Directory of Open Access Journals (Sweden)

    Yadav Arun K.

    2017-06-01

    Full Text Available In today’s world automotive industries are still putting efforts towards more autonomous vehicles (AVs. The main concern of introducing the autonomous technology is safety of driver. According to a survey 90% of accidents happen due to mistake of driver. The adaptive cruise control system (ACC is a system which combines cruise control with a collision avoidance system. The ACC system is based on laser and radar technologies. This system is capable of controlling the velocity of vehicle automatically to match the velocity of car, bus or truck in front of vehicle. If the lead vehicle gets slow down or accelerate, than ACC system automatically matches that velocity. The proposed paper is focusing on more accurate methods of detecting the preceding vehicle by using a radar and lidar sensors by considering the vehicle side slip and by controlling the distance between two vehicles. By using this approach i.e. logic for calculation of former vehicle distance and controlling the throttle valve of ACC equipped vehicle, an improvement in driving stability was achieved. The own contribution results with fuel efficient driving and with more safer and reliable driving system, but still some improvements are going on to make it more safe and reliable.

  3. Segmentation Scheme for Safety Enhancement of Engineered Safety Features Component Control System

    International Nuclear Information System (INIS)

    Lee, Sangseok; Sohn, Kwangyoung; Lee, Junku; Park, Geunok

    2013-01-01

    Common Caused Failure (CCF) or undetectable failure would adversely impact safety functions of ESF-CCS in the existing nuclear power plants. We propose the segmentation scheme to solve these problems. Main function assignment to segments in the proposed segmentation scheme is based on functional dependency and critical function success path by using the dependency depth matrix. The segment has functional independence and physical isolation. The segmentation structure is that prohibit failure propagation to others from undetectable failures. Therefore, the segmentation system structure has robustness to undetectable failures. The segmentation system structure has functional diversity. The specific function in the segment defected by CCF, the specific function could be maintained by diverse control function that assigned to other segments. Device level control signals and system level control signals are separated and also control signal and status signals are separated due to signal transmission paths are allocated independently based on signal type. In this kind of design, single device failure or failures on signal path in the channel couldn't result in the loss of all segmented functions simultaneously. Thus the proposed segmentation function is the design scheme that improves availability of safety functions. In conventional ESF-CCS, the single controller generates the signal to control the multiple safety functions, and the reliability is achieved by multiplication within the channel. This design has a drawback causing the loss of multiple functions due to the CCF (Common Cause Failure) and single failure Heterogeneous controller guarantees the diversity ensuring the execution of safety functions against the CCF and single failure, but requiring a lot of resources like manpower and cost. The segmentation technology based on the compartmentalization and functional diversification decreases the CCF and single failure nonetheless the identical types of controllers

  4. Segmentation Scheme for Safety Enhancement of Engineered Safety Features Component Control System

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Sangseok; Sohn, Kwangyoung [Korea Reliability Technology and System, Daejeon (Korea, Republic of); Lee, Junku; Park, Geunok [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-05-15

    Common Caused Failure (CCF) or undetectable failure would adversely impact safety functions of ESF-CCS in the existing nuclear power plants. We propose the segmentation scheme to solve these problems. Main function assignment to segments in the proposed segmentation scheme is based on functional dependency and critical function success path by using the dependency depth matrix. The segment has functional independence and physical isolation. The segmentation structure is that prohibit failure propagation to others from undetectable failures. Therefore, the segmentation system structure has robustness to undetectable failures. The segmentation system structure has functional diversity. The specific function in the segment defected by CCF, the specific function could be maintained by diverse control function that assigned to other segments. Device level control signals and system level control signals are separated and also control signal and status signals are separated due to signal transmission paths are allocated independently based on signal type. In this kind of design, single device failure or failures on signal path in the channel couldn't result in the loss of all segmented functions simultaneously. Thus the proposed segmentation function is the design scheme that improves availability of safety functions. In conventional ESF-CCS, the single controller generates the signal to control the multiple safety functions, and the reliability is achieved by multiplication within the channel. This design has a drawback causing the loss of multiple functions due to the CCF (Common Cause Failure) and single failure Heterogeneous controller guarantees the diversity ensuring the execution of safety functions against the CCF and single failure, but requiring a lot of resources like manpower and cost. The segmentation technology based on the compartmentalization and functional diversification decreases the CCF and single failure nonetheless the identical types of

  5. Safety implications of using programmable digital computers in nuclear safety and control systems

    International Nuclear Information System (INIS)

    Adams, D.M.; Rohrdanz, R.R.

    1982-01-01

    This papers describes the activities being conducted at the Idaho National Engineering Laboratory associated with the use of stored-program computers for protection and control systems. This project has recently been initiated and a preliminary report will be available. The use of computers in plant control and protection (and more generally in system important to safety) represents a major departure from the systems which have been used in the past. The design, development, and audit methods used for these systems are significantly different, thus requiring different skills and different perspectives

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

  7. Upgrading instrumentation and control systems for plant safety and operation

    International Nuclear Information System (INIS)

    Martin, M.; Prehler, H.J.; Schramm, W.

    1997-01-01

    Upgrading the electrical systems and instrumentation and control systems has become increasingly more important in the past few years for nuclear power plants currently in operation. As the requirements to be met in terms of plant safety and availability have become more stringent in the past few years, Western plants built in the sixties and seventies have been the subject of manifold backfitting and upgrading measures in the past. In the meantime, however, various nuclear power plants are facing much more thorough upgrading phases because of the difficulties in obtaining spare parts for older equipment systems. As digital technology has become widespread in many areas because of its advantages, and as applications are continuously expanding, conventional equipment and systems are losing more and more ground as a consequence of decreasing demand. Merely because of the pronounced decline in demand for conventional electronic components it is possible for equipment manufacturers to guarantee spare parts deliveries for older systems only for specific future periods of time. In addition, one-off manufacture entails high costs in purchases of spare parts. As a consequence of current thinking more and more focusing on availability and economy, upgrading of electrical systems and instrumentation and control systems is becoming a more and more topical question, for older plants even to ensure completion of full service life. (orig.) [de

  8. Programmable logic controller (PLC) for safety systems of nuclear plants

    International Nuclear Information System (INIS)

    Sen, S.K.; Karmakar, G.; Joseph, Jose; Patil, R.K.

    2002-01-01

    Full text: A programmable logic controller (PLC) has been developed by RCnD, BARC for use in the safety critical systems in nuclear power plants. This PLC uses qualified hardware developed in RCnD for use in NPP. The programming software conforms to IEC-61131 part 3. The application programming is done on function block diagram (FBD) editor and the FBD is automatically converted into code in high level language (C / C++). This feature makes the application easily decipherable and therefore easily subjected to reviews and other validation techniques. The key to make quality software for use in nuclear systems is to enforce various standards in the design and development of the software, something, which is not possible to do with a commercially available PLC. This PLC with its software completely transparent lends itself to rigorous verification and validation easily

  9. 30 CFR 7.103 - Safety system control test.

    Science.gov (United States)

    2010-07-01

    ... Areas of Underground Coal Mines Where Permissible Electric Equipment is Required § 7.103 Safety system... operate immediately when activated and stop the engine within 15 seconds. (6) The total intake air inlet...

  10. The safety implications of control systems program at ORNL

    International Nuclear Information System (INIS)

    Smith, O.L.

    1987-01-01

    Simulations of two pressurized water reactors (PWRs) point to several conclusions that bear on the principle interests of Unresolved Safety Issue A-47: (1) the simulated control systems of both plants exhibit considerable ability to respond to the investigated classes of off-normal disturbances; (2) overfill of the steam generators usually produced only minor cooling of the primary side; (3) despite protective features, substantial amounts of water could be injected into the steam lines because of low steam quality or high water level, but further analysis is needed to determine whether this creates the potential for water-hammer damage or other mass or momentum effects; and (4) potential core-uncovery scenarios explored steam generator tube rupture and other small breaks that might lead to loss of primary inventory without actuation of high pressure injection. The results indicated situations in which automatic actuation of high pressure injection would terminate the leak and others in which operator intervention appeared necessary

  11. Reliability Analysis Multiple Redundancy Controller for Nuclear Safety Systems

    International Nuclear Information System (INIS)

    Son, Gwangseop; Kim, Donghoon; Son, Choulwoong

    2013-01-01

    This controller is configured for multiple modular redundancy (MMR) composed of dual modular redundancy (DMR) and triple modular redundancy (TMR). The architecture of MRC is briefly described, and the Markov model is developed. Based on the model, the reliability and Mean Time To Failure (MTTF) are analyzed. In this paper, the architecture of MRC for nuclear safety systems is described. The MRC is configured for multiple modular redundancy (MMR) composed of dual modular redundancy (DMR) and triple modular redundancy (TMR). Markov models for MRC architecture was developed, and then the reliability was analyzed by using the model. From the reliability analyses for the MRC, it is obtained that the failure rate of each module in the MRC should be less than 2 Χ 10 -4 /hour and the MTTF average increase rate depending on FCF increment, i. e. ΔMTTF/ΔFCF, is 4 months/0.1

  12. The safety implications of control systems program at ORNL

    International Nuclear Information System (INIS)

    Smith, O.L.

    1987-01-01

    Simulations of two pressurized water reactors (PWRs) point to several conclusions that bear on the principle interests of Unresolved Safety Issue A-47: the simulated control systems of both plants exhibit considerable ability to respond to the investigated classes of off-normal disturbances; overfill of the steam generators usually produced only minor cooling of the primary side; despite protective features, substantial amounts of water could be injected into the steam lines because of low steam quality or high water level, but further analysis is needed to determine whether this creates the potential for water-hammer damage or other mass or momentum effects; and potential core-uncovery scenarios explored steam generator tube rupture and other small breaks that might lead to loss of primary inventory without actuation of high pressure injection. The results indicated situations in which automatic actuation of high pressure injection would terminate the leak and others in which operator intervention appeared necessary

  13. Safety and control of accelerator-driven subcritical systems

    Energy Technology Data Exchange (ETDEWEB)

    Rief, H. [Ispra Establishment (Italy); Takahashi, H. [Brookhaven National Laboratory, Long Island, NY (United States)

    1995-10-01

    To study control and safety of accelertor driven nuclear systems, a one point kinetic model was developed and programed. It deals with fast transients as a function of reactivity insertion. Doppler feedback, and the intensity of an external neutron source. The model allows for a simultaneous calculation of an equivalent critical reactor. It was validated by a comparison with a benchmark specified by the Nuclear Energy Agency Committee of Reactor Physics. Additional features are the possibility of inserting a linear or quadratic time dependent reactivity ramp which may account for gravity induced accidents like earthquakes, the possibility to shut down the external neutron source by an exponential decay law of the form exp({minus}t/{tau}), and a graphical display of the power and reactivity changes. The calculations revealed that such boosters behave quite benignly even if they are only slightly subcritical.

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

    International Nuclear Information System (INIS)

    Chi Zongbo; Wu Yuanqiang

    1997-01-01

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

  15. Nuclear safety considerations with emphasis on instrumentation and control systems

    International Nuclear Information System (INIS)

    Beare, J.W.

    1978-01-01

    The conceptual model of a nuclear power plant in Canada is that it consists basically of two kinds of systems. The first kind is the process systems, that is, those structures and components associated with the production of nuclear energy and its conversion to other forms of energy. The second kind is the special safety systems, whose purpose it is to protect the public in the event of a serious failure in the process systems which might otherwise lead to unacceptable radiological consequences. Quantitative limits are set on the unavailability of the special safety systems. These limits are low enough to be consistent with low overall risk and yet can be demonstrated by test during operation of the plant. Low unavailability is an important but not the only condition required for low unrealiability for the special safety systems. The special safety systems minimize the chance of a cross-linked failure particularly under the conditions experienced as a result of the more severe types of postulated serious process failures. Nuclear power plants must also withstand, without a major hazard to the public, certain rare events associated with natural phenomena or man-made activities off-site and also certain in-plant events such as fire or break-up of a turbine-generator which might have a cross-linking effect on process and safety systems. In the latest designs, Canadian nuclear power plants have emergency systems to deal with such events. The emergency systems have an enhanced degree of physical and functional separation from other plant systems. (author)

  16. Risk assessment of safety data link and network communication in digital safety feature control system of nuclear power plant

    International Nuclear Information System (INIS)

    Lee, Sang Hun; Son, Kwang Seop; Jung, Wondea; Kang, Hyun Gook

    2017-01-01

    Highlights: • Safety data communication risk assessment framework and quantitative scheme were proposed. • Fault-tree model of ESFAS unavailability due to safety data communication failure was developed. • Safety data link and network risk were assessed based on various ESF-CCS design specifications. • The effect of fault-tolerant algorithm reliability of safety data network on ESFAS unavailability was assessed. - Abstract: As one of the safety-critical systems in nuclear power plants (NPPs), the Engineered Safety Feature-Component Control System (ESF-CCS) employs safety data link and network communication for the transmission of safety component actuation signals from the group controllers to loop controllers to effectively accommodate various safety-critical field controllers. Since data communication failure risk in the ESF-CCS has yet to be fully quantified, the ESF-CCS employing data communication systems have not been applied in NPPs. This study therefore developed a fault tree model to assess the data link and data network failure-induced unavailability of a system function used to generate an automated control signal for accident mitigation equipment. The current aim is to provide risk information regarding data communication failure in a digital safety feature control system in consideration of interconnection between controllers and the fault-tolerant algorithm implemented in the target system. Based on the developed fault tree model, case studies were performed to quantitatively assess the unavailability of ESF-CCS signal generation due to data link and network failure and its risk effect on safety signal generation failure. This study is expected to provide insight into the risk assessment of safety-critical data communication in a digitalized NPP instrumentation and control system.

  17. System and software safety analysis for the ERA control computer

    International Nuclear Information System (INIS)

    Beerthuizen, P.G.; Kruidhof, W.

    2001-01-01

    The European Robotic Arm (ERA) is a seven degrees of freedom relocatable anthropomorphic robotic manipulator system, to be used in manned space operation on the International Space Station, supporting the assembly and external servicing of the Russian segment. The safety design concept and implementation of the ERA is described, in particular with respect to the central computer's software design. A top-down analysis and specification process is used to down flow the safety aspects of the ERA system towards the subsystems, which are produced by a consortium of companies in many countries. The user requirements documents and the critical function list are the key documents in this process. Bottom-up analysis (FMECA) and test, on both subsystem and system level, are the basis for safety verification. A number of examples show the use of the approach and methods used

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

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  19. Availability analysis of safety grade multiple redundant controller used in advanced nuclear safety systems

    International Nuclear Information System (INIS)

    Son, Kwang Seop; Kim, Dong Hoon; Park, Gee Yong; Kang, Hyun Gook

    2018-01-01

    Highlights: •The multiple redundant controller, SPLC is configured as the combination of DMR and TMR architecture. •We construct the Markov model of SPLC using the concept of the system unavailability rate. •To satisfy the availability requirement of safety grade controller, the fault coverage factor (FCF) should be ≥0.8 and the MTTR of each module should be ≤100 h when FCF is 0.9. •The availability of SPLC is better than that of PLC having iTMR architecture however it is poorer than iTMR considering the off-line test and inspection on the assumption that MTTR of each module is ≤200 h. -- Abstract: We analyze the availability of the Safety Programmable Logic Controller (SPLC) having multiple redundant architectures. In the SPLC, input/output and processor module are configured as triple modular redundancy (TMR), and backplane bus, power and communication modules are configured as dual modular redundancy (DMR). The voting logics for redundant architectures are based on the forwarding error detection. It means that the receivers perform the voting logics based on the status information of transmitters. To analyze the availability of SPLC, we construct the Markov model and simplify the model adopting the system unavailability rate. The results show that the fault coverage factor should be ≥0.8 and Mean Time To Repair (MTTR) should be ≤100 h in order to satisfy the requirement that the availability of the safety grade PLC should be ≥0.995. Also we evaluate the availability of SPLC comparing to other PLCs such as simplex, processor DMR (pDMR) and independent TMR (iTMR) PLCs used in the existing nuclear safety systems. The availability of SPLC is higher than those of the simplex, pDMR but is lower than that of iTMR for one month which is the periodic off-line test and inspection. That’s why the number of redundant modules used in PLC is more dominant to increasing the availability than the number of fault masking methods such as voting logics used

  20. 78 FR 979 - Petition for Positive Train Control Safety Plan Approval and System Certification of the...

    Science.gov (United States)

    2013-01-07

    ...] Petition for Positive Train Control Safety Plan Approval and System Certification of the Electronic Train... the Federal Railroad Administration (FRA) for Positive Train Control (PTC) Safety Plan (PTCSP) approval and system certification of the Electronic Train Management System (ETMS) as required by 49 U.S.C...

  1. Analysis approach for common cause failure on non-safety digital control system

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yun Goo; Oh, Eungse [Korea Hydro and Nuclear Power Co. Ltd., Daejeon (Korea, Republic of)

    2014-05-15

    The effects of common cause failure (CCF) on safety digital instrumentation and control (I and C) system had been considered in defense in depth and diversity coping analysis with safety analysis method. For the non-safety system, single failure had been considered for safety analysis. IEEE Std. 603-1991, Clause 5.6.3.1(2), 'Isolation' states that no credible failure on the non-safety side of an isolation device shall prevent any portion of a safety system from meeting its minimum performance requirements during and following any design basis event requiring that safety function. The software CCF is one of the credible failure on the non-safety side. In advanced digital I and C system, same hardware component is used for different control system and the defect in manufacture or common external event can generate CCF. Moreover, the non-safety I and C system uses complex software for its various function and software quality assurance for the development process is less severe than safety software for the cost effective design. Therefore the potential defects in software cannot be ignored and the effect of software CCF on non-safety I and C system is needed to be evaluated. This paper proposes the general process and considerations for the analysis of CCF on non-safety I and C system.

  2. Retrofit of safety and control systems in nuclear power plants

    International Nuclear Information System (INIS)

    Keiper, J.T.; Fassett, G.B.

    1986-01-01

    The modularity, compactness, compatibility, and licensability of the microcontrol system make it a cost-effective approach to obtain the benefits of digital control technology in the retrofit of nuclear power plants. Retrofit of individual loops or complete systems can be scheduled to meet the operational needs of the plant. The existing racks, panels, and cable systems can be utilized to the maximum extent to minimize the installed cost. Future expansion to total plant control or plant management is supported by the network communication module or gateway. The microcontrol module provides benefits now in improved operation, and future benefits in planned, controlled upgrading

  3. Research on conceptual design of simplified nuclear safety instrument and control system

    International Nuclear Information System (INIS)

    Huang Jie

    2015-01-01

    The Nuclear safety instrument and control system is directly related to the safety of the reactor. So redundant and diversity design is used to ensure the system's security and reliability. This make the traditional safety system large, more cabinets and wiring complexity. To solve these problem, we can adopt new technology to make the design more simple. The simplify conceptual design can make the system less cabinets, less wiring, but high security, strong reliability. (author)

  4. A survey of approaches combining safety and security for industrial control systems

    International Nuclear Information System (INIS)

    Kriaa, Siwar; Pietre-Cambacedes, Ludovic; Bouissou, Marc; Halgand, Yoran

    2015-01-01

    The migration towards digital control systems creates new security threats that can endanger the safety of industrial infrastructures. Addressing the convergence of safety and security concerns in this context, we provide a comprehensive survey of existing approaches to industrial facility design and risk assessment that consider both safety and security. We also provide a comparative analysis of the different approaches identified in the literature. - Highlights: • We raise awareness of safety and security convergence in numerical control systems. • We highlight safety and security interdependencies for modern industrial systems. • We give a survey of approaches combining safety and security engineering. • We discuss the potential of the approaches to model safety and security interactions

  5. Cold Vacuum Drying Safety Class Instrumentation and Control System Design Description SYS 93-2

    International Nuclear Information System (INIS)

    WHITEHURST, R.

    1999-01-01

    This document describes the Cold Vacuum Drying Facility (CVDF) Safety Class Instrumentation and Control system (SCIC). The SCIC provides safety functions and features to protect the environment, off-site and on-site personnel and equipment. The function of the SCIC is to provide automatic trip features, valve interlocks, alarms, indication and control for the cold vacuum drying process

  6. Cold Vacuum Drying Safety Class Instrumentation and Control System Design Description

    International Nuclear Information System (INIS)

    WHITEHURST, R.

    1999-01-01

    This document describes the Cold Vacuum Drying Facility (CVDF) Safety Class Instrumentation and Control system (SCIC). The SCIC provides safety functions and features to protect the environment, off-site and on-site personnel and equipment. The function of the SCIC is to provide automatic trip features, valve interlocks, alarms, indication and control for the cold vacuum drying process

  7. Safety System for Controlling Fluid Flow into a Suction Line

    Science.gov (United States)

    England, John Dwight (Inventor); Kelley, Anthony R. (Inventor); Cronise, Raymond J. (Inventor)

    2018-01-01

    A safety system includes a sleeve fitted within a pool's suction line at its inlet. The sleeve terminates with a plate that resides within the suction line. The plate has holes formed therethrough. A housing defining distinct channels is fitted in the sleeve so that the distinct channels lie within the sleeve. Each of the distinct channels has a first opening on one end thereof and a second opening on another end thereof. The second openings reside in the sleeve. The first openings are in fluid communication with the water in the pool, and are distributed around a periphery of an area of the housing that prevents coverage of all the first openings when a human interacts therewith. A first sensor is coupled to the sleeve to sense pressure therein, and a second pressure sensor is coupled to the plate to sense pressure in one of the plates' holes.

  8. 77 FR 22637 - Federal Motor Vehicle Safety Standards; Accelerator Control Systems

    Science.gov (United States)

    2012-04-16

    ... revise the Federal Motor Vehicle Safety Standard for accelerator control systems (ACS) in two ways. First... Standard (FMVSS) No. 124, Accelerator Control Systems,\\2\\ in two ways. First, we are proposing to update... February 2011 final report ``Technical Assessment of Toyota Electronic Throttle Control Systems,'' the...

  9. Control system of labour safety measures in the higher educational institution

    Directory of Open Access Journals (Sweden)

    O. G. Feoktistova

    2015-01-01

    Full Text Available The article examines a system of labour safety measures control. With the introduction of the integrated system of management the competitive ability of production and organization, the effectiveness of its activity rise, and sinnergicheskiy effect is also reached and the savings of all forms of resources are ensured. Objectives and methods of control system of labour safety measures in enterprises are developed, including in the educational institutions.

  10. JACoW Safety instrumented systems and the AWAKE plasma control as a use case

    CERN Document Server

    Blanco Viñuela, Enrique; Fernández Adiego, Borja; Speroni, Roberto

    2018-01-01

    Safety is likely the most critical concern in many process industries, yet there is a general uncertainty on the proper engineering to reduce the risks and ensure the safety of persons or material at the same time as providing the process control system. Some of the reasons for this misperception are unclear requirements, lack of functional safety engineering knowledge or incorrect protection functionalities attributed to the BPCS (Basic Process Control System). Occasionally the control engineers are not aware of the hazards inherent to an industrial process and this causes an incorrect design of the overall controls. This paper illustrates the engineering of the SIS (Safety Instrumented System) and the BPCS of the plasma vapour controls of the AWAKE R&D; project, the first proton-driven plasma wakefield acceleration experiment in the world. The controls design and implementation refers to the IEC61511/ISA84 standard, including technological choices, design, operation and maintenance. Finally, the publica...

  11. Performance of SPNDs used in control and safety systems

    International Nuclear Information System (INIS)

    Fernando, M.P.S.; Raj, Manish; Kumar, A.N.

    2006-01-01

    Large sized reactor such as 540 MWe Pressurised Heavy Water Reactor (PHWR) requires continuous in core monitoring of local flux in order to provide effective control and protection. About 198 self powered neutron detectors (SPNDs) of the straight individually replaceable type are distributed in the reactor core. For purposes of reactor regulation, 42 prompt responding cobalt SPNDs called zone control detectors (ZCDs) are housed in vertical flux units (VFUs) and these are uniformly distributed in 14 power zones. The in core detectors used for spatial control by ZCCs do not accurately represent average zone power as they sense the flux over a small volume. Flux mapping system (FMS) comprising of 102 vanadium SPNDs in 26 VFUs, provide accurate measure of neutron flux, even though they have slow response to change in neutron flux levels. For reactor protection system-1 (RPS-1), 36 cobalt SPNDs are placed in VFUs and become part of core overpower protection system-1 (COPPS-1). Similarly, for RPS-2, 18 cobalt SPNDs are placed in horizontal flux units (HFUs) and become part of the COPPS-2. The present study discusses the performance of in core SPNDs used in TAPP-4 by comparing the measured fluxes with detailed simulations. The performances of SPNDs are evaluated at different power levels and several full power day of reactor operation. (author)

  12. Impacts of safety on the design of light remotely-piloted helicopter flight control systems

    International Nuclear Information System (INIS)

    Di Rito, G.; Schettini, F.

    2016-01-01

    This paper deals with the architecture definition and the safety assessment of flight control systems for light remotely-piloted helicopters for civil applications. The methods and tools to be used for these activities are standardised for conventional piloted aircraft, while they are currently a matter of discussion in case of light remotely-piloted systems flying into unsegregated airspaces. Certification concerns are particularly problematic for aerial systems weighing from 20 to 150 kgf, since the airworthiness permission is granted by national authorities. The lack of specific requirements actually requires to analyse both the existing standards for military applications and the certification guidelines for civil systems, up to derive the adequate safety objectives. In this work, after a survey on applicable certification documents for the safety objectives definition, the most relevant functional failures of a light remotely-piloted helicopter are identified and analysed via Functional Hazard Assessment. Different architectures are then compared by means of Fault-Tree Analysis, highlighting the contributions to the safety level of the main elements of the flight control system (control computers, servoactuators, antenna) and providing basic guidelines on the required redundancy level. - Highlights: • A method for architecture definition and safety assessment of light RW‐UAS flight control systems is proposed. • Relevant UAS failures are identified and analysed via Functional Hazard Assessment and Fault‐Tree Analysis. • The key safety elements are control computers, servoactuators and TX/RX system. • Single‐simplex flight control systems have inadequate safety levels. • Dual‐duplex flight control systems demonstrate to be safety compliant, with safety budgets dominated by servoactuators.

  13. Division of Cyber Safety and Security Responsibilities Between Control System Owners and Suppliers

    OpenAIRE

    Skotnes , Ruth

    2016-01-01

    Part 2: CONTROL SYSTEMS SECURITY; International audience; The chapter discusses the important issue of responsibility for information and communications technology (ICT) – or cyber – safety and security for industrial control systems and the challenges involved in dividing the responsibility between industrial control system owners and suppliers in the Norwegian electric power supply industry. Industrial control system owners are increasingly adopting information and communications technologi...

  14. A Practical Risk Assessment Methodology for Safety-Critical Train Control Systems

    Science.gov (United States)

    2009-07-01

    This project proposes a Practical Risk Assessment Methodology (PRAM) for analyzing railroad accident data and assessing the risk and benefit of safety-critical train control systems. This report documents in simple steps the algorithms and data input...

  15. The engineering project and reliability research of the safety interlock slow control system in BESIII

    International Nuclear Information System (INIS)

    Zhang Yinhong; Zhao Jingwei; Li Xiaonan; Xie Xiaoxi; Gao Cuishan; Bai Jingzhi; Chen Xihui; Min Jian; Nie Zhendong

    2008-01-01

    The new safety interlock slow control system of BESIII is designed to ensure that the BESIII interior equipments and the accelerator control center to work in coordination, and to guarantee the safety of the operating staff and all the important equipments at the same time. This paper introduces the hardware and software design of safety interlock system from the engineering requirements angle, including a detailed research on the software implementation technique of the state machine on PLC and the reliability of the system. (authors)

  16. Design measures to increase safety and reliability of power station control and protection systems

    International Nuclear Information System (INIS)

    Edelmann, J.; Spieth, W.

    1977-06-01

    The paper reviews a few criteria which exert a considerable influence on the safety and reliability of monitoring and control systems. When judging the safety and reliability of a system, it is of importance not only to look at the failures of just one part of a system but also to take into account the effect these failures have on the overall process. In this respect there is a marked difference between a centralized and a decentralized system. With the technical equipment nowadays at our disposal a high safety standard has been reached. Redundant and dynamic protection systems make the occurrence of a dangerous failure hypothetic. (Author)

  17. Design a Smart Control Strategy to Implement an Intelligent Energy Safety and Management System

    OpenAIRE

    Jing-Min Wang; Ming-Ta Yang

    2014-01-01

    The energy saving and electricity safety are today a cause for increasing concern for homes and buildings. Integrating the radio frequency identification (RFID) and ZigBee wireless sensor network (WSN) mature technologies, the paper designs a smart control strategy to implement an intelligent energy safety and management system (IESMS) which performs energy measuring, controlling, monitoring, and saving of the power outlet system. The presented RFID and billing module is used to identify user...

  18. Towards a decision support system for control of multiple food safety hazards in raw milk production

    NARCIS (Netherlands)

    Spiegel, van der M.; Sterrenburg, P.; Haasnoot, W.; Fels-Klerx, van der H.J.

    2013-01-01

    Decision support systems (DSS) for controlling multiple food safety hazards in raw milk production have not yet been developed, but the underlying components are fragmentarily available. This article presents the state-of-the-art of essential DSS elements for judging food safety compliance of raw

  19. On safety classification of instrumentation and control systems and their components

    International Nuclear Information System (INIS)

    Yastrebenetskij, M.A.; Rozen, Yu.V.

    2004-01-01

    Safety classification of instrumentation and control systems (I and C) and their components (hardware, software, software-hardware complexes) is described: - evaluation of classification principles and criteria in Ukrainian standards and rules; comparison between Ukrainian and international principles and criteria; possibility and ways of coordination of Ukrainian and international standards related to (I and C) safety classification

  20. Verification of the safety communication protocol in train control system using colored Petri net

    International Nuclear Information System (INIS)

    Chen Lijie; Tang Tao; Zhao Xianqiong; Schnieder, Eckehard

    2012-01-01

    This paper deals with formal and simulation-based verification of the safety communication protocol in ETCS (European Train Control System). The safety communication protocol controls the establishment of safety connection between train and trackside. Because of its graphical user interface and modeling flexibility upon the changes in the system conditions, this paper proposes a composition Colored Petri Net (CPN) representation for both the logic and the timed model. The logic of the protocol is proved to be safe by means of state space analysis: the dead markings are correct; there are no dead transitions; being fair. Further analysis results have been obtained using formal and simulation-based verification approach. The timed models for the open transmit system and the application process are created for the purpose of performance analysis of the safety communication protocol. The models describe the procedure of data transmission and processing, and also provide relevant timed and stochastic factors, as well as time delay and lost packet, which may influence the time for establishment of safety connection of the protocol. Time for establishment of safety connection of the protocol in normal state is verified by formal verification, and then time for establishment of safety connection with different probability of lost packet is simulated. After verification it is found that the time for establishment of safety connection of the safety communication protocol satisfies the safety requirements.

  1. Analysis Method of Common Cause Failure on Non-safety Digital Control System

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yun Goo; Oh, Eun Gse [KHNP, Daejeon (Korea, Republic of)

    2014-08-15

    The effects of common cause failure on safety digital instrumentation and control system had been considered in defense in depth analysis with safety analysis method. However, the effects of common cause failure on non-safety digital instrumentation and control system also should be evaluated. The common cause failure can be included in credible failure on the non-safety system. In the I and C architecture of nuclear power plant, many design feature has been applied for the functional integrity of control system. One of that is segmentation. Segmentation defenses the propagation of faults in the I and C architecture. Some of effects from common cause failure also can be limited by segmentation. Therefore, in this paper there are two type of failure mode, one is failures in one control group which is segmented, and the other is failures in multiple control group because that the segmentation cannot defense all effects from common cause failure. For each type, the worst failure scenario is needed to be determined, so the analysis method has been proposed in this paper. The evaluation can be qualitative when there is sufficient justification that the effects are bounded in previous safety analysis. When it is not bounded in previous safety analysis, additional analysis should be done with conservative assumptions method of previous safety analysis or best estimation method with realistic assumptions.

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  3. New Methodology for a Comprehensive Modular Safety Control System in a Cyclotron Site

    International Nuclear Information System (INIS)

    Kaufman, Y.; Kravitz, M.; Arad, M.; Osovizky, A.; Paran, J.; Sarussi, B.; Ellenbogen, M.; Tal, N.

    2004-01-01

    This Paper describes a new methodology for a comprehensive modular Safety Control System (SCS), for a cyclotron site. The developed SCS is a modular approach for controlling the production procedures, safety conditions and documentation aspects in the Cyclotron site. Usually, the safety conditions in cyclotron sites are maintained by a variety of sensors. The cyclotron is supplied from the manufacturer with a self-integrated control system for its operation, yet the comprehensive SCS has to be defined and setup by the customer. Therefore, customers face a lot of integration problems in trying to combine all the signals from the different safety systems such as radiation monitoring, environmental and access control, in order to maintain proper safety working conditions. The presented SCS design provides main user interface and the complete safety solution required by including preset control logic definitions and open logic for specific user applications. The knowledge for the preset control logic definitions was gathered in previous projects. Failure Mode and Effects Analysis (FMEA) method has been implemented on the SCS to analyze the potential failure modes and their impact on the product reliability

  4. New design of engineered safety features-component control system to improve performance and reliability

    International Nuclear Information System (INIS)

    Kim, S.T.; Jung, H.W.; Lee, S.J.; Cho, C.H.; Kim, D.H.; Kim, H.

    2006-01-01

    Full text: Full text: The Engineered Safety Features-Component Control System (ESF-CCS) controls the engineered safety features of a Nuclear Power Plant such as Solenoid Operated Valves (SOV), Motor Operated Valves (MOV), pumps, dampers, etc. to mitigate the effects of a Design Basis Accident (DBA) or an abnormal operation. ESF-CCS serves as an interface system between the Plant Protection System (PPS) and remote actuation devices. ESF-CCS is composed of fault tolerant Group Controllers GC, Loop Controllers (LC), ESF-CCS Test and Interface Processor (ETIP) and Cabinet Operator Module (COM) and Control Channel Gateway (CCG) etc. GCs in each division are designed to be fully independent triple configuration, which perform system level NSSS and BOP ESFAS logic (2-out-of-4 logic and l-out-of-2 logic, respectively) making it possible to test each GC individually during normal operation. In the existing configuration, the safety-related plant component control is part of the Plant Control System (PCS) non-safety system. For increased safety and reliability, this design change incorporates this part into the LCs, and is therefore designed according to the safety-critical system procedures. The test and diagnosis capabilities of ETIP and COM are reinforced. By means of an automatic periodic test for all main functions of the system, it is possible to quickly determine an abnormal status of the system, and to decrease the elapsed time for tests, thus effectively increasing availability. ESF-CCS consists of four independent divisions (A, B, C, and D) in the Advanced Power Reactor 1400 (APR1400). One prototype division is being manufactured and will be tested

  5. Who is in control of road safety? A STAMP control structure analysis of the road transport system in Queensland, Australia.

    Science.gov (United States)

    Salmon, Paul M; Read, Gemma J M; Stevens, Nicholas J

    2016-11-01

    Despite significant progress, road trauma continues to represent a global safety issue. In Queensland (Qld), Australia, there is currently a focus on preventing the 'fatal five' behaviours underpinning road trauma (drug and drink driving, distraction, seat belt wearing, speeding, and fatigue), along with an emphasis on a shared responsibility for road safety that spans road users, vehicle manufacturers, designers, policy makers etc. The aim of this article is to clarify who shares the responsibility for road safety in Qld and to determine what control measures are enacted to prevent the fatal five behaviours. This is achieved through the presentation of a control structure model that depicts the actors and organisations within the Qld road transport system along with the control and feedback relationships that exist between them. Validated through a Delphi study, the model shows a diverse set of actors and organisations who share the responsibility for road safety that goes beyond those discussed in road safety policies and strategies. The analysis also shows that, compared to other safety critical domains, there are less formal control structures in road transport and that opportunities exist to add new controls and strengthen existing ones. Relationships that influence rather than control are also prominent. Finally, when compared to other safety critical domains, the strength of road safety controls is brought into question. Copyright © 2016 Elsevier Ltd. All rights reserved.

  6. Technical considerations for the development of an engineering safety features control system with PLC

    International Nuclear Information System (INIS)

    Lee, C. K.; Kim, C. H.; Han, J. B.; Kim, H.; Lee, S. S.

    2002-01-01

    Technical considerations are summarized for the development of an ESFCS(Engineered Safety Features Control System) with PLC (Programmable Logic Controller). The ESFCS is required for the mitigation of plant accident conditions and therefore developed in conformance with the design requirements applied to the safety critical system. The design of ESFCS primarily considered its safety, and the system has an architecture that will be able to minimize spurious actuation. The PLC based functional distribution and redundant design features are adopted, and the fieldbus is applied in the communication of information and control signals between PLC processors. It is expected that the ESFCS will have several advanced design features compared with the conventional systems supplied by foreign vendors

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

    International Nuclear Information System (INIS)

    Popovic, B.; Pesic, M.

    1962-01-01

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

  8. Experience in the review of utility control room design review and safety parameter display system programs

    International Nuclear Information System (INIS)

    Moore, V.A.

    1985-01-01

    The Detailed Control Room Design Review (DCRDR) and the Safety Parameter Display System (SPDS) had their origins in the studies and investigations conducted as the result of the TMI-2 accident. The President's Commission (Kemeny Commission) critized NRC for not examining the man-machine interface, over-emphasizing equipment, ignoring human beings, and tolerating outdated technology in control rooms. The Commission's Special Inquiry Group (Rogovin Report) recommended greater application of human factors engineering including better instrumentation displays and improved control room design. The NRC Lessons Learned Task Force concluded that licensees should review and improve control rooms using NRC Human engineering guidelines, and install safety parameter display systems (then called the safety staff vector). The TMI Action Plan Item I.D.1 and I.D.2 were based on these recommendations

  9. Safety of Mixed Model Access Control in a Multilevel System

    Science.gov (United States)

    2014-06-01

    42  H.  FIREWALL AND IPS LANGUAGES...Research Laboratory AIS automated information system ANOA advance notice of arrival APT advanced persistent threat BFM boundary flow modeling...of Investigation FW firewall GENSER general service xvi GUI graphical user interface HAG high-assurance guard HGS high-grade service H-H-H High

  10. Impact Safety Control Strategy for the Battery System of an Example Electric Bus

    Directory of Open Access Journals (Sweden)

    Zhen-po Wang

    2015-01-01

    Full Text Available This paper proposes a side impact safety control strategy for the battery system, aiming at defusing the hazards of unacceptable behaviors of the battery system such as high-voltage hazards. Based on some collision identification metrics, a side impact discrimination algorithm and a side impact severity algorithm are developed for electric buses. Based on the study on the time to break for power battery, the side impact discrimination algorithm response time is about 20 ms posing a great challenge to the side impact discrimination algorithm. At the same time, the reliability of the impact safety control strategy developed in this paper is evaluated for other plausible side impact signals generated by finite element analysis. The results verify that the impact safety control strategy exhibits robust performance and is able to trigger a breaking signal for power battery system promptly and accurately.

  11. Implementation of amplifiers, control and safety subsystems of radiofrequency system of VINCY Cyclotron

    International Nuclear Information System (INIS)

    Drndarevic, V.; Obradovic, M.; Samardic, B.; Djuric, B.; Bojovic, B.; Trajic, M.I.; Golubicic, Z.; Smiljakovic, V.

    1996-01-01

    Concept and design of power amplifiers, control subsystem and safety subsystems for the RF system of the VINCY cyclotron are described. The power amplifiers subsystem consists of two amplifiers of 30 kW nominal power that operate in class B or class C. High stability of voltage amplitude of 5x10 -4 and phase stability between two resonators better than ± 0.5 0 in the range from 16.5 to 31 MHz is being providing by RF control subsystem. Autonomous safety system serves to protect staff from high voltage and to protect equipment from damage. (author)

  12. Radiological safety and control

    International Nuclear Information System (INIS)

    Kim, Jang Hee; Kim, Ki Sub

    1995-01-01

    The practical objective of radiological safety control is intended for achievement and maintenance of appropreately safe condition in environmental control for activities involving exposure from the use of radiation. In order to establish these objectives, we should be to prevent deterministic effects and to limit the occurrence stochastic effects to level deemed to be acceptable by the application of general principles of radiation protection and systems of dose limitation based on ICRP recommendations. 34 tabs., 19 figs., 11 refs. (Author) .new

  13. Radiation safety and control

    International Nuclear Information System (INIS)

    Kim, Jang Hee; Kim, Gi Sub.

    1996-12-01

    The principal objective of radiological safety control is intended for achievement and maintenance of appropriately safe condition in environmental control for activities involving exposure from the use of radiation. In order to establish these objective, we should be to prevent deterministic effects and to limit the occurrence stochastic effects to level deemed to be acceptable by the application of general principles of radiation protection and systems of dose limitation based on ICRP recommendations. (author). 22 tabs., 13 figs., 11 refs

  14. Design of Instrumentation and Control Systems for Nuclear Power Plants. Specific Safety Guide

    International Nuclear Information System (INIS)

    2016-01-01

    This publication is a revision and combination of two Safety Guides, IAEA Safety Standards Series No. NS-G-1.1 and No. NS-G-1.3. The revision takes into account developments in instrumentation and control (I&C) systems since the publication of the earlier Safety Guides. The main changes relate to the continuing development of computer applications and the evolution of the methods necessary for their safe, secure and practical use. In addition, account is taken of developments in human factors engineering and the need for computer security. This Safety Guide references and takes into account other IAEA Safety Standards and Nuclear Security Series publications that provide guidance relating to I&C design

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

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  16. Safety critical FPGA-based NPP instrumentation and control systems: assessment, development and implementation

    Energy Technology Data Exchange (ETDEWEB)

    Bakhmach, E. S.; Siora, A. A.; Tokarev, V. I. [Research and Production Corporation Radiy, 29 Geroev Stalingrada Str., Kirovograd 25006 (Ukraine); Kharchenko, V. S.; Sklyar, V. V.; Andrashov, A. A., E-mail: marketing@radiy.co [Center for Safety Infrastructure-Oriented Research and Analysis, 37 Astronomicheskaya Str., Kharkiv 61085 (Ukraine)

    2010-10-15

    The stages of development, production, verification, licensing and implementation methods and technologies of safety critical instrumentation and control systems for nuclear power plants (NPP) based on FPGA (Field Programmable Gates Arrays) technologies are described. A life cycle model and multi-version technologies of dependability and safety assurance of FPGA-based instrumentation and control systems are discussed. An analysis of NPP instrumentation and control systems construction principles developed by Research and Production Corporation Radiy using FPGA-technologies and results of these systems implementation and operation at Ukrainian and Bulgarian NPP are presented. The RADIY{sup TM} platform has been designed and developed by Research and Production Corporation Radiy, Ukraine. The main peculiarity of the RADIY{sup TM} platform is the use of FPGA as programmable components for logic control operation. The FPGA-based RADIY{sup TM} platform used for NPP instrumentation and control systems development ensures sca lability of system functions types, volume and peculiarities (by changing quantity and quality of sensors, actuators, input/output signals and control algorithms); sca lability of dependability (safety integrity) (by changing a number of redundant channel, tiers, diagnostic and reconfiguration procedures); sca lability of diversity (by changing types, depth and method of diversity selection). (Author)

  17. Safety critical FPGA-based NPP instrumentation and control systems: assessment, development and implementation

    International Nuclear Information System (INIS)

    Bakhmach, E. S.; Siora, A. A.; Tokarev, V. I.; Kharchenko, V. S.; Sklyar, V. V.; Andrashov, A. A.

    2010-10-01

    The stages of development, production, verification, licensing and implementation methods and technologies of safety critical instrumentation and control systems for nuclear power plants (NPP) based on FPGA (Field Programmable Gates Arrays) technologies are described. A life cycle model and multi-version technologies of dependability and safety assurance of FPGA-based instrumentation and control systems are discussed. An analysis of NPP instrumentation and control systems construction principles developed by Research and Production Corporation Radiy using FPGA-technologies and results of these systems implementation and operation at Ukrainian and Bulgarian NPP are presented. The RADIY TM platform has been designed and developed by Research and Production Corporation Radiy, Ukraine. The main peculiarity of the RADIY TM platform is the use of FPGA as programmable components for logic control operation. The FPGA-based RADIY TM platform used for NPP instrumentation and control systems development ensures sca lability of system functions types, volume and peculiarities (by changing quantity and quality of sensors, actuators, input/output signals and control algorithms); sca lability of dependability (safety integrity) (by changing a number of redundant channel, tiers, diagnostic and reconfiguration procedures); sca lability of diversity (by changing types, depth and method of diversity selection). (Author)

  18. Validation of Safety-Critical Systems for Aircraft Loss-of-Control Prevention and Recovery

    Science.gov (United States)

    Belcastro, Christine M.

    2012-01-01

    Validation of technologies developed for loss of control (LOC) prevention and recovery poses significant challenges. Aircraft LOC can result from a wide spectrum of hazards, often occurring in combination, which cannot be fully replicated during evaluation. Technologies developed for LOC prevention and recovery must therefore be effective under a wide variety of hazardous and uncertain conditions, and the validation framework must provide some measure of assurance that the new vehicle safety technologies do no harm (i.e., that they themselves do not introduce new safety risks). This paper summarizes a proposed validation framework for safety-critical systems, provides an overview of validation methods and tools developed by NASA to date within the Vehicle Systems Safety Project, and develops a preliminary set of test scenarios for the validation of technologies for LOC prevention and recovery

  19. Software quality assurance and software safety in the Biomed Control System

    International Nuclear Information System (INIS)

    Singh, R.P.; Chu, W.T.; Ludewigt, B.A.; Marks, K.M.; Nyman, M.A.; Renner, T.R.; Stradtner, R.

    1989-01-01

    The Biomed Control System is a hardware/software system used for the delivery, measurement and monitoring of heavy-ion beams in the patient treatment and biology experiment rooms in the Bevalac at the Lawrence Berkeley Laboratory (LBL). This paper describes some aspects of this system including historical background philosophy, configuration management, hardware features that facilitate software testing, software testing procedures, the release of new software quality assurance, safety and operator monitoring. 3 refs

  20. Assessment of shaft safety and management system of controlling engineering information

    Energy Technology Data Exchange (ETDEWEB)

    Liu Rui-xin; Xu Yan-chun [Yanzhou Mining Group Ltd., Zoucheng (China)

    2008-02-15

    Evaluating shaft safety and establishing a system for controlling engineering information is very important because more than 90 shafts in thick alluvial areas suddenly have shaft wall fracturing or breaking problems and there are more than a few hundred shafts of similar geologic conditions. Taking shaft control in the Yangzhou Coal Mining Group as an example, an assessment and management system and related software were established. This system includes basic information of the mine, measurement results and analysis, and functions of empirical and theoretical forecasting and finite element analysis, which are confirmed to be very effective for guiding shaft well control engineering in practice. 8 refs., 3 figs., 2 tabs.

  1. Development of FPGA-based safety-related instrumentation and control systems

    Energy Technology Data Exchange (ETDEWEB)

    Oda, N.; Tanaka, A.; Izumi, M.; Tarumi, T.; Sato, T. [Toshiba Corporation, Isogo Nuclear Engineering Center, Yokohama (Japan)

    2004-07-01

    Toshiba has developed systems which perform signal processing by field programmable gate arrays (FPGA) for safety-related instrumentation and control systems. FPGA is a device which consists only of defined digital circuit: hardware, which performs defined processing. FPGA-based system solves issues existing both in the conventional systems operated by analog circuits (analog-based system) and the systems operated by central processing units (CPU-based system). The advantages of applying FPGA are to keep the long-life supply of products, improving testability (verification), and to reduce the drift which may occur in analog-based system. Considering application to safety-related systems, nonvolatile and non rewritable FPGA which is impossible to be changed after once manufactured has been adopted in Toshiba FPGA-based system. The systems which Toshiba developed this time are Power range Monitor (PRM) and Trip Module (TM). These systems are compatible with the conventional analog-based systems and the CPU-based systems. Therefore, requested cost for upgrading will be minimized. Toshiba is planning to expand application of FPGA-based technology by adopting this development method to the other safety-related systems from now on. (authors)

  2. Physics related to control and safety of hybrid systems; Physique associee au controle et a la surete des systemes hybrides

    Energy Technology Data Exchange (ETDEWEB)

    Gueton, O

    2001-12-01

    Regarding nuclear waste management, ADS can be considered as large minor actinides burners. In a first part, a critical analysis of different reactor types shows that fast spectrum, helium coolant and nitride fuel, containing 100% minor actinides, agree perfectly with the high transmutation requirements of ADS. The control and safety demonstration of this system represents the main purpose of this study. Understanding spatial and dynamic behaviour of ADS flux is absolutely necessary. For this purpose, we have defined an indicator to quantify spatial decoupling. It shows, on the one hand, point kinetic deficiency to study local transients, and on the other hand, perturbations propagation differences between ADS and critical cores. Then, in a more concrete approach, accidental sequences (source transient, beam de-focalization, reactivity insertions, loss of flow, depressurization) are evaluated for this core, strongly loaded with minor actinides. It is shown that the automatic beam shutdown leads to preserve large safety margins for all studied transients. The accelerator emergency stop is induced by an unexpected evolution of the core control parameters. These parameters, except reactivity, can be directly measured in subcritical systems like in critical ones. Concerning reactivity, we suggest a new method for its absolute determination in ADS: at the time of reactor start-up, the reactivity must be calibrated by coupling two methods of relative reactivity measurements (pulsed source and Approached Source Multiplication) for successive subcritical levels. After that, the on-line follow-up of reactivity is obtained from this calibration like in a critical core. (authors)

  3. Study concerning the power plant control and safety equipment by integrated distributed systems

    International Nuclear Information System (INIS)

    Optea, I.; Oprea, M.; Stanescu, P.

    1995-01-01

    The paper deals with the trends existing in the field of nuclear control and safety equipment and systems, proposing a high-efficiency integrated system. In order to enhance the safety of the plant and reliability of the structure system and components, we present a concept based on the latest computer technology with an open, distributed system, connected by a local area network with high redundancy. A modern conception for the control and safety system is to integrate all the information related to the reactor protection, active engineered safeguard and auxiliary systems parameters, offering a fast flow of information between all the agencies concerned so that situations can be quickly assessed. The integrated distributed control is based on a high performance operating system for realtime applications, flexible enough for transparent networking and modular for demanding configurations. The general design considerations for nuclear reactors instrumentation reliability and testing methods for real-time functions under dynamic regime are presented. Taking into account the fast progress in information technology, we consider the replacement of the old instrumentation of Cernavoda-1 NPP by a modern integrated system as an economical and efficient solution for the next units. (Author) 20 Refs

  4. Design lessons from using programmable controllers in the MFTF-B personnel safety and interlocks system

    International Nuclear Information System (INIS)

    Branum, J.D.

    1983-01-01

    Applying programmable controllers in critical applications such as personnel safety and interlocks systems requires special considerations in the design of both hardware and software. All modern programmable controller systems feature extensive internal diagnostic capabilities to protect against problems such as program memory errors; however most, if not all present designs lack an intrinsic capability for detecting and countering failures on the field-side of their I/O modules. Many of the most common styles of I/O modules can also introduce potentially dangerous sneak circuits, even without component failure. This paper presents the most significant lessons learned to date in the design of the MFTF-B Personnel Safety and Interlocks System, which utilizes two non-redundant programmable controllers with over 800 I/O points each. Specific problems recognized during the design process as well as those discovered during initial testing and operation are discussed along with their specific solutions in hardware and software

  5. Probabilistic safety assessment for instrumentation and control systems in nuclear power plants: an overview

    International Nuclear Information System (INIS)

    Lu, Lixuan; Jiang, Jin

    2004-01-01

    Deregulation in the electricity market has resulted in a number of challenges in the nuclear power industry. Nuclear power plants must find innovative ways to remain competitive by reducing operating costs without jeopardizing safety. Instrumentation and Control (I and C) systems not only play important roles in plant operation, but also in reducing the cost of power generation while maintaining and/or enhancing safety. Therefore, it is extremely important that I and C systems are managed efficiently and economically. With the increasing use of digital technologies, new methods are needed to solve problems associated with various aspects of digital I and C systems. Probabilistic Safety Assessment (PSA) has proved to be an effective method for safety analysis and risk-based decisions, even though challenges are still present. This paper provides an overview of PSA applications in three areas of digital I and C systems in nuclear power plants. These areas are Graded Quality Assurance, Surveillance Testing, and Instrumentation and Control System Design. In addition, PSA application in the regulation of nuclear power plants that adopt digital I and C systems is also investigated. (author)

  6. Safety Systems

    Science.gov (United States)

    Halligan, Tom

    2009-01-01

    Colleges across the country are rising to the task by implementing safety programs, response strategies, and technologies intended to create a secure environment for teachers and students. Whether it is preparing and responding to a natural disaster, health emergency, or act of violence, more schools are making campus safety a top priority. At…

  7. Operation safety of control systems. Principles and methods; Surete de fonctionnement des systemes de commande. Principes et methodes

    Energy Technology Data Exchange (ETDEWEB)

    Aubry, J.F. [Institut National Polytechnique, 54 - Nancy (France); Chatelet, E. [Universite de Technologie de Troyes, 10 (France)

    2008-09-15

    This article presents the main operation safety methods that can be implemented to design safe control systems taking into account the behaviour of the different components with each other (binary 'operation/failure' behaviours, non-consistent behaviours and 'hidden' failures, dynamical behaviours and temporal aspects etc). To take into account these different behaviours, advanced qualitative and quantitative methods have to be used which are described in this article: 1 - qualitative methods of analysis: functional analysis, preliminary risk analysis, failure mode and failure effects analyses; 2 - quantitative study of systems operation safety: binary representation models, state space-based methods, event space-based methods; 3 - application to the design of control systems: safe specifications of a control system, qualitative analysis of operation safety, quantitative analysis, example of application; 4 - conclusion. (J.S.)

  8. A safety assessment methodology applied to CNS/ATM-based air traffic control system

    Energy Technology Data Exchange (ETDEWEB)

    Vismari, Lucio Flavio, E-mail: lucio.vismari@usp.b [Safety Analysis Group (GAS), School of Engineering at University of Sao Paulo (Poli-USP), Av. Prof. Luciano Gualberto, Trav.3, n.158, Predio da Engenharia de Eletricidade, Sala C2-32, CEP 05508-900, Sao Paulo (Brazil); Batista Camargo Junior, Joao, E-mail: joaocamargo@usp.b [Safety Analysis Group (GAS), School of Engineering at University of Sao Paulo (Poli-USP), Av. Prof. Luciano Gualberto, Trav.3, n.158, Predio da Engenharia de Eletricidade, Sala C2-32, CEP 05508-900, Sao Paulo (Brazil)

    2011-07-15

    In the last decades, the air traffic system has been changing to adapt itself to new social demands, mainly the safe growth of worldwide traffic capacity. Those changes are ruled by the Communication, Navigation, Surveillance/Air Traffic Management (CNS/ATM) paradigm , based on digital communication technologies (mainly satellites) as a way of improving communication, surveillance, navigation and air traffic management services. However, CNS/ATM poses new challenges and needs, mainly related to the safety assessment process. In face of these new challenges, and considering the main characteristics of the CNS/ATM, a methodology is proposed at this work by combining 'absolute' and 'relative' safety assessment methods adopted by the International Civil Aviation Organization (ICAO) in ICAO Doc.9689 , using Fluid Stochastic Petri Nets (FSPN) as the modeling formalism, and compares the safety metrics estimated from the simulation of both the proposed (in analysis) and the legacy system models. To demonstrate its usefulness, the proposed methodology was applied to the 'Automatic Dependent Surveillance-Broadcasting' (ADS-B) based air traffic control system. As conclusions, the proposed methodology assured to assess CNS/ATM system safety properties, in which FSPN formalism provides important modeling capabilities, and discrete event simulation allowing the estimation of the desired safety metric.

  9. A safety assessment methodology applied to CNS/ATM-based air traffic control system

    International Nuclear Information System (INIS)

    Vismari, Lucio Flavio; Batista Camargo Junior, Joao

    2011-01-01

    In the last decades, the air traffic system has been changing to adapt itself to new social demands, mainly the safe growth of worldwide traffic capacity. Those changes are ruled by the Communication, Navigation, Surveillance/Air Traffic Management (CNS/ATM) paradigm , based on digital communication technologies (mainly satellites) as a way of improving communication, surveillance, navigation and air traffic management services. However, CNS/ATM poses new challenges and needs, mainly related to the safety assessment process. In face of these new challenges, and considering the main characteristics of the CNS/ATM, a methodology is proposed at this work by combining 'absolute' and 'relative' safety assessment methods adopted by the International Civil Aviation Organization (ICAO) in ICAO Doc.9689 , using Fluid Stochastic Petri Nets (FSPN) as the modeling formalism, and compares the safety metrics estimated from the simulation of both the proposed (in analysis) and the legacy system models. To demonstrate its usefulness, the proposed methodology was applied to the 'Automatic Dependent Surveillance-Broadcasting' (ADS-B) based air traffic control system. As conclusions, the proposed methodology assured to assess CNS/ATM system safety properties, in which FSPN formalism provides important modeling capabilities, and discrete event simulation allowing the estimation of the desired safety metric.

  10. A Novel Control Algorithm for Integration of Active and Passive Vehicle Safety Systems in Frontal Collisions

    Directory of Open Access Journals (Sweden)

    Daniel Wallner

    2010-10-01

    Full Text Available The present paper investigates an approach to integrate active and passive safety systems of passenger cars. Worldwide, the introduction of Integrated Safety Systems and Advanced Driver Assistance Systems (ADAS is considered to continue the today

  11. Safety and security analysis for distributed control system in nuclear power plants

    International Nuclear Information System (INIS)

    Lu Zhigang; Liu Baoxu

    2011-01-01

    The Digital Distributed Control System (DCS) is the core that manages all monitoring and operation tasks in a Nuclear Power Plant (NPP). So, Digital Distributed Control System in Nuclear Power Plant has strict requirements for control and automation device safety and security due to many factors. In this article, factors of safety are analyzed firstly, while placing top priority on reliability, quality of supply and stability have also been carefully considered. In particular, advanced digital and electronic technologies are adopted to maintain sufficient reliability and supervisory capabilities in nuclear power plants. Then, security of networking and information technology have been remarked, several design methodologies considering the security characteristics are suggested. Methods and technologies of this article are being used in testing and evaluation for a real implement of a nuclear power plant in China. (author)

  12. Safety-related instrumentation and control systems for nuclear power plants

    International Nuclear Information System (INIS)

    1984-01-01

    This Safety Guide deals mainly with design requirements for those I and C systems that are important to safety but are not safety systems. The Guide is intended to expand paragraphs 3.1, 3.2 and 3.3 of the Code of Practice on Design for Safety of Nuclear Power Plants (IAEA Safety Series No.50-C-D) in the area of I and C systems important to safety and refers to them as safety-related I and C systems. It also gives guidance and enumerates requirements for multiplexing and the use of the digital computers employed in this area

  13. Safety control and risk management

    International Nuclear Information System (INIS)

    Rasmussen, J.

    1987-01-01

    The acceptable probability of major accidents in nuclear power is very small, and can not be determined from direct empirical evidence. Therefore, control of the level of safety is a complex problem. The difficulty is related to the fact that a variable, 'safety', which is not accessible to direct measurement, is to be tightly controlled. Control, therefore, depends on a systematic, analytical prediction of the target state, i.e., the level of safety, from indirect evidence. From a control theoretic point of view this means that safety is controlled by a system which includes openloop as well as closed loop control paths. The aim of the paper is to take a general systems view on the complex mechanisms involved in the control of safety of industrial installations like nuclear power. From this, the role of probabilistic risk analysis is evaluated and needs for further development discussed. (author)

  14. Maintaining and improving the control and safety systems for the Electromagnetic Calorimeter of the CMS experiment

    CERN Document Server

    Di Calafiori, Diogo Raphael; Dissertori, Günther; Holme, Oliver; Jovanovic, Dragoslav; Lustermann, Werner; Zelepoukine, Serguei

    2012-01-01

    This paper presents the current architecture of the control and safety systems designed and implemented for the Electromagnetic Calorimeter (ECAL) of the Compact Muon Solenoid (CMS) experiment at the Large Hadron Collider (LHC). An evaluation of system performance during all CMS physics data taking periods is reported, with emphasis on how software and hardware solutions are used to overcome limitations, whilst maintaining and improving reliability and robustness. The outcomes of the CMS ECAL Detector Control System (DCS) Software Analysis Project were a fundamental step towards the integration of all control system applications and the consequent piece-by-piece software improvements allowed a smooth transition to the latest revision of the system. The ongoing task of keeping the system in-line with new hardware technologies and software platforms specified by the CMS DCS Group is discussed. The structure of the comprehensive support service with detailed incident logging is presented in addition to a complet...

  15. Probabilistic safety assessment for instrumentation and control systems in nuclear power plants. A literature survey

    International Nuclear Information System (INIS)

    Lu, Lixuan; Jiang, Jin

    2003-01-01

    Deregulation in electricity market will create a great deal of challenges for Nuclear Power Plants (NPP). To stay competitive, NPP will need to find new ways to reduce their operation costs. In NPP, Instrumentation and Control (I and C) systems play an important role in reducing the cost of producing electricity while maintaining and/or enhancing safety. Therefore, it is extremely important that one should manage the I and C systems more efficiently and economically. Meanwhile, obsolescence problem associated with I and C systems encouraged the usage of advanced digital techniques in I and C systems. Thus, new methodologies are needed to analyze the reliability and determine the maintenance strategy for the digital I and C systems. Probabilistic Safety Assessment (PSA) has been probed to be a promising method to deal with this issue. This paper provides a literature survey on the development of digital I and C systems in NPP, followed by a detailed review of PSA including its benefits, limitations and the future direction of its development. Most importantly, potential applications of PSA in various aspects of I and C systems are brought into perspective throughout the paper. Furthermore, the applicability of PSA in the regulation of safety-related I and C systems is demonstrated. Detailed information on PSA applications in 1) the resource allocation for I and C systems: 2) the determination of surveillance testing strategies; and 3) I and C system designs, is provided. (author)

  16. Improving the safety and protective automatic actions of the CMS electromagnetic calorimeter detector control system

    CERN Document Server

    Jimenez Estupinan, Raul; Cirkovic, Predrag; Di Calafiori, Diogo Raphael; Dissertori, Guenther; Djambazov, Lubomir; Jovanovic, Dragoslav; Lustermann, Werner; Milenovic, Predrag; Zelepoukine, Serguei

    2017-01-01

    The CMS ECAL Detector Control System (DCS) features several monitoring mechanisms able to react and perform automatic actions based on pre-defined action matrices. The DCS is capable of early detection of anomalies inside the ECAL and on its off-detector support systems, triggering automatic actions to mitigate the impact of these events and preventing them from escalating to the safety system. The treatment of such events by the DCS allows for a faster recovery process, better understanding of the development of issues, and in most cases, actions with higher granularity than the safety system. This paper presents the details of the DCS automatic action mechanisms, as well as their evolution based on several years of CMS ECAL operations.

  17. Probabilistic safety assessment for digital instrumentation and control systems in nuclear power plants - a review

    International Nuclear Information System (INIS)

    Lu, L.; Jiang, J.

    2003-01-01

    Deregulation in electricity market has created a great deal of challenges for nuclear power industries [1]. To stay competitive, Nuclear Power Plants (NPPs) will have to find ways to reduce their operational costs and to improve the plant safety. Instrumentation and Control (I and C) systems play an important role in this regard. Thus, new methodologies need to be developed to manage the operation of I and C systems more economically without jeopardizing the overall plant safety. Probabilistic Safety Assessment (PSA) technique is one of the promising methods to deal with such an issue, because PSA analyzes various system operational issues from a probabilistic sense, rather than a worst-case approach. However, there are several limitations when PSA is applied to I and C systems directly. A possible solution to this problem can be found by incorporating PSA with several other approaches. To better understand the issues involved, an attempt has been made in this paper to carry out a literature survey on this and related subject, particularly the effort will be made on: 1) the development of digital I and C systems in NPP, 2) PSA and its potential benefits and limitations, and 3) applications of PSA in various aspects of I and C systems including the resource allocation, the determination of surveillance testing strategies and the design of I and C systems. Finally, some solutions to overcome the aforementioned obstacles when applying PSA in I and C systems are also examined critically. (author)

  18. Requirements and analysis of electromagnetic compatibility of safety-related instrumentation and control system in nuclear power plants

    International Nuclear Information System (INIS)

    Liu Sujuan

    2002-01-01

    The state-of-the-art instrumentation and control system and the influence of their application to the electromagnetic compatibility is analyzed. Based on the present situation of nuclear safety in China and relevant experiences from other countries, the author tries to probe into the requirements and test methods about how safety-related instrument and control system to accommodate electromagnetic interference, radio-frequency interference and power surges in the environments of nuclear power plant so as to develop Chinese safety standards

  19. ACCIDENT ANALYSES & CONTROL OPTIONS IN SUPPORT OF THE SLUDGE WATER SYSTEM SAFETY ANALYSIS

    Energy Technology Data Exchange (ETDEWEB)

    WILLIAMS, J.C.

    2003-11-15

    This report documents the accident analyses and nuclear safety control options for use in Revision 7 of HNF-SD-WM-SAR-062, ''K Basins Safety Analysis Report'' and Revision 4 of HNF-SD-SNF-TSR-001, ''Technical Safety Requirements - 100 KE and 100 KW Fuel Storage Basins''. These documents will define the authorization basis for Sludge Water System (SWS) operations. This report follows the guidance of DOE-STD-3009-94, ''Preparation Guide for US. Department of Energy Nonreactor Nuclear Facility Safety Analysis Reports'', for calculating onsite and offsite consequences. The accident analysis summary is shown in Table ES-1 below. While this document describes and discusses potential control options to either mitigate or prevent the accidents discussed herein, it should be made clear that the final control selection for any accident is determined and presented in HNF-SD-WM-SAR-062.

  20. Safety system status monitoring

    International Nuclear Information System (INIS)

    Lewis, J.R.; Morgenstern, M.H.; Rideout, T.H.; Cowley, P.J.

    1984-03-01

    The Pacific Northwest Laboratory has studied the safety aspects of monitoring the preoperational status of safety systems in nuclear power plants. The goals of the study were to assess for the NRC the effectiveness of current monitoring systems and procedures, to develop near-term guidelines for reducing human errors associated with monitoring safety system status, and to recommend a regulatory position on this issue. A review of safety system status monitoring practices indicated that current systems and procedures do not adequately aid control room operators in monitoring safety system status. This is true even of some systems and procedures installed to meet existing regulatory guidelines (Regulatory Guide 1.47). In consequence, this report suggests acceptance criteria for meeting the functional requirements of an adequate system for monitoring safety system status. Also suggested are near-term guidelines that could reduce the likelihood of human errors in specific, high-priority status monitoring tasks. It is recommended that (1) Regulatory Guide 1.47 be revised to address these acceptance criteria, and (2) the revised Regulatory Guide 1.47 be applied to all plants, including those built since the issuance of the original Regulatory Guide

  1. Safety system status monitoring

    Energy Technology Data Exchange (ETDEWEB)

    Lewis, J.R.; Morgenstern, M.H.; Rideout, T.H.; Cowley, P.J.

    1984-03-01

    The Pacific Northwest Laboratory has studied the safety aspects of monitoring the preoperational status of safety systems in nuclear power plants. The goals of the study were to assess for the NRC the effectiveness of current monitoring systems and procedures, to develop near-term guidelines for reducing human errors associated with monitoring safety system status, and to recommend a regulatory position on this issue. A review of safety system status monitoring practices indicated that current systems and procedures do not adequately aid control room operators in monitoring safety system status. This is true even of some systems and procedures installed to meet existing regulatory guidelines (Regulatory Guide 1.47). In consequence, this report suggests acceptance criteria for meeting the functional requirements of an adequate system for monitoring safety system status. Also suggested are near-term guidelines that could reduce the likelihood of human errors in specific, high-priority status monitoring tasks. It is recommended that (1) Regulatory Guide 1.47 be revised to address these acceptance criteria, and (2) the revised Regulatory Guide 1.47 be applied to all plants, including those built since the issuance of the original Regulatory Guide.

  2. Safety of High Speed Ground Transportation Systems : Analytical Methodology for Safety Validation of Computer Controlled Subsystems : Volume 2. Development of a Safety Validation Methodology

    Science.gov (United States)

    1995-01-01

    This report describes the development of a methodology designed to assure that a sufficiently high level of safety is achieved and maintained in computer-based systems which perform safety cortical functions in high-speed rail or magnetic levitation ...

  3. Reactivity control in HTR power plants with respect to passive safety system. Summary

    Energy Technology Data Exchange (ETDEWEB)

    Barnert, H; Kugeler, K [Forschungszentrum Juelich GmbH (Germany). Inst. fuer Sicherheitsforschung und Reaktortechnik

    1996-12-01

    The R and D and Demonstration of the High Temperature Reactor (HTR) is described in overview. The HTR-MODULE power plant, as the most advanced concept, is taken for the description of the reactivity control in general. The idea of the ``modularization of the core`` of the HTR has been developed as the answer on the experiences of the core melt accident at Three Miles Island. The HTR module has two shutdown systems: The ``6 rods``-system for hot shutdown at the ``18 small absorber pebbles units`` - system for cold shutdown. With respect to the definition of ``Passive Systems`` of IAEA-TECDOC-626 the total reactivity control system of the HTR-MODULE is a passive system of category D, because it is an emergency reactor shutdown system based on gravity driven rods, and devices, activated by fail-safe trip logic. But reactivity control of the HTR does not only consist of these engineered safety system but does have a self-acting stabilization by the negative temperature coefficient of the reactivity, being rather effective in reactivity control. Examples from computer calculations are presented, and, in addition, experimental results from the ``Stuck Rod Experiment`` at the AVR reactor in Juelich. On the basis of this the proposal is made that ``self-acting stabilization as a quality of the function`` should be discussed as a new category in addition to the active and passive engineered safety systems, structures and components of IAEA-TECDOC-626. The requirements for a future ``catastrophe-free`` nuclear technology are presented. In the appendix the 7th amendment of the atomic energy act of the Federal Republic of Germany, effective 28 July 94, is given. (author).

  4. A model of Occupational Safety and Health Management System (OSHMS) for promoting and controlling health and safety in textile industry.

    Science.gov (United States)

    Manimaran, S; Rajalakshmi, R; Bhagyalakshmi, K

    2015-01-01

    The development of Occupational Safety and Health Management System in textile industry will rejuvenate the workers and energize the economy as a whole. In India, especially in Tamil Nadu, approximately 1371 textile business is running with the help of 38,461 workers under Ginning, Spinning, Weaving, Garment and Dyeing sectors. Textile industry of contributes to the growth of Indian economy but it fails to foster education and health as key components of human development and help new democracies. The present work attempts to measure and develop OSHMS which reduce the hazards and risk involved in textile industry. Among all other industries textile industry is affected by enormous hazards and risk because of negligence by management and Government. It is evident that managements are not abiding by law when an accident has occurred. Managements are easily deceiving workers and least bothered about the Quality of Work Life (QWL). A detailed analysis of factors promoting safety and health to the workers has been done by performing confirmatory factor analysis, evaluating Risk Priority Number and the framework of OHMS has been conceptualized using Structural Equation Model. The data have been collected using questionnaire and interview method. The study finds occupation health for worker in Textile industry is affected not only by safety measure but also by technology and management. The work shows that difficulty in identifying the cause and effect of hazards, the influence of management in controlling and promoting OSHMS under various dimensions. One startling fact is existence of very low and insignificance correlation between health factors and outcome.

  5. Design and research of safety monitor and control system based on CAN BUS

    International Nuclear Information System (INIS)

    Wen Xinling; Chen Yu; Zhang Zhen; Zhao Yubin

    2007-01-01

    In Order to protect machine operator under danger work area in producing-manufacturing industry, we present a distributed safety monitor and control system based on CAN BUS technology. The detection signal is collected based on the photo-voltage characteristics of the infrared sensor and it was processed with the core of AT89C51. The microprocessor controls the CAN BUS controller SJA1000/transceiver PCA82C250 to structure CAN BUS communication system to transmit the data. Through the serial interface MAX232 connected main controller with each control node, PC can monitor and control each machine in real time and renew control scheme. This paper introduces composition principle and the methods of hardware design in detail. Experiments shown that the system has yield control precision of 0.1 mm, defend distance more than 15 m and the measurement accuracy of 100%. Moreover, it can realize to reform FA431 and monitor cotton-breaking, yarn-breaking and product quality. Productivity is improved about 25%-35%. (authors)

  6. PWR hybrid computer model for assessing the safety implications of control systems

    International Nuclear Information System (INIS)

    Smith, O.L.; Renier, J.P.; Difilippo, F.C.; Clapp, N.E.; Sozer, A.; Booth, R.S.; Craddick, W.G.; Morris, D.G.

    1986-03-01

    The ORNL study of safety-related aspects of nuclear power plant control systems consists of two interrelated tasks: (1) failure mode and effects analysis (FMEA) that identified single and multiple component failures that might lead to significant plant upsets and (2) computer models that used these failures as initial conditions and traced the dynamic impact on the control system and remainder of the plant. This report describes the simulation of Oconee Unit 1, the first plant analyzed. A first-principles, best-estimate model was developed and implemented on a hybrid computer consisting of AD-4 analog and PDP-10 digital machines. Controls were placed primarily on the analog to use its interactive capability to simulate operator action. 48 refs., 138 figs., 15 tabs

  7. Integrated Chassis Control System with Fail Safety Using Optimum Yaw Moment Distribution

    International Nuclear Information System (INIS)

    Yim, Seongjin

    2014-01-01

    This paper presents an integrated chassis control system with fail safety using optimum yaw moment distribution for a vehicle with steer-by-wire and brake-by-wire devices. The proposed system has two-level structure: upper- and lower-level controllers. In the upper-level controller, the control yaw moment is computed with sliding mode control theory. In the lower-level controller, the control yaw moment is distributed into the tire forces of active front steering(AFS) and electronic stability control(ESC) with the weighted pseudo-inverse based control allocation(WPCA) method. By setting the variable weights in WPCA, it is possible to take the sensor/actuator failure into account. In this framework, it is necessary to optimize the variables weights in order to enhance the yaw moment distribution. For this purpose, simulation-based tuning is proposed. To show the effectiveness of the proposed method, simulations are conducted on a vehicle simulation package, CarSim

  8. Integrated Chassis Control System with Fail Safety Using Optimum Yaw Moment Distribution

    Energy Technology Data Exchange (ETDEWEB)

    Yim, Seongjin [Seoul Nat' l Univ. of Sci. and Tech., Seoul (Korea, Republic of)

    2014-03-15

    This paper presents an integrated chassis control system with fail safety using optimum yaw moment distribution for a vehicle with steer-by-wire and brake-by-wire devices. The proposed system has two-level structure: upper- and lower-level controllers. In the upper-level controller, the control yaw moment is computed with sliding mode control theory. In the lower-level controller, the control yaw moment is distributed into the tire forces of active front steering(AFS) and electronic stability control(ESC) with the weighted pseudo-inverse based control allocation(WPCA) method. By setting the variable weights in WPCA, it is possible to take the sensor/actuator failure into account. In this framework, it is necessary to optimize the variables weights in order to enhance the yaw moment distribution. For this purpose, simulation-based tuning is proposed. To show the effectiveness of the proposed method, simulations are conducted on a vehicle simulation package, CarSim.

  9. Control maintenance training program for special safety systems at Bruce B

    International Nuclear Information System (INIS)

    Reinwald, G.

    1997-01-01

    It was recognized from the early days of commissioning of Bruce B that Control Maintenance staff would require a level of expertise to be able to maintain Special Safety Systems in proper running order. In the early 80's this was achieved through hands on experience during the original commissioning, troubleshooting and placing of the various systems in service. Control maintenance procedures were developed and implemented as the new systems came available for commissioning, as were operating manuals,training manuals etc. Under the development of the Maintenance Manager, a Conduct of Maintenance section was organized. One of the responsibilities of this section was to develop a series of Maintenance Administrative Procedures (MAPs) that set the standards for maintenance activities including training

  10. Means to improve underground coal mine safety by automated control of methane drainage systems

    Directory of Open Access Journals (Sweden)

    Babut Gabriel Bujor

    2017-01-01

    Full Text Available Based on the critical analysis of the presently employed management of methane drainage systems operation in Jiu Valley collieries, the paper aims to assess the basic elements required to develop an automated monitoring and control system of these. The results obtained after studies and researches carried out also allowed formulating certain proposals regarding the modification of manual control procedures of methane drainage systems operation, in order to correlate them with the prescriptions of legislation requirements from countries having a well-developed mining industry. Putting in practice the mentioned proposals could have immediate and beneficial effects on increasing the methane drainage process efficiency, leading meanwhile to an improved working environment and, implicitly, to a higher level of occupational safety and health in Jiu Valley collieries.

  11. Control and safety systems for TRIGA irradiation facilities C5 and C9

    International Nuclear Information System (INIS)

    Talpalariu Cornel Talpalariu Jeni Crucean Mircea Matei Corina

    2008-01-01

    Full text: The Institute for Nuclear Research conducted research for designing and manufacturing of microprocessor equipment for some irradiation facilities operating by the TRIGA reactor. This equipment has accumulated a wide operating time allowing the conclusions referring to reliability, ergonomics, and design of the operating facilities. Based upon these studies a new program was initiated for the design and manufacturing of a modern equipment with improved reliability and flexibility performances. The system provides the user with a multitude of options, numerical and analog interfaces, keyboard and high reliability local display. The main functional components of the system are: - 8 PID full options regulating loops; - 8 safety analog channels having 4 preset trips; - watch dog restart and fault tolerant facilities; - 8 high precision analog with an input of 0 - 15 mV from thermocouple; - 8 computer controlled power supplies of 220 V, 1 kWA; - alphanumeric display and keyboard; - fault tolerant analog scanner. A real improvement of the system is the future remote control computer, a PC AT Pentium working like a system controller, real time data acquisition, and operator's adviser. This new facility allows the operator to set the trips or to control remotely all the power supply and step-by-step positioner of irradiation device. Software design for acquisition and data processing provides modern techniques for operator interfacing, representation recording and protection of test results. Software implementation keeps a special organization supported by a real time executive that is the best method to achieve the performance required. Following this objective, the software structure consists of: 1. Tasks as follows: - testing parameters setup; - data processing routines; - engineering and electrical conversion; - numerical / graphical data representation; - test results recording routines and data base management. 2. Drivers as follows: - A/I and D

  12. Study of fieldbus technology confiability when applied in a Sterilization plant control and safety systems

    International Nuclear Information System (INIS)

    Karma, D.; Sampa, M.H.O.; Rela, P.R.

    2001-01-01

    Several sterilization processes have been used in these years for treatment of countless products. Some processes use high temperatures, thermal shocks and chemical agents. With the discovery of the ionizing radiation and its posterior technological developments turned possible the application of that process, in 1960, also in the sterilization, denominated radiation sterilization. This process became also applied in another areas of health and industrial as food conservation, gemstones enhancement and others. The radiation sterilization requests an effective control and it needs a high level of safety. The commercial use of the computers applied in industrial automation provides and the domain of new technologies in this field provides news applications then new designs now is possible. The Fieldbus technology, a new digital communication protocol, like a Local Area Network, can be an alternative in the cobalt-60 irradiation plant. This paper show preliminary study about confiability in systems using Fieldbus technology. This technology was simulated in sterilization plant control and safety systems and the fail probability was quantified using Fail Tree Analysis Method. Fieldbus technology can be used in sterilization plants because the confiability in this systems is like PLCs and relays systems, was the conclusion

  13. Outline of the requirements of application of computer based instrumentation and control systems in the systems important to safety on Bohunice NPPs

    International Nuclear Information System (INIS)

    Bacurik, J.

    1997-01-01

    The most important regulatory requirements and issues are described related to the review, evaluation and assessment of computer-based safety-related IandC systems, with emphasis on safety instrumentation and control. These aspects include safety classification and categorization of IandC, ranking of applicable codes and standards, design evaluation on the system level, and software assessment. (author)

  14. PWR hybrid computer model for assessing the safety implications of control systems

    International Nuclear Information System (INIS)

    Smith, O.L.; Booth, R.S.; Clapp, N.E.; DiFilippo, F.C.; Renier, J.P.; Sozer, A.

    1985-01-01

    The ORNL study of safety-related aspects of control systems consists of two interrelated tasks, (1) a failure mode and effects analysis that, in part, identifies single and multiple component failures that may lead to significant plant upsets, and (2) a hybrid computer model that uses these failures as initial conditions and traces the dynamic impact on the control system and remainder of the plant. The second task is reported here. The initial step in model development was to define a suitable interface between the FMEA and computer simulation tasks. This involved identifying primary plant components that must be simulated in dynamic detail and secondary components that can be treated adequately by the FMEA alone. The FMEA in general explores broader spectra of initiating events that may collapse into a reduced number of computer runs. A portion of the FMEA includes consideration of power supply failures. Consequences of the transients may feedback on the initiating causes, and there may be an interactive relationship between the FMEA and the computer simulation. Since the thrust of this program is to investigate control system behavior, the controls are modeled in detail to accurately reproduce characteristic response under normal and off-normal transients. The balance of the model, including neutronics, thermohydraulics and component submodels, is developed in sufficient detail to provide a suitable support for the control system

  15. KAERI software safety guideline for developing safety-critical software in digital instrumentation and control system of nuclear power plant

    International Nuclear Information System (INIS)

    Lee, Jang Soo; Kim, Jang Yeol; Eum, Heung Seop.

    1997-07-01

    Recently, the safety planning for safety-critical software systems is being recognized as the most important phase in the software life cycle, and being developed new regulatory positions and standards by the regulatory and the standardization organization. The requirements for software important to safety of nuclear reactor are described in such positions and standards. Most of them are describing mandatory requirements, what shall be done, for the safety-critical software. The developers of such a software. However, there have been a lot of controversial factors on whether the work practices satisfy the regulatory requirements, and to justify the safety of such a system developed by the work practices, between the licenser and the licensee. We believe it is caused by the reason that there is a gap between the mandatory requirements (What) and the work practices (How). We have developed a guidance to fill such gap, which can be useful for both licenser and licensee to conduct a justification of the safety in the planning phase of developing the software for nuclear reactor protection systems. (author). 67 refs., 13 tabs., 2 figs

  16. KAERI software safety guideline for developing safety-critical software in digital instrumentation and control system of nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jang Soo; Kim, Jang Yeol; Eum, Heung Seop

    1997-07-01

    Recently, the safety planning for safety-critical software systems is being recognized as the most important phase in the software life cycle, and being developed new regulatory positions and standards by the regulatory and the standardization organization. The requirements for software important to safety of nuclear reactor are described in such positions and standards. Most of them are describing mandatory requirements, what shall be done, for the safety-critical software. The developers of such a software. However, there have been a lot of controversial factors on whether the work practices satisfy the regulatory requirements, and to justify the safety of such a system developed by the work practices, between the licenser and the licensee. We believe it is caused by the reason that there is a gap between the mandatory requirements (What) and the work practices (How). We have developed a guidance to fill such gap, which can be useful for both licenser and licensee to conduct a justification of the safety in the planning phase of developing the software for nuclear reactor protection systems. (author). 67 refs., 13 tabs., 2 figs.

  17. Full scale impact testing for environmental and safety control of energy material shipping container systems

    International Nuclear Information System (INIS)

    Seagren, R.D.

    1978-01-01

    Heavily-shielded energy material shipping systems, similar in size and weight to those presently employed to transport irradiated reactor fuel elements, are being destructively tested under dynamic conditions. In these tests, the outer and inner steel shells interact in a complex manner with the massive biological shielding in the system. Results obtained from these tests provide needed information for new design concepts. Containment failure (and the resulting release of radioactive material to the environment which might occur in an extremely severe accident) is most likely through the seals and other ancillary features of the shipping systems. Analyses and experiments provide engineering data on the behavior of these shipping systems under severe accident conditions and information for predicting potential survivability and environmental control with a rational margin of safety

  18. Procedures for controlling the risks of reliability, safety, and availability of technical systems

    International Nuclear Information System (INIS)

    1987-01-01

    The reference book covers four sections. Apart from the fundamental aspects of the reliability problem, of risk and safety and the relevant criteria with regard to reliability, the material presented explains reliability in terms of maintenance, logistics and availability, and presents procedures for reliability assessment and determination of factors influencing the reliability, together with suggestions for systems technical integration. The reliability assessment consists of diagnostic and prognostic analyses. The section on factors influencing reliability discusses aspects of organisational structures, programme planning and control, and critical activities. (DG) [de

  19. Safety Information System Guide

    International Nuclear Information System (INIS)

    Bullock, M.G.

    1977-03-01

    This Guide provides guidelines for the design and evaluation of a working safety information system. For the relatively few safety professionals who have already adopted computer-based programs, this Guide may aid them in the evaluation of their present system. To those who intend to develop an information system, it will, hopefully, inspire new thinking and encourage steps towards systems safety management. For the line manager who is working where the action is, this Guide may provide insight on the importance of accident facts as a tool for moving ideas up the communication ladder where they will be heard and acted upon; where what he has to say will influence beneficial changes among those who plan and control his operations. In the design of a safety information system, it is suggested that the safety manager make friends with a computer expert or someone on the management team who has some feeling for, and understanding of, the art of information storage and retrieval as a new and better means for communication

  20. A Fiber Bragg Grating-Based Monitoring System for Roof Safety Control in Underground Coal Mining

    Directory of Open Access Journals (Sweden)

    Yiming Zhao

    2016-10-01

    Full Text Available Monitoring of roof activity is a primary measure adopted in the prevention of roof collapse accidents and functions to optimize and support the design of roadways in underground coalmines. However, traditional monitoring measures, such as using mechanical extensometers or electronic gauges, either require arduous underground labor or cannot function properly in the harsh underground environment. Therefore, in this paper, in order to break through this technological barrier, a novel monitoring system for roof safety control in underground coal mining, using fiber Bragg grating (FBG material as a perceived element and transmission medium, has been developed. Compared with traditional monitoring equipment, the developed, novel monitoring system has the advantages of providing accurate, reliable, and continuous online monitoring of roof activities in underground coal mining. This is expected to further enable the prevention of catastrophic roof collapse accidents. The system has been successfully implemented at a deep hazardous roadway in Zhuji Coal Mine, China. Monitoring results from the study site have demonstrated the advantages of FBG-based sensors over traditional monitoring approaches. The dynamic impacts of progressive face advance on roof displacement and stress have been accurately captured by the novel roadway roof activity and safety monitoring system, which provided essential references for roadway support and design of the mine.

  1. Density Control of Multi-Agent Systems with Safety Constraints: A Markov Chain Approach

    Science.gov (United States)

    Demirer, Nazli

    The control of systems with autonomous mobile agents has been a point of interest recently, with many applications like surveillance, coverage, searching over an area with probabilistic target locations or exploring an area. In all of these applications, the main goal of the swarm is to distribute itself over an operational space to achieve mission objectives specified by the density of swarm. This research focuses on the problem of controlling the distribution of multi-agent systems considering a hierarchical control structure where the whole swarm coordination is achieved at the high-level and individual vehicle/agent control is managed at the low-level. High-level coordination algorithms uses macroscopic models that describes the collective behavior of the whole swarm and specify the agent motion commands, whose execution will lead to the desired swarm behavior. The low-level control laws execute the motion to follow these commands at the agent level. The main objective of this research is to develop high-level decision control policies and algorithms to achieve physically realizable commanding of the agents by imposing mission constraints on the distribution. We also make some connections with decentralized low-level motion control. This dissertation proposes a Markov chain based method to control the density distribution of the whole system where the implementation can be achieved in a decentralized manner with no communication between agents since establishing communication with large number of agents is highly challenging. The ultimate goal is to guide the overall density distribution of the system to a prescribed steady-state desired distribution while satisfying desired transition and safety constraints. Here, the desired distribution is determined based on the mission requirements, for example in the application of area search, the desired distribution should match closely with the probabilistic target locations. The proposed method is applicable for both

  2. System safety education focused on flight safety

    Science.gov (United States)

    Holt, E.

    1971-01-01

    The measures necessary for achieving higher levels of system safety are analyzed with an eye toward maintaining the combat capability of the Air Force. Several education courses were provided for personnel involved in safety management. Data include: (1) Flight Safety Officer Course, (2) Advanced Safety Program Management, (3) Fundamentals of System Safety, and (4) Quantitative Methods of Safety Analysis.

  3. The use of microprocessors at TRIUMF in the control of radiation safety interlock systems

    International Nuclear Information System (INIS)

    King, L.

    1988-01-01

    At TRIUMF the cyclotron vault, all primary beam lines, and each experimental area has a dedicated control unit to manage the safety interlock control of the area lockup sequence, beam blocker drive and area access. Typically each area has 24 devices which are monitored to control 16 outputs. These control units (Area Safety Units) were first implemented through the use of relay logic. The relay logic was reliable but difficult to modify to incorporate changes to the areas. In 1979 it was decided to use microprocessors in the form of single board computers to control the Area Safety Units. The details of the hardware and software is discussed as well as the advantages of microprocessor control

  4. Contribution at the evaluation of safety softwares in nuclear power plants control systems

    International Nuclear Information System (INIS)

    Soubies, B.; Le Meur, M.; Henry, J.Y.; Boulc'h, J.

    1993-06-01

    The introduction of programmable systems such the SPIN (Numerical Integrated Protection System) has conducted at particular dispositions for the conception and the use of such systems. The utilization of such systems until 1983 has conducted at modifications in the maintenance procedures. The new methods used for the N4 project in the evaluation of safety softwares are given in this report

  5. 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

  6. Multi-core System Architecture for Safety-critical Control Applications

    DEFF Research Database (Denmark)

    Li, Gang

    and size, and high power consumption. Increasing the frequency of a processor is becoming painful now due to the explosive power consumption. Furthermore, components integrated into a single-core processor have to be certified to the highest SIL, due to that no isolation is provided in a traditional single...... certification cost. Meanwhile, hardware platforms with improved processing power are required to execute the applications of larger size. To tackle the two issues mentioned above, the state of the art approaches are using more Electronic Control Units (ECU) in a federated architecture or increasing......-core processor. A promising alternative to improve processing power and provide isolation is to adopt a multi-core architecture with on-chip isolation. In general, a specific multi-core architecture can facilitate the development and certification of safety-related systems, due to its physical isolation between...

  7. Safety logic systems of PFBR

    International Nuclear Information System (INIS)

    Sambasivan, S. Ilango

    2004-01-01

    Full text : PFBR is provided with two independent, fast acting and diverse shutdown systems to detect any abnormalities and to initiate safety action. Each system consists of sensors, signal processing systems, logics, drive mechanisms and absorber rods. The absorber rods of the first system are Control and Safety Rods (CSR) and that of the second are called as Diverse Safety Rods (DSR). There are nine CSR and three DSR. While CSR are used for startup, control of reactor power, controlled shutdown and SCRAM, the DSR are used only for SCRAM. The respective drive mechanisms are called as CSRDM and DSRDM. Each of these two systems is capable of executing the shutdown satisfactorily with single failure criteria. Two independent safety logic systems based on diverse principles have been designed for the two shut down systems. The analog outputs of the sensors of Core Monitoring Systems comprising of reactor flux monitoring, core temperature monitoring, failed fuel detection and core flow monitoring systems are processed and converted into binary signals depending on their instantaneous values. Safety logic systems receive the binary signals from these core-monitoring systems and process them logically to protect the reactor against postulated initiating events. Neutronic and power to flow (P/Q) signals form the inputs to safety logic system-I and temperature signals are inputs to the safety logic system II. Failed fuel detection signals are processed by both the shut down systems. The two logic systems to actuate the safety rods are also based on two diverse designs and implemented with solid-state devices to meet all the requirements of safety systems. Safety logic system I that caters to neutronic and P/Q signals is designed around combinational logic and has an on-line test facility to detect struck at faults. The second logic system is based on dynamic logic and hence is inherently safe. This paper gives an overview of the two logic systems that have been

  8. Engineered and Administrative Safety Systems for the Control of Prompt Radiation Hazards at Accelerator Facilities

    International Nuclear Information System (INIS)

    Liu, James C.; SLAC; Vylet, Vashek; Walker, Lawrence S.

    2007-01-01

    The ANSI N43.1 Standard, currently in revision (ANSI 2007), sets forth the requirements for accelerator facilities to provide adequate protection for the workers, the public and the environment from the hazards of ionizing radiation produced during and from accelerator operations. The Standard also recommends good practices that, when followed, provide a level of radiation protection consistent with those established for the accelerator communities. The N43.1 Standard is suitable for all accelerator facilities (using electron, positron, proton, or ion particle beams) capable of producing radiation, subject to federal or state regulations. The requirements (see word 'shall') and recommended practices (see word 'should') are prescribed in a graded approach that are commensurate with the complexity and hazard levels of the accelerator facility. Chapters 4, 5 and 6 of the N43.1 Standard address specially the Radiation Safety System (RSS), both engineered and administrative systems, to mitigate and control the prompt radiation hazards from accelerator operations. The RSS includes the Access Control System (ACS) and Radiation Control System (RCS). The main requirements and recommendations of the N43.1 Standard regarding the management, technical and operational aspects of the RSS are described and condensed in this report. Clearly some aspects of the RSS policies and practices at different facilities may differ in order to meet the practical needs for field implementation. A previous report (Liu et al. 2001a), which reviews and summarizes the RSS at five North American high-energy accelerator facilities, as well as the RSS references for the 5 labs (Drozdoff 2001; Gallegos 1996; Ipe and Liu 1992; Liu 1999; Liu 2001b; Rokni 1996; TJNAF 1994; Yotam et al. 1991), can be consulted for the actual RSS implementation at various laboratories. A comprehensive report describing the RSS at the Stanford Linear Accelerator Center (SLAC 2006) can also serve as a reference

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

    International Nuclear Information System (INIS)

    Nasr, M.; Nasser, O.

    1988-01-01

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

  10. Safety concepts and their implications with respect to systems, instrumentation (automatic) control and hardware

    International Nuclear Information System (INIS)

    Paziaud, A.; Walther, M.

    1982-01-01

    This overview of instrumentation and control in the French Nuclear Power Plants sets out the importance of safety requirements. As a matter of fact, the amount of equipment increases proportionally to the increase in safety requirements, resulting in higher costs in spite of the decrease in the prices of each component owing to the advance in electronics. However the improved reliability should improve the plant capacity factor and, as a consequence, improve both the power output and the safety which is often endangered by minor failures starting severe accidents. (orig.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1962-07-01

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

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

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  13. Lightweight monitoring and control system for coal mine safety using REST style.

    Science.gov (United States)

    Cheng, Bo; Cheng, Xin; Chen, Junliang

    2015-01-01

    The complex environment of a coal mine requires the underground environment, devices and miners to be constantly monitored to ensure safe coal production. However, existing coal mines do not meet these coverage requirements because blind spots occur when using a wired network. In this paper, we develop a Web-based, lightweight remote monitoring and control platform using a wireless sensor network (WSN) with the REST style to collect temperature, humidity and methane concentration data in a coal mine using sensor nodes. This platform also collects information on personnel positions inside the mine. We implement a RESTful application programming interface (API) that provides access to underground sensors and instruments through the Web such that underground coal mine physical devices can be easily interfaced to remote monitoring and control applications. We also implement three different scenarios for Web-based, lightweight remote monitoring and control of coal mine safety and measure and analyze the system performance. Finally, we present the conclusions from this study and discuss future work. Copyright © 2014 ISA. Published by Elsevier Ltd. All rights reserved.

  14. Regenerative braking strategies, vehicle safety and stability control systems: critical use-case proposals

    Science.gov (United States)

    Oleksowicz, Selim A.; Burnham, Keith J.; Southgate, Adam; McCoy, Chris; Waite, Gary; Hardwick, Graham; Harrington, Cian; McMurran, Ross

    2013-05-01

    The sustainable development of vehicle propulsion systems that have mainly focused on reduction of fuel consumption (i.e. CO2 emission) has led, not only to the development of systems connected with combustion processes but also to legislation and testing procedures. In recent years, the low carbon policy has made hybrid vehicles and fully electric vehicles (H/EVs) popular. The main virtue of these propulsion systems is their ability to restore some of the expended energy from kinetic movement, e.g. the braking process. Consequently new research and testing methods for H/EVs are currently being developed. This especially concerns the critical 'use-cases' for functionality tests within dynamic events for both virtual simulations, as well as real-time road tests. The use-case for conventional vehicles for numerical simulations and road tests are well established. However, the wide variety of tests and their great number (close to a thousand) creates a need for selection, in the first place, and the creation of critical use-cases suitable for testing H/EVs in both virtual and real-world environments. It is known that a marginal improvement in the regenerative braking ratio can significantly improve the vehicle range and, therefore, the economic cost of its operation. In modern vehicles, vehicle dynamics control systems play the principal role in safety, comfort and economic operation. Unfortunately, however, the existing standard road test scenarios are insufficient for H/EVs. Sector knowledge suggests that there are currently no agreed tests scenarios to fully investigate the effects of brake blending between conventional and regenerative braking as well as the regenerative braking interaction with active driving safety systems (ADSS). The paper presents seven manoeuvres, which are considered to be suitable and highly informative for the development and examination of H/EVs with regenerative braking capability. The critical manoeuvres presented are considered to be

  15. System architecture of Detector Control and safety for the ATLAS Inner Detector Upgrade

    International Nuclear Information System (INIS)

    Ferrere, D.; Kersten, S.

    2011-01-01

    In the current ATLAS Upgrade plan a new Inner Detector (ID) based upon silicon sensor technology is being considered. The operational monitoring and control of the ID will be very demanding. The Detector Control System (DCS) is a common tool that is essential for the operational safety of a system. Even at this early stage the DCS system architecture has to be defined such that it is well integrated and optimized for its later implementation and use. For example the DCS diagnostics for the front-end (FE) chips is a serious option being considered that needs an early requirement and specification definition. In addition one of the main constraints is the service reuse between the service patch panels of the ATLAS ID and the counting room that limits the number of electrical lines to be reused. Conceptual differences in terms of readout architecture and layout have been identified between the strip and the pixel detector that lead to two distinct architectures. Nevertheless, the limitation of available electrical lines going to the counting room as well as the low material budget requirements inside the ID volume are two major constraints that lead the ID to consider an on-detector radiation hard integrated circuitry for the slow control. At this stage of the project, the definitions of the logical actions and protocol for the ADCs of such a chip are still being specified. In addition the experience gained from the current ID will be essential for the guidance of tuning the future DCS architecture in the coming years.

  16. 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

  17. 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.)

  18. Model-based safety analysis of a control system using Simulink and Simscape extended models

    Directory of Open Access Journals (Sweden)

    Shao Nian

    2017-01-01

    Full Text Available The aircraft or system safety assessment process is an integral part of the overall aircraft development cycle. It is usually characterized by a very high timely and financial effort and can become a critical design driver in certain cases. Therefore, an increasing demand of effective methods to assist the safety assessment process arises within the aerospace community. One approach is the utilization of model-based technology, which is already well-established in the system development, for safety assessment purposes. This paper mainly describes a new tool for Model-Based Safety Analysis. A formal model for an example system is generated and enriched with extended models. Then, system safety analyses are performed on the model with the assistance of automation tools and compared to the results of a manual analysis. The objective of this paper is to improve the increasingly complex aircraft systems development process. This paper develops a new model-based analysis tool in Simulink/Simscape environment.

  19. Radiological safety and control

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Sea Young; Yoo, Y S; Lee, J C; Lee, T Y; Lee, J L; Kim, B W; Lee, B J; Chung, K K; Chung, R I; Kim, J S; Lee, H S; Han, Y D; Lee, J I; Lee, K C; Yoon, J H; Sul, C W; Kim, C K; Yoon, K S; Seo, K W; Yoon, Y C [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1996-01-01

    This report describes the annual results of radiological safety and control program of 1995. This program consists of working area monitoring including HANARO, personnel radiation monitoring, education for radiation protection. As a result, the objectives of radiation protection have been achieved satisfactorily through the activities mentioned above. Also, the calibration services were provided to insure accurate radiation measurement in the radiation working places. 21 figs., 39 tabs., 5 refs. (Author) .new.

  20. Enhancement of a radiation safety system through the use of a microprocessor-controlled speech synthesizer

    International Nuclear Information System (INIS)

    Keefe, D.J.; McDowell, W.P.

    1980-01-01

    A speech synthesizer is being used to differentiate eight separate safety alarms on a high energy accelerator at Argonne National Laboratory. A single board microcomputer monitors eight signals from an existing radiation safety logic circuit. The microcomputer is programmed to output the proper code at the proper time and sequence to a speech synthesizer which supplies the audio input to a local public address system. This eliminates the requirement for eight different alarm tones and the personnel training required to differentiate among them. A twenty-word vocabulary was found adequate to supply the necessary safety announcements. The article describes the techniques used to interface the speech synthesizer into the existing safety logic circuit

  1. Evaluation of safety implications of control systems in LWR nuclear power plants

    International Nuclear Information System (INIS)

    Szukiewicz, A.J.

    1989-06-01

    An in-depth evaluation was performed on non-safety-related control systems (see Section 1) that are typically used during normal plant operation on four nuclear steam supply system plants: a General Electric Company boiling-water reactor, a Westinghouse 3-loop pressurized-water reactor (PWR), a Babcock ampersand Wilcox Co. (B ampersand W) once-through steam generator PWR, and a Combustion Engineering PWR design. A study was also conducted to determine the generic applicability of the results to the class of plants represented by the specific plants analyzed. Generic conclusions were then developed. Steam generator and reactor vessel overfill events and reactor vessel overcooling events were identified as major classes of events having the potential to be more severe than previously analyzed. Specific substasks of this issue were to study these events to determine the need for preventive and/or mitigating design measures. This report describes the technical studies performed by the laboratories, the NRC staff assessment of the results, the generic applicability of the evaluations, and the technical findings resulting from these studies. This final report contains the staff's responses to, and resolution of, the public comments that were solicited and received before September 16,1988, in response to the draft reports issued for public comment on May 27, 1988. 39 refs, 1 fig., 7 tabs

  2. Introduction of the system of hazard analysis critical control point to ensure the safety of irradiated food

    International Nuclear Information System (INIS)

    Sajet, A.S.

    2014-01-01

    Hazard Analysis Critical Control Point (HACCP) is a preventive system for food safety. It identifies safety risks faced by food. Identified points are controlled ensuring product safety. Because of presence of many of the pathogenic microorganisms and parasites in food which caused cases of food poisoning and many diseases transmitted through food, the current methods of food production could not prevent food contamination or prevent the growth of these pathogens completely because of being a part of the normal flora in the environment. Irradiation technology helped to control diseases transmitted through food, caused by pathological microorganisms and parasites present in food. The application of a system based on risk analysis as a means of risk management in food chain, demonstrated the importance of food irradiation. (author)

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

    International Nuclear Information System (INIS)

    Kim, Ung Soo

    1997-02-01

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

  4. Safety control system and its interface to EPICS for the off-line front end of the SPES project

    International Nuclear Information System (INIS)

    Vasquez, J.; Andrighetto, A.; Bassato, G.; Costa, L.; Giacchini, M.; Bertocco, M.

    2012-01-01

    The SPES (Selective Production of Exotic Species) project is based on a facility for the production of neutron-rich radioactive ion beams using the isotope separation on-line technique. The SPES off-line front-end apparatus involves a number of subsystems and procedures that are potentially dangerous both for human operators and for the equipment. The high voltage power supply, the ion source complex power supplies, the target chamber handling systems and the laser source are some example of these subsystems. For that reason, a safety control system has been developed. It is based on Schneider Electrics Preventa family safety modules that control the power supply of critical subsystems in combination with safety detectors that monitor critical variables. A Programmable Logic Controller (PLC), model BMXP342020 from the Schneider Electrics Modicon M340 family, is used for monitoring the status of the system as well as controlling the sequence of some operations in automatic way. A touch screen, model XBTGT5330 from the Schneider Electrics Magelis family, is used as Human Machine Interface (HMI) and communicates with the PLC using MODBUS-TCP. Additionally, an interface to the EPICS control network was developed using a home-made MODBUS-TCP EPICS driver in order to integrate it to the control system of the Front End as well as present the status of the system to the users on the main control panel. (authors)

  5. SCOPE safety-controls optimization by performance evaluation: A systematic approach for safety-related decisions at the Hanford Tank Remediation System. Phase 1, final report

    Energy Technology Data Exchange (ETDEWEB)

    Bergeron, K.D.; Williams, D.C.; Slezak, S.E.; Young, M.L. [and others

    1996-12-01

    The Department of Energy`s Hanford Tank Waste Remediation system poses a significant challenge for hazard management because of the uncertainty that surrounds many of the variables that must be considered in decisions on safety and control strategies. As a result, site managers must often operate under excessively conservative and expensive assumptions. This report describes a systematic approach to quantifying the uncertainties surrounding the critical parameters in control decisions (e.g., condition of the tanks, kinds of wastes, types of possible accidents) through the use of expert elicitation methods. The results of the elicitations would then be used to build a decision support system and accident analysis model that would allow managers to see how different control strategies would affect the cost and safety of a facility configuration.

  6. SCOPE safety-controls optimization by performance evaluation: A systematic approach for safety-related decisions at the Hanford Tank Remediation System. Phase 1, final report

    International Nuclear Information System (INIS)

    Bergeron, K.D.; Williams, D.C.; Slezak, S.E.; Young, M.L.

    1996-12-01

    The Department of Energy's Hanford Tank Waste Remediation system poses a significant challenge for hazard management because of the uncertainty that surrounds many of the variables that must be considered in decisions on safety and control strategies. As a result, site managers must often operate under excessively conservative and expensive assumptions. This report describes a systematic approach to quantifying the uncertainties surrounding the critical parameters in control decisions (e.g., condition of the tanks, kinds of wastes, types of possible accidents) through the use of expert elicitation methods. The results of the elicitations would then be used to build a decision support system and accident analysis model that would allow managers to see how different control strategies would affect the cost and safety of a facility configuration

  7. Safety system function trends

    International Nuclear Information System (INIS)

    Johnson, C.

    1989-01-01

    This paper describes research to develop risk-based indicators of plant safety performance. One measure of the safety-performance of operating nuclear power plants is the unavailability of important safety systems. Brookhaven National Laboratory and Science Applications International Corporation are evaluating ways to aggregate train-level or component-level data to provide such an indicator. This type of indicator would respond to changes in plant safety margins faster than the currently used indicator of safety system unavailability (i.e., safety system failures reported in licensee event reports). Trends in the proposed indicator would be one indication of trends in plant safety performance and maintenance effectiveness. This paper summarizes the basis for such an indicator, identifies technical issues to be resolved, and illustrates the potential usefullness of such indicators by means of computer simulations and case studies

  8. Safety of mechanical devices. Safety of automation systems

    International Nuclear Information System (INIS)

    Pahl, G.; Schweizer, G.; Kapp, K.

    1985-01-01

    The paper deals with the classic procedures of safety engineering in the sectors mechanical engineering, electrical and energy engineering, construction and transport, medicine technology and process technology. Particular stress is laid on the safety of automation systems, control technology, protection of mechanical devices, reactor safety, mechanical constructions, transport systems, railway signalling devices, road traffic and protection at work in chemical plans. (DG) [de

  9. Towards the certification of non-deterministic control systems for safety-critical applications: analysing aviation analogies for possible certification strategies

    CSIR Research Space (South Africa)

    Burger, CR

    2011-11-01

    Full Text Available Current certification criteria for safety-critical systems exclude non-deterministic control systems. This paper investigates the feasibility of using human-like monitoring strategies to achieve safe non-deterministic control using multiple...

  10. Tests on instrumentation and control systems important to safety in nuclear power stations. Systempruefung der leittechnischen Einrichtungen des Sicherheitssystems in Kernkraftwerken

    Energy Technology Data Exchange (ETDEWEB)

    1985-01-01

    The rule applies to the reactor protection system, to the protection and state boundaries, to control devices important to safety, and to danger alarms of the classes S and I. The system inspection of the control devices of the safety system comprises in-service testing and recurrent testing.

  11. Nuclear power plants - Instrumentation and control systems important for safety - Classification (International Electrotechnical Commission Standard Publication 1226:1993)

    International Nuclear Information System (INIS)

    Stefanik, J.

    1996-01-01

    This international standard established a method of classification of the information and command functions for nuclear power plants, and the I and C and equipment that provide those functions, into categories that designate the importance for safety of the functions, and the associated systems and equipment. The resulting classification then determines relevant design criteria. The design criteria are the measures of quality by which the adequacy of each functions, and the associated systems and equipment in relation to its importance to plant safety is ensured. In this standard, the criteria are those of functionality, reliability, performance, environmental durability and quality assurance. This standard is applicable to all the information and command functions, and the instrumentation and control systems and equipment that provide those functions. The functions, systems and equipment under consideration provide automated protection, closed or open loop control, and information to the operating staff. They keep the NPP conditions inside the safe operating envelope and provide automatic actions, or enable manual actions, that mitigate accidents or prevent or minimize radioactive releases to the site or wider environment. The functions, and the associated systems and equipment that fulfill these roles safeguard the health and safety of the NPP operators and the public. This standard complements, and does not replace or supersede, the Safety Guides and Codes of Practice published by the International Atomic Energy Agency

  12. Compositional Synthesis of Safety Controllers

    NARCIS (Netherlands)

    Kuijper, W.

    2012-01-01

    In my thesis I investigate compositional techniques for synthesis of safety controllers. A safety controller, in this context, is a state machine that gives the set of safe control outputs for every possible sequence of observations from the plant under control. Compositionality, in this context,

  13. Microbiological performance of Hazard Analysis Critical Control Point (HACCP)-based food safety management systems: A case of Nile perch processing company

    NARCIS (Netherlands)

    Kussaga, J.B.; Luning, P.A.; Tiisekwa, B.P.M.; Jacxsens, L.

    2017-01-01

    This study aimed at giving insight into microbiological safety output of a Hazard Analysis Critical Control Point (HACCP)-based Food Safety Management System (FSMS) of a Nile perch exporting company by using a combined assessment, This study aimed at giving insight into microbiological safety output

  14. LOFT integral test system final safety analysis report

    International Nuclear Information System (INIS)

    1974-03-01

    Safety analyses are presented for the following LOFT Reactor systems: engineering safety features; support buildings and facilities; instrumentation and controls; electrical systems; and auxiliary systems. (JWR)

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

    International Nuclear Information System (INIS)

    Ridha Roslan; Nik Mohd Faiz Khairuddin

    2013-01-01

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

  16. 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

  17. NASA aviation safety reporting system

    Science.gov (United States)

    1981-01-01

    Aviation safety reports that relate to loss of control in flight, problems that occur as a result of similar sounding alphanumerics, and pilot incapacitation are presented. Problems related to the go around maneuver in air carrier operations, and bulletins (and FAA responses to them) that pertain to air traffic control systems and procedures are included.

  18. The verification methodologies for a software modeling of Engineered Safety Features- Component Control System (ESF-CCS)

    International Nuclear Information System (INIS)

    Lee, Young-Jun; Cheon, Se-Woo; Cha, Kyung-Ho; Park, Gee-Yong; Kwon, Kee-Choon

    2007-01-01

    The safety of a software is not guaranteed through a simple testing of the software. The testing reviews only the static functions of a software. The behavior, dynamic state of a software is not reviewed by a software testing. The Ariane5 rocket accident and the failure of the Virtual Case File Project are determined by a software fault. Although this software was tested thoroughly, the potential errors existed internally. There are a lot of methods to solve these problems. One of the methods is a formal methodology. It describes the software requirements as a formal specification during a software life cycle and verifies a specified design. This paper suggests the methods which verify the design to be described as a formal specification. We adapt these methods to the software of a ESF-CCS (Engineered Safety Features-Component Control System) and use the SCADE (Safety Critical Application Development Environment) tool for adopting the suggested verification methods

  19. Research on the evaluation model of the software reliability in nuclear safety class digital instrumentation and control system

    International Nuclear Information System (INIS)

    Liu Ying; Yang Ming; Li Fengjun; Ma Zhanguo; Zeng Hai

    2014-01-01

    In order to analyze the software reliability (SR) in nuclear safety class digital instrumentation and control system (D-I and C), firstly, the international software design standards were analyzed, the standards' framework was built, and we found that the D-I and C software standards should follow the NUREG-0800 BTP7-14, according to the NRC NUREG-0800 review of requirements. Secondly, the quantitative evaluation model of SR using Bayesian Belief Network and thirteen sub-model frameworks were established. Thirdly, each sub-models and the weight of corresponding indexes in the evaluation model were analyzed. Finally, the safety case was introduced. The models lay a foundation for review and quantitative evaluation on the SR in nuclear safety class D-I and C. (authors)

  20. Efficiency of the functioning of the state control system for the safety and quality of animal products in Ukraine

    Directory of Open Access Journals (Sweden)

    I. Kyryliuk

    2017-12-01

    Full Text Available The study reveals the results of evaluating the effectiveness of the state control system (supervision on the safety and individual indicators of the quality of livestock products in Ukraine. The necessity of application of such components of efficiency as legislation, management and its organizational structure, inspection and laboratory service, information, training and communications is substantiated. It has been determined that during a sufficiently long period of time (until 2015, the system of state control (supervision was archaic and actually focused on the principles of command and administrative economy. The modern tendencies and specifics of the improvement of the Ukrainian control system in the direction of its harmonization with the European one are shown. The emphasis was on significant volumes of work that needed to be done in a very short time, as well as in the absence of adequate funding and appropriate skilled specialists. The emergence of clarity and unambiguousness in determining the responsibility of market operators for violating the legislation requirements in the field of production and circulation of animal origin food products was emphasized. Along with the achievements, there were identified systemic problems related to the technical regulation of safety assurance processes and individual quality indicators in Ukraine. Also it was noted and revealed that legislation in the area of guaranteeing the quality and safety of livestock products in Ukraine remains incomplete and not fully developed. The necessity of development of a number of by-laws and allocation of necessary financing for effective functioning of the state control system over product safety is substantiated. Article specified on the presence of insufficient number of professional inspection and laboratory services is underlined. The mechanisms of avoiding corruption risks and excessive pressure on the subjects of the livestock production market are

  1. Modeling goals and functions of control and safety systems - theoretical foundations and extensions of MFM

    International Nuclear Information System (INIS)

    Lind, M.

    2005-10-01

    Multilevel Flow Modeling (MFM) has proven to be an effective modeling tool for reasoning about plant failure and control strategies and is currently exploited for operator support in diagnosis and on-line alarm analysis. Previous MFM research was focussed on representing goals and functions of process plants which generate, transform and distribute mass and energy. However, only a limited consideration has been given to the problems of modeling the control systems. Control functions are indispensable for operating any industrial plant. But modeling of control system functions has proven to be a more challenging problem than modeling functions of energy and mass processes. The problems were discussed by Lind and tentative solutions has been proposed but have not been investigated in depth until recently, partly due to the lack of an appropriate theoretical foundation. The purposes of the present report are to show that such a theoretical foundation for modeling goals and functions of control systems can be built from concepts and theories of action developed by Von Wright and to show how the theoretical foundation can be used to extend MFM with concepts for modeling control systems. The theoretical foundations has been presented in detail elsewhere by the present author without the particular focus on modeling control actions and MFM adopted here. (au)

  2. Modeling goals and functions of control and safety systems -theoretical foundations and extensions of MFM

    Energy Technology Data Exchange (ETDEWEB)

    Lind, M. [Oersted - DTU, Kgs. Lyngby (Denmark)

    2005-10-01

    Multilevel Flow Modeling (MFM) has proven to be an effective modeling tool for reasoning about plant failure and control strategies and is currently exploited for operator support in diagnosis and on-line alarm analysis. Previous MFM research was focussed on representing goals and functions of process plants which generate, transform and distribute mass and energy. However, only a limited consideration has been given to the problems of modeling the control systems. Control functions are indispensable for operating any industrial plant. But modeling of control system functions has proven to be a more challenging problem than modeling functions of energy and mass processes. The problems were discussed by Lind and tentative solutions has been proposed but have not been investigated in depth until recently, partly due to the lack of an appropriate theoretical foundation. The purposes of the present report are to show that such a theoretical foundation for modeling goals and functions of control systems can be built from concepts and theories of action developed by Von Wright and to show how the theoretical foundation can be used to extend MFM with concepts for modeling control systems. The theoretical foundations has been presented in detail elsewhere by the present author without the particular focus on modeling control actions and MFM adopted here. (au)

  3. A holistic strategy for quality and safety control of traditional Chinese medicines by the “iVarious” standard system

    Directory of Open Access Journals (Sweden)

    Anzhen Chen

    2017-10-01

    Full Text Available An effective quality control system is the key to ensuring the quality, safety and efficacy of traditional Chinese medicines (TCMs. However, the current quality standard research lacks the top-design and systematic design, mostly based on specific technologies or evaluation methods. To resolve the challenges and questions of quality control of TCMs, a brand-new quality standard system, named “iVarious”, was proposed. The system comprises eight elements in a modular format. Meaning of every element was specifically illustrated via corresponding research instances. Furthermore, frankincense study was taken as an example for demonstrating standards and research process, based on the “iVarious” system. This system highlighted a holistic strategy for effectiveness, security, integrity and systematization of quality and safety control standards of TCMs. The establishment of “iVarious” integrates multi-disciplinary technologies and progressive methods, basis elements and key points of standard construction. The system provides a novel idea and technological demonstration for regulation establishment of TCMs quality standards.

  4. A holistic strategy for quality and safety control of traditional Chinese medicines by the "iVarious" standard system.

    Science.gov (United States)

    Chen, Anzhen; Sun, Lei; Yuan, Hang; Wu, Aiying; Lu, Jingguang; Ma, Shuangcheng

    2017-10-01

    An effective quality control system is the key to ensuring the quality, safety and efficacy of traditional Chinese medicines (TCMs). However, the current quality standard research lacks the top-design and systematic design, mostly based on specific technologies or evaluation methods. To resolve the challenges and questions of quality control of TCMs, a brand-new quality standard system, named "iVarious", was proposed. The system comprises eight elements in a modular format. Meaning of every element was specifically illustrated via corresponding research instances. Furthermore, frankincense study was taken as an example for demonstrating standards and research process, based on the "iVarious" system. This system highlighted a holistic strategy for effectiveness, security, integrity and systematization of quality and safety control standards of TCMs. The establishment of "iVarious" integrates multi-disciplinary technologies and progressive methods, basis elements and key points of standard construction. The system provides a novel idea and technological demonstration for regulation establishment of TCMs quality standards.

  5. Controls in new construction reactors-factory testing of the non-safety portion of the Lungmen nuclear power plant distributed control system

    International Nuclear Information System (INIS)

    Wu, Y. S.; Dick, J. W.; Tetirick, C. W.

    2006-01-01

    The construction permit for Taipower's Lungmen Nuclear Units 1 and 2, two ABWR plants, was issued on March 17, 1999[1], The construction of these units is progressing actively at site. The digital I and C system supplied by GE, which is designated as the Distributed Control and Information System (DCIS) in this project, is being implemented primarily at one vendor facility. In order to ensure the reliability, safety and availability of the DCIS, it is required to comprehensively test the whole DCIS in factory. This article describes the test requirements and acceptance criteria for functional testing of the Non-Safety Distributed Control and Information system (DCIS) for Taiwan Power's Lungmen Units 1 and 2 GE selected Invensys as the equipment supplier for this Non-Safety portion of DCIS. The DCIS system of the Lungmen Units is a physically distributed control system. Field transmitters are connected to hard I/O terminal inputs on the Invensys I/A system. Once the signal is digitized on FBMs (Field Bus Modules) in Remote Multiplexing Units (RMUs), the signal is passed into an integrated control software environment. Control is based on the concept of compounds and blocks where each compound is a logical collection of blocks that performs a control function. Each point identified by control compound and block can be individually used throughout the DCIS system by referencing its unique name. In the Lungmen Project control logic and HSI (Human System Interface) requirements are divided into individual process systems called MPLs (Master Parts List). Higher-level Plant Computer System (PCS) algorithms access control compounds and blocks in these MPLs to develop functions. The test requirements and acceptance criteria for the DCIS system of the Lungmen Project are divided into three general categories (see 1,2,3 below) of verification, which in turn are divided into several specific tests: 1. DCIS System Physical Checks a) RMU Test - To confirm that the hard I

  6. Controls in new construction reactors-factory testing of the non-safety portion of the Lungmen nuclear power plant distributed control system

    Energy Technology Data Exchange (ETDEWEB)

    Wu, Y. S. [Taiwan Power Company, 242, Roosevelt Road, Sec. 3, Taipei 100, Taiwan (China); Dick, J. W. [Invensys System Inc., 33 Commercial St., Foxboro, MA 02035 (United States); Tetirick, C. W. [GE Energy, 1989 Little Orchard Street, San Jose, CA 95125-1030 (United States)

    2006-07-01

    The construction permit for Taipower's Lungmen Nuclear Units 1 and 2, two ABWR plants, was issued on March 17, 1999[1], The construction of these units is progressing actively at site. The digital I and C system supplied by GE, which is designated as the Distributed Control and Information System (DCIS) in this project, is being implemented primarily at one vendor facility. In order to ensure the reliability, safety and availability of the DCIS, it is required to comprehensively test the whole DCIS in factory. This article describes the test requirements and acceptance criteria for functional testing of the Non-Safety Distributed Control and Information system (DCIS) for Taiwan Power's Lungmen Units 1 and 2 GE selected Invensys as the equipment supplier for this Non-Safety portion of DCIS. The DCIS system of the Lungmen Units is a physically distributed control system. Field transmitters are connected to hard I/O terminal inputs on the Invensys I/A system. Once the signal is digitized on FBMs (Field Bus Modules) in Remote Multiplexing Units (RMUs), the signal is passed into an integrated control software environment. Control is based on the concept of compounds and blocks where each compound is a logical collection of blocks that performs a control function. Each point identified by control compound and block can be individually used throughout the DCIS system by referencing its unique name. In the Lungmen Project control logic and HSI (Human System Interface) requirements are divided into individual process systems called MPLs (Master Parts List). Higher-level Plant Computer System (PCS) algorithms access control compounds and blocks in these MPLs to develop functions. The test requirements and acceptance criteria for the DCIS system of the Lungmen Project are divided into three general categories (see 1,2,3 below) of verification, which in turn are divided into several specific tests: 1. DCIS System Physical Checks a) RMU Test - To confirm that the hard

  7. Study of fieldbus technology applied in a sterilization plant control and safety systems

    International Nuclear Information System (INIS)

    Karam Junior, Dib

    2000-01-01

    Several sterilization processes have been used in these years for treatment of countless products. Some processes use high temperatures, thermal shocks and chemical agents. With the discovery of the ionizing radiation and its posterior technological developments turned possible application of that process, in 1960, also in the the sterilization, denominated radiation sterilization. This process became also applied in another areas of health and industrial as food conservation, gemstones enhancement and others. The radiation sterilization requests an effective control and it needs a high level of safety. The commercial use of the computers applied in industrial automation provides and the domain of new technologies in this field provides new applications then new designs now is possible. The Fieldbus technology, a new digital communication protocol, like a Local Area Network, can be an alternative in the cobalt-60 irradiation plant. The present work suggests, evaluates, qualifies and quantifies this possibility. (author)

  8. Optimal Design of Safety Instrumented Systems for Pressure Control of Methanol Separation Columns in the Bisphenol a Manufacturing Process

    Directory of Open Access Journals (Sweden)

    In-Bok Lee

    2016-12-01

    Full Text Available A bisphenol A production plant possesses considerable potential risks in the top of the methanol separation column, as pressurized acetone, methanol, and water are processed at an elevated temperature, especially in the event of an abnormal pressure increase due to a sudden power outage. This study assesses the potential risks in the methanol separation column through hazard and operability assessments and evaluates the damages in the case of fire and explosion accident scenarios. The study chooses three leakage scenarios: a 5-mm puncture on the methanol separation column, a 50-mm diameter fracture of a discharge pipe and a catastrophic rupture, and, simulated using Phast (Ver. 6.531, the concentration distribution of scattered methanol, thermal radiation distribution of fires, and overpressure distribution of vapor cloud explosions. Implementation of a safety-instrumented system equipped with two-out-of-three voting as a safety measure can detect overpressure at the top of the column and shut down the main control valve and the emergency shutoff valve simultaneously. By applying a safety integrity level of three, the maximal release volume of the safety relief valve can be reduced and, therefore, the design capacity of the flare stack can also be reduced. Such integration will lead to improved safety at a reduced cost.

  9. Software system safety

    Science.gov (United States)

    Uber, James G.

    1988-01-01

    Software itself is not hazardous, but since software and hardware share common interfaces there is an opportunity for software to create hazards. Further, these software systems are complex, and proven methods for the design, analysis, and measurement of software safety are not yet available. Some past software failures, future NASA software trends, software engineering methods, and tools and techniques for various software safety analyses are reviewed. Recommendations to NASA are made based on this review.

  10. 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

  11. 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

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

    International Nuclear Information System (INIS)

    Coura, J.G.

    1986-01-01

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

  13. Experience on operational safety improvement of control and operation support systems

    International Nuclear Information System (INIS)

    Itoh, N.; Nakagawa, T.; Mano, K.

    1988-01-01

    Japanese nuclear industry started in 1956 and about 30 years have passed since that time. Through these years, we have made a lot of efforts and developments in the field of Control and Instrumentation (C and I) system. The above 30 years and following years can be divided into four major periods. The first one is the period of research, the second of domestic production, the third of improvement, and the fourth of advancement. Improvements of C and I system, which we have made in those periods have made a great contribution to enhancement of reliability, availability and operability of nuclear power plants. Fig. 1 shows TEPCO's nuclear power plant (BWR) construction experience and technical trend of C and I system in Japan. This paper is to introduce the efforts and operational experience on control and operation support systems

  14. Programmable electronic safety systems

    International Nuclear Information System (INIS)

    Parry, R.R.

    1993-01-01

    Traditionally safety systems intended for protecting personnel from electrical and radiation hazards at particle accelerator laboratories have made extensive use of electromechanical relays. These systems have the advantage of high reliability and allow the designer to easily implement fail-safe circuits. Relay based systems are also typically simple to design, implement, and test. As systems, such as those presently under development at the Superconducting Super Collider Laboratory (SSCL), increase in size, and the number of monitored points escalates, relay based systems become cumbersome and inadequate. The move toward Programmable Electronic Safety Systems is becoming more widespread and accepted. In developing these systems there are numerous precautions the designer must be concerned with. Designing fail-safe electronic systems with predictable failure states is difficult at best. Redundancy and self-testing are prime examples of features that should be implemented to circumvent and/or detect failures. Programmable systems also require software which is yet another point of failure and a matter of great concern. Therefore the designer must be concerned with both hardware and software failures and build in the means to assure safe operation or shutdown during failures. This paper describes features that should be considered in developing safety systems and describes a system recently installed at the Accelerator Systems String Test (ASST) facility of the SSCL

  15. Programmable Electronic Safety Systems

    International Nuclear Information System (INIS)

    Parry, R.

    1993-05-01

    Traditionally safety systems intended for protecting personnel from electrical and radiation hazards at particle accelerator laboratories have made extensive use of electromechanical relays. These systems have the advantage of high reliability and allow the designer to easily implement failsafe circuits. Relay based systems are also typically simple to design, implement, and test. As systems, such as those presently under development at the Superconducting Super Collider Laboratory (SSCL), increase in size, and the number of monitored points escalates, relay based systems become cumbersome and inadequate. The move toward Programmable Electronic Safety Systems is becoming more widespread and accepted. In developing these systems there are numerous precautions the designer must be concerned with. Designing fail-safe electronic systems with predictable failure states is difficult at best. Redundancy and self-testing are prime examples of features that should be implemented to circumvent and/or detect failures. Programmable systems also require software which is yet another point of failure and a matter of great concern. Therefore the designer must be concerned with both hardware and software failures and build in the means to assure safe operation or shutdown during failures. This paper describes features that should be considered in developing safety systems and describes a system recently installed at the Accelerator Systems String Test (ASST) facility of the SSCL

  16. Optimisation of active suspension control inputs for improved performance of active safety systems

    Science.gov (United States)

    Čorić, Mirko; Deur, Joško; Xu, Li; Tseng, H. Eric; Hrovat, Davor

    2018-01-01

    A collocation-type control variable optimisation method is used to investigate the extent to which the fully active suspension (FAS) can be applied to improve the vehicle electronic stability control (ESC) performance and reduce the braking distance. First, the optimisation approach is applied to the scenario of vehicle stabilisation during the sine-with-dwell manoeuvre. The results are used to provide insights into different FAS control mechanisms for vehicle performance improvements related to responsiveness and yaw rate error reduction indices. The FAS control performance is compared to performances of the standard ESC system, optimal active brake system and combined FAS and ESC configuration. Second, the optimisation approach is employed to the task of FAS-based braking distance reduction for straight-line vehicle motion. Here, the scenarios of uniform and longitudinally or laterally non-uniform tyre-road friction coefficient are considered. The influences of limited anti-lock braking system (ABS) actuator bandwidth and limit-cycle ABS behaviour are also analysed. The optimisation results indicate that the FAS can provide competitive stabilisation performance and improved agility when compared to the ESC system, and that it can reduce the braking distance by up to 5% for distinctively non-uniform friction conditions.

  17. Practical Applications of Cosmic Ray Science: Spacecraft, Aircraft, Ground Based Computation and Control Systems and Human Health and Safety

    Science.gov (United States)

    Atwell, William; Koontz, Steve; Normand, Eugene

    2012-01-01

    In this paper we review the discovery of cosmic ray effects on the performance and reliability of microelectronic systems as well as on human health and safety, as well as the development of the engineering and health science tools used to evaluate and mitigate cosmic ray effects in earth surface, atmospheric flight, and space flight environments. Three twentieth century technological developments, 1) high altitude commercial and military aircraft; 2) manned and unmanned spacecraft; and 3) increasingly complex and sensitive solid state micro-electronics systems, have driven an ongoing evolution of basic cosmic ray science into a set of practical engineering tools (e.g. ground based test methods as well as high energy particle transport and reaction codes) needed to design, test, and verify the safety and reliability of modern complex electronic systems as well as effects on human health and safety. The effects of primary cosmic ray particles, and secondary particle showers produced by nuclear reactions with spacecraft materials, can determine the design and verification processes (as well as the total dollar cost) for manned and unmanned spacecraft avionics systems. Similar considerations apply to commercial and military aircraft operating at high latitudes and altitudes near the atmospheric Pfotzer maximum. Even ground based computational and controls systems can be negatively affected by secondary particle showers at the Earth's surface, especially if the net target area of the sensitive electronic system components is large. Accumulation of both primary cosmic ray and secondary cosmic ray induced particle shower radiation dose is an important health and safety consideration for commercial or military air crews operating at high altitude/latitude and is also one of the most important factors presently limiting manned space flight operations beyond low-Earth orbit (LEO).

  18. CERN safety system monitoring - SSM

    International Nuclear Information System (INIS)

    Hakulinen, T.; Ninin, P.; Valentini, F.; Gonzalez, J.; Salatko-Petryszcze, C.

    2012-01-01

    CERN SSM (Safety System Monitoring) is a system for monitoring state-of-health of the various access and safety systems of the CERN site and accelerator infrastructure. The emphasis of SSM is on the needs of maintenance and system operation with the aim of providing an independent and reliable verification path of the basic operational parameters of each system. Included are all network-connected devices, such as PLCs (local purpose control unit), servers, panel displays, operator posts, etc. The basic monitoring engine of SSM is a freely available system-monitoring framework Zabbix, on top of which a simplified traffic-light-type web-interface has been built. The web-interface of SSM is designed to be ultra-light to facilitate access from hand-held devices over slow connections. The underlying Zabbix system offers history and notification mechanisms typical of advanced monitoring systems. (authors)

  19. Study on the evaluation system for the coal safety management based on risk pre-control

    Institute of Scientific and Technical Information of China (English)

    LI Xin-chun; XU Hai-xia; WANG Pei; SONG Xue-feng

    2009-01-01

    The new type of risk management is process management.First,the hazard sources are identified before coal mine accidents occur,and then the pre-control measure and information monitoring method based on classifying the hidden hazard sources are given.Lastly,the risk pre-alarm and risk control method are confirmed,the management standard and management measure are used to eliminate the hidden hazard sources.In this study,an evaluation system is built to evaluate the result of risk management.

  20. The LHC personnel safety system

    International Nuclear Information System (INIS)

    Ninin, P.; Valentini, F.; Ladzinski, T.

    2011-01-01

    Large particle physics installations such as the CERN Large Hadron Collider require specific Personnel Safety Systems (PSS) to protect the personnel against the radiological and industrial hazards. In order to fulfill the French regulation in matter of nuclear installations, the principles of IEC 61508 and IEC 61513 standard are used as a methodology framework to evaluate the criticality of the installation, to design and to implement the PSS.The LHC PSS deals with the implementation of all physical barriers, access controls and interlock devices around the 27 km of underground tunnel, service zones and experimental caverns of the LHC. The system shall guarantee the absence of personnel in the LHC controlled areas during the machine operations and, on the other hand, ensure the automatic accelerator shutdown in case of any safety condition violation, such as an intrusion during beam circulation. The LHC PSS has been conceived as two separate and independent systems: the LHC Access Control System (LACS) and the LHC Access Safety System (LASS). The LACS, using off the shelf technologies, realizes all physical barriers and regulates all accesses to the underground areas by identifying users and checking their authorizations.The LASS has been designed according to the principles of the IEC 61508 and 61513 standards, starting from a risk analysis conducted on the LHC facility equipped with a standard access control system. It consists in a set of safety functions realized by a dedicated fail-safe and redundant hardware guaranteed to be of SIL3 class. The integration of various technologies combining electronics, sensors, video and operational procedures adopted to establish an efficient personnel safety system for the CERN LHC accelerator is presented in this paper. (authors)

  1. 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)

  2. Systems Safety and Engineering Division

    Data.gov (United States)

    Federal Laboratory Consortium — Volpe's Systems Safety and Engineering Division conducts engineering, research, and analysis to improve transportation safety, capacity, and resiliency. We provide...

  3. The Daresbury personnel safety system

    International Nuclear Information System (INIS)

    Poole, D.E.; Ring, T.

    1989-01-01

    The personnel safety system designed for the SRS at Daresbury is a unified system covering the three accelerators of the source itself, the beamlines and the experimental stations. The system has also been applied to the experimental areas of the Nuclear Structure Facility, and is therefore established as a site standard. A dual guardline interlock module forms a building block for a relay based interlock system completely independent of the machine control system, although comprehensive monitoring of the system status via the control system computer is a feature. An outline of the design criteria adopted for the system is presented together with a more detailed description of the philosophy of the guardline logic and the way this is implemented in a standard modular form. The emphasis is on the design features of a modern microprocessor based variant of the original SRS system. Experience with the original system during build-up and operation of the SRS facility is described. 2 refs., 4 figs

  4. INMACS: Operating experience of a mature, computer-assisted control system for nuclear material inventory and criticality safety

    International Nuclear Information System (INIS)

    Ross, A.M.

    1983-01-01

    This paper describes the operating experience of INMACS, the Integrated Nuclear Material Accounting and Control System used in the Recycle Fuel Fabrication Laboratories at Chalk River. Since commissioning was completed in 1977, INMACS has checked and recorded approximately 3000 inventory-related transactions involved in fabricating thermal-recycle fuels of (U,Pu)0 2 and (Th,Pu)0 2 . No changes have been necessary to INMACS programs that are used by laboratory staff when moving or processing nuclear material. The various utility programs have allowed efficient management and surveillance of the INMACS data base. Hardware failures and the nuisance of system unavailability at the laboratory terminals have been minimized by regular preventative maintenance. The original efforts in the design and rigorous testing of programs have helped INMACS to be accepted enthusiastically by old and new staff of the laboratories. The work required for nuclear material inventory control is done efficiently and in an atmosphere of safety

  5. Safety Verification for Probabilistic Hybrid Systems

    DEFF Research Database (Denmark)

    Zhang, Lijun; She, Zhikun; Ratschan, Stefan

    2010-01-01

    The interplay of random phenomena and continuous real-time control deserves increased attention for instance in wireless sensing and control applications. Safety verification for such systems thus needs to consider probabilistic variations of systems with hybrid dynamics. In safety verification o...... on a number of case studies, tackled using a prototypical implementation....

  6. Seismic simulation and functional performance evaluation of a safety related, seismic category I control room emergency air cleaning system

    International Nuclear Information System (INIS)

    Manley, D.K.; Porco, R.D.; Choi, S.H.

    1985-01-01

    Under a nuclear contract MSA was required to design, manufacture, seismically test and functionally test a complete Safety Related, Seismic Category I, Control Room Emergency Air Cleaning System before shipment to the Yankee Atomic Electric Company, Yankee Nuclear Station in Rowe, Massachusetts. The installation of this system was required to satisfy the NRC requirements of NUREG-0737, Section III, D.3.4, ''Control Room Habitability''. The filter system tested was approximately 3 ft. wide by 8 ft. high by 18 ft. long and weighed an estimated 8300 pounds. It had a design flow rate of 3000 SCFM and contained four stages of filtration - prefilters, upstream and downstream HEPA filters and Type II sideload charcoal adsorber cells. The filter train design followed the guidelines set forth by ANSI/ASME N509-1980. Seismic Category I Qualification Testing consisted of resonance search testing and triaxial random multifrequency testing. In addition to ANSI/ASME N510-1980 testing, triaxial response accelerometers were placed at specific locations on designated prefilters, HEPA filters, charcoal adsorbers and test canisters along with accelerometers at the corresponding filter seal face locations. The purpose of this test was to demonstrate the integrity of the filters, filter seals, and monitor seismic response levels which is directly related to the system's ability to function during a seismic occurrence. The Control Room Emergency Air Cleaning System demonstrated the ability to withstand the maximum postulated earthquake for the plant site by remaining structurally sound and functional

  7. 76 FR 55829 - Federal Motor Vehicle Safety Standards; Electronic Stability Control Systems

    Science.gov (United States)

    2011-09-09

    ... April 2007 final rule described NHTSA's intent to begin formal work to develop a global technical... specifies two sizes of outriggers. The Alliance noted that European and Asian markets have a larger... requirements of the Vehicle Safety Act, the Regulatory Flexibility Act, Executive Order 13132 (Federalism...

  8. Safety of huge systems

    International Nuclear Information System (INIS)

    Kondo, Jiro.

    1995-01-01

    Recently accompanying the development of engineering technology, huge systems tend to be constructed. The disaster countermeasures of huge cities become large problems as the concentration of population into cities is conspicuous. To make the expected value of loss small, the knowledge of reliability engineering is applied. In reliability engineering, even if a part of structures fails, the safety as a whole system must be ensured, therefore, the design having margin is carried out. The degree of margin is called redundancy. However, such design concept makes the structure of a system complex, and as the structure is complex, the possibility of causing human errors becomes high. At the time of huge system design, the concept of fail-safe is effective, but simple design must be kept in mind. The accident in Mihama No. 2 plant of Kansai Electric Power Co. and the accident in Chernobyl nuclear power station, and the accident of Boeing B737 airliner and the fatigue breakdown are described. The importance of safety culture was emphasized as the method of preventing human errors. Man-system interface and management system are discussed. (K.I.)

  9. Final Technical Report on Quantifying Dependability Attributes of Software Based Safety Critical Instrumentation and Control Systems in Nuclear Power Plants

    International Nuclear Information System (INIS)

    Smidts, Carol; Huang, Fuqun; Li, Boyuan; Li, Xiang

    2016-01-01

    With the current transition from analog to digital instrumentation and control systems in nuclear power plants, the number and variety of software-based systems have significantly increased. The sophisticated nature and increasing complexity of software raises trust in these systems as a significant challenge. The trust placed in a software system is typically termed software dependability. Software dependability analysis faces uncommon challenges since software systems' characteristics differ from those of hardware systems. The lack of systematic science-based methods for quantifying the dependability attributes in software-based instrumentation as well as control systems in safety critical applications has proved itself to be a significant inhibitor to the expanded use of modern digital technology in the nuclear industry. Dependability refers to the ability of a system to deliver a service that can be trusted. Dependability is commonly considered as a general concept that encompasses different attributes, e.g., reliability, safety, security, availability and maintainability. Dependability research has progressed significantly over the last few decades. For example, various assessment models and/or design approaches have been proposed for software reliability, software availability and software maintainability. Advances have also been made to integrate multiple dependability attributes, e.g., integrating security with other dependability attributes, measuring availability and maintainability, modeling reliability and availability, quantifying reliability and security, exploring the dependencies between security and safety and developing integrated analysis models. However, there is still a lack of understanding of the dependencies between various dependability attributes as a whole and of how such dependencies are formed. To address the need for quantification and give a more objective basis to the review process -- therefore reducing regulatory uncertainty

  10. Intradermal influenza vaccination of healthy adults using a new microinjection system: a 3-year randomised controlled safety and immunogenicity trial

    Directory of Open Access Journals (Sweden)

    Beran Jiri

    2009-04-01

    Full Text Available Abstract Background Intradermal vaccination provides direct and potentially more efficient access to the immune system via specialised dendritic cells and draining lymphatic vessels. We investigated the immunogenicity and safety during 3 successive years of different dosages of a trivalent, inactivated, split-virion vaccine against seasonal influenza given intradermally using a microinjection system compared with an intramuscular control vaccine. Methods In a randomised, partially blinded, controlled study, healthy volunteers (1150 aged 18 to 57 years at enrolment received three annual vaccinations of intradermal or intramuscular vaccine. In Year 1, subjects were randomised to one of three groups: 3 μg or 6 μg haemagglutinin/strain/dose of inactivated influenza vaccine intradermally, or a licensed inactivated influenza vaccine intramuscularly containing 15 μg/strain/dose. In Year 2 subjects were randomised again to one of two groups: 9 μg/strain/dose intradermally or 15 μg intramuscularly. In Year 3 subjects were randomised a third time to one of two groups: 9 μg intradermally or 15 μg intramuscularly. Randomisation lists in Year 1 were stratified for site. Randomisation lists in Years 2 and 3 were stratified for site and by vaccine received in previous years to ensure the inclusion of a comparable number of subjects in a vaccine group at each centre each year. Immunogenicity was assessed 21 days after each vaccination. Safety was assessed throughout the study. Results In Years 2 and 3, 9 μg intradermal was comparably immunogenic to 15 μg intramuscular for all strains, and both vaccines met European requirements for annual licensing of influenza vaccines. The 3 μg and 6 μg intradermal formulations were less immunogenic than intramuscular 15 μg. Safety of the intradermal and intramuscular vaccinations was comparable in each year of the study. Injection site erythema and swelling was more common with the intradermal route. Conclusion

  11. Technical self reliance of digital safety systems

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Kee Choon; Lee, Dong Young [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Kook Hun [Doosan Heavy Industries and Construction, Changwon (Korea, Republic of); Choi, Seung Gap [POSCON, Pohang (Korea, Republic of)

    2009-04-15

    This paper summarizes the development results of the Korea Nuclear Instrumentation and Control System (KNICS) project sponsored by the Korean government. In this project, Man Machine Interface System (MMIS) architecture, two digital platforms, and several control systems are developed. One platform is a programmable Logic Controller (PLC) for a safety system and another platform is a Distributed Control System (DCS) for a non safety system. With the POSAFE Q PLC, a Reactor Protection System (RPS) and an Engineered Safety Feature Component Control System (ESF CCS) are developed. A Power Control System (PCS) is developed based on the DCS. The safety grade platform and the digital safety systems obtained approval for the Topical Report from the Korean regulatory body in February of 2009. Also a Korean utility and a vendor company determined KNICS results to apply them to the planned Nuclear Power Plant (NPP) in March 2009. This paper introduces the technical self reliance experiences of the safety grade platform and the digital safety systems developed in the KNICS R and D project.

  12. Final Technical Report on Quantifying Dependability Attributes of Software Based Safety Critical Instrumentation and Control Systems in Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Smidts, Carol [The Ohio State Univ., Columbus, OH (United States); Huang, Funqun [The Ohio State Univ., Columbus, OH (United States); Li, Boyuan [The Ohio State Univ., Columbus, OH (United States); Li, Xiang [The Ohio State Univ., Columbus, OH (United States)

    2016-03-25

    With the current transition from analog to digital instrumentation and control systems in nuclear power plants, the number and variety of software-based systems have significantly increased. The sophisticated nature and increasing complexity of software raises trust in these systems as a significant challenge. The trust placed in a software system is typically termed software dependability. Software dependability analysis faces uncommon challenges since software systems’ characteristics differ from those of hardware systems. The lack of systematic science-based methods for quantifying the dependability attributes in software-based instrumentation as well as control systems in safety critical applications has proved itself to be a significant inhibitor to the expanded use of modern digital technology in the nuclear industry. Dependability refers to the ability of a system to deliver a service that can be trusted. Dependability is commonly considered as a general concept that encompasses different attributes, e.g., reliability, safety, security, availability and maintainability. Dependability research has progressed significantly over the last few decades. For example, various assessment models and/or design approaches have been proposed for software reliability, software availability and software maintainability. Advances have also been made to integrate multiple dependability attributes, e.g., integrating security with other dependability attributes, measuring availability and maintainability, modeling reliability and availability, quantifying reliability and security, exploring the dependencies between security and safety and developing integrated analysis models. However, there is still a lack of understanding of the dependencies between various dependability attributes as a whole and of how such dependencies are formed. To address the need for quantification and give a more objective basis to the review process -- therefore reducing regulatory uncertainty

  13. Metal food packaging design based on hazard analysis critical control point (HACCP system in canned food safety

    Directory of Open Access Journals (Sweden)

    Li Xingyi

    2016-06-01

    Full Text Available This study aims to design metal food packaging with hazard analysis critical control point (HACCP. First, theory of HACCP was introduced in detail. Taking empty cans provided by Wuxi Huapeng Food Packaging Company as an example, we studied migration of bisphenol compounds in coating of food can to food stimulant. Moreover, packaging design of luncheon meat can was taken as an example to confirm whether HACCP system could effectively control migration of phenolic substance. Results demonstrated that, coating of such empty were more likely to contain multiple bisphenol compounds such as bisphenol A (BPA, and bisphenol A diglycidyl ether (BADGE was considered as the leading bisphenol pollutant; food stimulant of different types, storage temperature and time could all impact migration of bisphenol compounds. HACCP system was proved to be effective in controlling hazards of phenolic substance in luncheon meat can and could reduce various phenolic substance indexes to an acceptable range. Therefore, HACCP can control migration of phenolic substance and recontamination of food and thus ensure food safety.

  14. The safety interlocking system at the NAC

    International Nuclear Information System (INIS)

    Visser, K.; Mostert, H.

    1984-01-01

    The central safety interlocking system (CSIS) controls the higher level of interlocking between the various cyclotron subsystems. It ensures the safe operation of the entire cyclotron facility as regards personnel safety and proper instrument operation. The system consists of a micro-processor with a ROM-based safety interlocking program, relay output modules providing ''safety OK'' instructions to all interlocked apparatus, alarm input modules connected to transducers providing binary alarm status signals and an interface to the central control computer. All solid state electronic components of the system are situated in a low level radiation area and are interfaced to cyclotron equipment by means of 24 V relays

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  16. A comparison of the difference of requirements between functional safety and nuclear safety controllers

    Energy Technology Data Exchange (ETDEWEB)

    Chen, C.K.; Lee, C.L.; Shyu, S.S. [Inst. of Nuclear Energy Research, Taoyuan, Taiwan (China)

    2014-07-01

    In order to establish self-reliant capabilities of nuclear I&C systems in Taiwan, Taiwan's Nuclear I&C System (TNICS) project had been established by Institute of Nuclear Energy Research (INER). A Triple Modular Redundant (TMR) safety controller (SCS-2000) has been completed and gone through the IEC 61508 Safety Integrity Level 3 (SIL3) certification of Functional Safety for industries. Based on the certification processes, the difference of requirements between Functional Safety and Nuclear Safety controllers in term of hardware and software are addressed in this study. Besides, the measures used to determine and verify the reliability of the safety control system design are presented. (author)

  17. Automated safety control by video cameras

    NARCIS (Netherlands)

    Lefter, I.; Rothkrantz, L.; Somhorst, M.

    2012-01-01

    At this moment many surveillance systems are installed in public domains to control the safety of people and properties. They are constantly watched by human operators who are easily overloaded. To support the human operators, a surveillance system model is designed that detects suspicious behaviour

  18. Environmental Health Impacts of Nuclear Fuel Cycle With Emphasis to Monitoring and Radiological Safety Control System

    International Nuclear Information System (INIS)

    Gad Allah, A.A.; El- Shanshory, A.I.

    2010-01-01

    facilities, as well as their health impacts instruments and monitors systems for radiological control have been reviewed and evaluated

  19. Application range affected by software failures in safety relevant instrumentation and control systems of nuclear power plants

    International Nuclear Information System (INIS)

    Jopen, Manuela; Mbonjo, Herve; Sommer, Dagmar; Ulrich, Birte

    2017-03-01

    This report presents results that have been developed within a BMUB-funded research project (Promotion Code 3614R01304). The overall objective of this project was to broaden the knowledge base of GRS regarding software failures and their impact in software-based instrumentation and control (I and C) systems. To this end, relevant definitions and terms in standards and publications (DIN, IEEE standards, IAEA standards, NUREG publications) as well as in the German safety requirements for nuclear power plants were analyzed first. In particular, it was found that the term ''software fault'' is defined differently and partly contradictory in the considered literature sources. For this reason, a definition of software fault was developed on the basis of the software life cycle of software-based I and C systems within the framework of this project, which takes into account the various aspects relevant to software faults and their related effects. It turns out that software failures result from latent faults in a software-based control system, which can lead to a non-compliant behavior of a software-based I and C system. Hereby a distinction should be made between programming faults and specification faults. In a further step, operational experience with software failures in software-based I and C systems in nuclear facilities and in nonnuclear sector was investigated. The identified events were analyzed with regard to their cause and impacts and the analysis results were summarized. Based on the developed definition of software failure and on the COMPSIS-classification scheme for events related to software based I and C systems, the COCS-classification scheme was developed to classify events from operating experience with software failures, in which the events are classified according to the criteria ''cause'', ''affected system'', ''impact'' and ''CCF potential''. This classification scheme was applied to evaluate the events identified in the framework of this project

  20. Information management system for the control of the data of the safety and radiological protection on a national scale

    International Nuclear Information System (INIS)

    Valdes Ramos, Maryzury; Prendes Alonso, Miguel; Arnau Fernadez, Alma

    2005-01-01

    The Center for Radiation Protection and Hygiene (CPHR) and the National Center for Nuclear Safety (CNSN), have been working in the last years in the design and improvement of a computing tool that allows the management of all the important information, which should be controlled by the Regulatory Authority. The results obtained with the design and implementation of the Integrated System of Data (RASSYN) for the management of the National Regulatory Authority's information in the country are shown in this paper. The software allows an efficient management of the information related to several regulatory aspects such as: the radiation sources in the national territory; the practices associated to the sources; the personnel associated to the practices and their doses; the instruments for the measurement; the waste management; the radiological events; the conditions and requirements of the given authorizations and the inspections results

  1. Development of the safety regulation technology for digital instrumentation and control systems

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Chul Hwan; Ahn, Sang Phil; Yuh, Sang Min; Kang, Yoon Sik; Ahn, Sang Min; Kim, In Su; Kim, Dong Su; Kang, Jin Su [Sungkyunkwan Univ., Seoul (Korea, Republic of)

    2001-03-15

    A general nuclear power stations desire to continue operating for as long as possible. But, cable in nuclear power stations can be subjected to energetic radiation to various degrees as well as other environmental stimuli. It is of major importance to know how these materials will respond and how any resulting aging phenomena can be monitored. From the point of view of nuclear safety the interest in these cables derives from the fact that they are of vital importance. Many researches on cable condition monitoring have been conducting for many years and various techniques of condition monitoring have been suggested until now. We will develop the combination of CM methods using merits of preliminary methods and CM methods using signal processing technique.

  2. Development of the safety regulation technology for digital instrumentation and control systems

    International Nuclear Information System (INIS)

    Kim, Chul Hwan; Ahn, Sang Phil; Yuh, Sang Min; Kang, Yoon Sik; Ahn, Sang Min; Kim, In Su; Kim, Dong Su; Kang, Jin Su

    2001-03-01

    A general nuclear power stations desire to continue operating for as long as possible. But, cable in nuclear power stations can be subjected to energetic radiation to various degrees as well as other environmental stimuli. It is of major importance to know how these materials will respond and how any resulting aging phenomena can be monitored. From the point of view of nuclear safety the interest in these cables derives from the fact that they are of vital importance. Many researches on cable condition monitoring have been conducting for many years and various techniques of condition monitoring have been suggested until now. We will develop the combination of CM methods using merits of preliminary methods and CM methods using signal processing technique

  3. Safety significance evaluation system

    International Nuclear Information System (INIS)

    Lew, B.S.; Yee, D.; Brewer, W.K.; Quattro, P.J.; Kirby, K.D.

    1991-01-01

    This paper reports that the Pacific Gas and Electric Company (PG and E), in cooperation with ABZ, Incorporated and Science Applications International Corporation (SAIC), investigated the use of artificial intelligence-based programming techniques to assist utility personnel in regulatory compliance problems. The result of this investigation is that artificial intelligence-based programming techniques can successfully be applied to this problem. To demonstrate this, a general methodology was developed and several prototype systems based on this methodology were developed. The prototypes address U.S. Nuclear Regulatory Commission (NRC) event reportability requirements, technical specification compliance based on plant equipment status, and quality assurance assistance. This collection of prototype modules is named the safety significance evaluation system

  4. Probabilistic safety assessment and optimal control of hazardous technological systems. A marked point process approach

    International Nuclear Information System (INIS)

    Holmberg, J.

    1997-04-01

    The thesis models risk management as an optimal control problem for a stochastic process. The approach classes the decisions made by management into three categories according to the control methods of a point process: (1) planned process lifetime, (2) modification of the design, and (3) operational decisions. The approach is used for optimization of plant shutdown criteria and surveillance test strategies of a hypothetical nuclear power plant

  5. Probabilistic safety assessment and optimal control of hazardous technological systems. A marked point process approach

    Energy Technology Data Exchange (ETDEWEB)

    Holmberg, J [VTT Automation, Espoo (Finland)

    1997-04-01

    The thesis models risk management as an optimal control problem for a stochastic process. The approach classes the decisions made by management into three categories according to the control methods of a point process: (1) planned process lifetime, (2) modification of the design, and (3) operational decisions. The approach is used for optimization of plant shutdown criteria and surveillance test strategies of a hypothetical nuclear power plant. 62 refs. The thesis includes also five previous publications by author.

  6. Evaluating safety management system implementation

    International Nuclear Information System (INIS)

    Preuss, M.

    2009-01-01

    Canada is committed to not only maintaining, but also improving upon our record of having one of the safest aviation systems in the world. The development, implementation and maintenance of safety management systems is a significant step towards improving safety performance. Canada is considered a world leader in this area and we are fully engaged in implementation. By integrating risk management systems and business practices, the aviation industry stands to gain better safety performance with less regulatory intervention. These are important steps towards improving safety and enhancing the public's confidence in the safety of Canada's aviation system. (author)

  7. Practical Applications of Cosmic Ray Science: Spacecraft, Aircraft, Ground-Based Computation and Control Systems, and Human Health and Safety

    Science.gov (United States)

    Atwell, William; Koontz, Steve; Normand, Eugene

    2012-01-01

    Three twentieth century technological developments, 1) high altitude commercial and military aircraft; 2) manned and unmanned spacecraft; and 3) increasingly complex and sensitive solid state micro-electronics systems, have driven an ongoing evolution of basic cosmic ray science into a set of practical engineering tools needed to design, test, and verify the safety and reliability of modern complex technological systems. The effects of primary cosmic ray particles and secondary particle showers produced by nuclear reactions with the atmosphere, can determine the design and verification processes (as well as the total dollar cost) for manned and unmanned spacecraft avionics systems. Similar considerations apply to commercial and military aircraft operating at high latitudes and altitudes near the atmospheric Pfotzer maximum. Even ground based computational and controls systems can be negatively affected by secondary particle showers at the Earth s surface, especially if the net target area of the sensitive electronic system components is large. Finally, accumulation of both primary cosmic ray and secondary cosmic ray induced particle shower radiation dose is an important health and safety consideration for commercial or military air crews operating at high altitude/latitude and is also one of the most important factors presently limiting manned space flight operations beyond low-Earth orbit (LEO). In this paper we review the discovery of cosmic ray effects on the performance and reliability of microelectronic systems as well as human health and the development of the engineering and health science tools used to evaluate and mitigate cosmic ray effects in ground-based atmospheric flight, and space flight environments. Ground test methods applied to microelectronic components and systems are used in combinations with radiation transport and reaction codes to predict the performance of microelectronic systems in their operating environments. Similar radiation transport

  8. AEC controlled area safety program

    International Nuclear Information System (INIS)

    Hendricks, D.W.

    1969-01-01

    The detonation of underground nuclear explosives and the subsequent data recovery efforts require a comprehensive pre- and post-detonation safety program for workers within the controlled area. The general personnel monitoring and environmental surveillance program at the Nevada Test Site are presented. Some of the more unusual health-physics aspects involved in the operation of this program are also discussed. The application of experience gained at the Nevada Test Site is illustrated by description of the on-site operational and safety programs established for Project Gasbuggy. (author)

  9. AEC controlled area safety program

    Energy Technology Data Exchange (ETDEWEB)

    Hendricks, D W [Nevada Operations Office, Atomic Energy Commission, Las Vegas, NV (United States)

    1969-07-01

    The detonation of underground nuclear explosives and the subsequent data recovery efforts require a comprehensive pre- and post-detonation safety program for workers within the controlled area. The general personnel monitoring and environmental surveillance program at the Nevada Test Site are presented. Some of the more unusual health-physics aspects involved in the operation of this program are also discussed. The application of experience gained at the Nevada Test Site is illustrated by description of the on-site operational and safety programs established for Project Gasbuggy. (author)

  10. Design of the Magnum-PSI safety, control and data acquisition system

    NARCIS (Netherlands)

    van der Linden, G. W.; Wijnoltz, F.; Scholten, J.; Busch, P. J.; Poelman, A. J.; Smeets, P. H.; de Groot, B.; Koppers, W. R.

    2008-01-01

    The FOM-Institute for Plasma Physics Rijnhuizen has started the construction of Magnum-PSI, a magnetized (3 T), steady-state, large area (80cm(2)) high-flux (up to 10(24) H(+)ions m(-2) s(-1)) plasma generator. The aim of this linear plasma device is to provide a controlled, highly accessible

  11. 46 CFR 62.35-50 - Tabulated monitoring and safety control requirements for specific systems.

    Science.gov (United States)

    2010-10-01

    ... Status Main (Propulsion steam) turbine (2) (2) (2) (4, 5) Manual trip Main propulsion, diesel (1) (1) (1... lubrication Pressure Low Main propulsion, controllable pitch propeller Hydraulic oil Pressure High, Low...) Gas turbine (8) (8) (8) (8) (5) Engines and turbines Jacking/turning gear Engaged (8) Fuel oil (9) (9...

  12. The ATLAS Detector Safety System

    CERN Multimedia

    Helfried Burckhart; Kathy Pommes; Heidi Sandaker

    The ATLAS Detector Safety System (DSS) has the mandate to put the detector in a safe state in case an abnormal situation arises which could be potentially dangerous for the detector. It covers the CERN alarm severity levels 1 and 2, which address serious risks for the equipment. The highest level 3, which also includes danger for persons, is the responsibility of the CERN-wide system CSAM, which always triggers an intervention by the CERN fire brigade. DSS works independently from and hence complements the Detector Control System, which is the tool to operate the experiment. The DSS is organized in a Front- End (FE), which fulfills autonomously the safety functions and a Back-End (BE) for interaction and configuration. The overall layout is shown in the picture below. ATLAS DSS configuration The FE implementation is based on a redundant Programmable Logical Crate (PLC) system which is used also in industry for such safety applications. Each of the two PLCs alone, one located underground and one at the s...

  13. A study on the revision of nuclear safety act to build the foundation of nuclear export and import control system in Korea

    International Nuclear Information System (INIS)

    Yang, Seung Hyo; Choi, Sun Do

    2012-01-01

    Nuclear related items require export and import control beyond the multilateral export control system according to Safeguard Agreement, Additional Protocol and bilateral agreements. Besides Korea as a nuclear supplier is needed to actively cope with its export control system, which is being reinforced internationally. In regard to this trend, this study drew the revision plan of present Nuclear Safety Act to found the nuclear export and import control system in Korea by examining the related legislations and analyzing the implementation status of nuclear export and import control

  14. A study on the revision of nuclear safety act to build the foundation of nuclear export and import control system in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Seung Hyo; Choi, Sun Do [Korea Institute of Nuclear Nonproliferation and Control, Daejeon (Korea, Republic of)

    2012-10-15

    Nuclear related items require export and import control beyond the multilateral export control system according to Safeguard Agreement, Additional Protocol and bilateral agreements. Besides Korea as a nuclear supplier is needed to actively cope with its export control system, which is being reinforced internationally. In regard to this trend, this study drew the revision plan of present Nuclear Safety Act to found the nuclear export and import control system in Korea by examining the related legislations and analyzing the implementation status of nuclear export and import control.

  15. Software Quality Assurance for Nuclear Safety Systems

    International Nuclear Information System (INIS)

    Sparkman, D R; Lagdon, R

    2004-01-01

    The US Department of Energy has undertaken an initiative to improve the quality of software used to design and operate their nuclear facilities across the United States. One aspect of this initiative is to revise or create new directives and guides associated with quality practices for the safety software in its nuclear facilities. Safety software includes the safety structures, systems, and components software and firmware, support software and design and analysis software used to ensure the safety of the facility. DOE nuclear facilities are unique when compared to commercial nuclear or other industrial activities in terms of the types and quantities of hazards that must be controlled to protect workers, public and the environment. Because of these differences, DOE must develop an approach to software quality assurance that ensures appropriate risk mitigation by developing a framework of requirements that accomplishes the following goals: (sm b ullet) Ensures the software processes developed to address nuclear safety in design, operation, construction and maintenance of its facilities are safe (sm b ullet) Considers the larger system that uses the software and its impacts (sm b ullet) Ensures that the software failures do not create unsafe conditions Software designers for nuclear systems and processes must reduce risks in software applications by incorporating processes that recognize, detect, and mitigate software failure in safety related systems. It must also ensure that fail safe modes and component testing are incorporated into software design. For nuclear facilities, the consideration of risk is not necessarily sufficient to ensure safety. Systematic evaluation, independent verification and system safety analysis must be considered for software design, implementation, and operation. The software industry primarily uses risk analysis to determine the appropriate level of rigor applied to software practices. This risk-based approach distinguishes safety

  16. The aviation safety reporting system

    Science.gov (United States)

    Reynard, W. D.

    1984-01-01

    The aviation safety reporting system, an accident reporting system, is presented. The system identifies deficiencies and discrepancies and the data it provides are used for long term identification of problems. Data for planning and policy making are provided. The system offers training in safety education to pilots. Data and information are drawn from the available data bases.

  17. Integrated therapy safety management system.

    Science.gov (United States)

    Podtschaske, Beatrice; Fuchs, Daniela; Friesdorf, Wolfgang

    2013-09-01

    The aim is to demonstrate the benefit of the medico-ergonomic approach for the redesign of clinical work systems. Based on the six layer model, a concept for an 'integrated therapy safety management' is drafted. This concept could serve as a basis to improve resilience. The concept is developed through a concept-based approach. The state of the art of safety and complexity research in human factors and ergonomics forms the basis. The findings are synthesized to a concept for 'integrated therapy safety management'. The concept is applied by way of example for the 'medication process' to demonstrate its practical implementation. The 'integrated therapy safety management' is drafted in accordance with the six layer model. This model supports a detailed description of specific work tasks, the corresponding responsibilities and related workflows at different layers by using the concept of 'bridge managers'. 'Bridge managers' anticipate potential errors and monitor the controlled system continuously. If disruptions or disturbances occur, they respond with corrective actions which ensure that no harm results and they initiate preventive measures for future procedures. The concept demonstrates that in a complex work system, the human factor is the key element and final authority to cope with the residual complexity. The expertise of the 'bridge managers' and the recursive hierarchical structure results in highly adaptive clinical work systems and increases their resilience. The medico-ergonomic approach is a highly promising way of coping with two complexities. It offers a systematic framework for comprehensive analyses of clinical work systems and promotes interdisciplinary collaboration. © 2013 The Authors. British Journal of Clinical Pharmacology © 2013 The British Pharmacological Society.

  18. Integrated therapy safety management system

    Science.gov (United States)

    Podtschaske, Beatrice; Fuchs, Daniela; Friesdorf, Wolfgang

    2013-01-01

    Aims The aim is to demonstrate the benefit of the medico-ergonomic approach for the redesign of clinical work systems. Based on the six layer model, a concept for an ‘integrated therapy safety management’ is drafted. This concept could serve as a basis to improve resilience. Methods The concept is developed through a concept-based approach. The state of the art of safety and complexity research in human factors and ergonomics forms the basis. The findings are synthesized to a concept for ‘integrated therapy safety management’. The concept is applied by way of example for the ‘medication process’ to demonstrate its practical implementation. Results The ‘integrated therapy safety management’ is drafted in accordance with the six layer model. This model supports a detailed description of specific work tasks, the corresponding responsibilities and related workflows at different layers by using the concept of ‘bridge managers’. ‘Bridge managers’ anticipate potential errors and monitor the controlled system continuously. If disruptions or disturbances occur, they respond with corrective actions which ensure that no harm results and they initiate preventive measures for future procedures. The concept demonstrates that in a complex work system, the human factor is the key element and final authority to cope with the residual complexity. The expertise of the ‘bridge managers’ and the recursive hierarchical structure results in highly adaptive clinical work systems and increases their resilience. Conclusions The medico-ergonomic approach is a highly promising way of coping with two complexities. It offers a systematic framework for comprehensive analyses of clinical work systems and promotes interdisciplinary collaboration. PMID:24007448

  19. Test to prove the resistance to incidents of components of electric and control systems in the safety containment of nuclear power plants

    International Nuclear Information System (INIS)

    1982-01-01

    The marginal program for proving the suitability of safety-relevant components of electric and control systems in the safety containment during a loss-of-coolant incident is described. Variant test conditions are established in the component-specific test program. Special attention has been paid to the representation of the course of pressure and temperature for the performance test of the valve room of the Nuclear Power Plant Philippsburg 2. (DG) [de

  20. Analysis and design on airport safety information management system

    Directory of Open Access Journals (Sweden)

    Yan Lin

    2017-01-01

    Full Text Available Airport safety information management system is the foundation of implementing safety operation, risk control, safety performance monitor, and safety management decision for the airport. The paper puts forward the architecture of airport safety information management system based on B/S model, focuses on safety information processing flow, designs the functional modules and proposes the supporting conditions for system operation. The system construction is helpful to perfecting the long effect mechanism driven by safety information, continually increasing airport safety management level and control proficiency.

  1. Safety and efficacy of subcutaneous tocilizumab in adults with systemic sclerosis (faSScinate) : a phase 2, randomised, controlled trial

    NARCIS (Netherlands)

    Khanna, Dinesh; Denton, Christopher P.; Jahreis, Angelika; van Laar, Jacob M.; Frech, Tracy M.; Anderson, Marina E.; Baron, Murray; Chung, Lorinda; Fierlbeck, Gerhard; Lakshminarayanan, Santhanam; Allanore, Yannick; Pope, Janet E.; Riemekasten, Gabriela; Steen, Virginia; Müller-Ladner, Ulf; Lafyatis, Robert; Stifano, Giuseppina; Spotswood, Helen; Chen-Harris, Haiyin; Dziadek, Sebastian; Morimoto, Alyssa; Sornasse, Thierry; Siegel, Jeffrey; Furst, Daniel E.

    2016-01-01

    Background Systemic sclerosis is a rare disabling autoimmune disease with few treatment options. The efficacy and safety of tocilizumab, an interleukin 6 receptor-α inhibitor, was assessed in the faSScinate phase 2 trial in patients with systemic sclerosis. Methods We did this double-blind,

  2. Intermediate probabilistic safety assessment approach for safety critical digital systems

    International Nuclear Information System (INIS)

    Taeyong, Sung; Hyun Gook, Kang

    2001-01-01

    Even though the conventional probabilistic safety assessment methods are immature for applying to microprocessor-based digital systems, practical needs force to apply it. In the Korea, UCN 5 and 6 units are being constructed and Korean Next Generation Reactor is being designed using the digital instrumentation and control equipment for the safety related functions. Korean regulatory body requires probabilistic safety assessment. This paper analyzes the difficulties on the assessment of digital systems and suggests an intermediate framework for evaluating their safety using fault tree models. The framework deals with several important characteristics of digital systems including software modules and fault-tolerant features. We expect that the analysis result will provide valuable design feedback. (authors)

  3. Regulatory Control of Radiation Sources. Safety Guide

    International Nuclear Information System (INIS)

    2009-01-01

    This Safety Guide is intended to assist States in implementing the requirements established in Safety Standards Series No. GS-R-1, Legal and Governmental Infrastructure for Nuclear, Radiation, Radioactive Waste and Transport Safety, for a national regulatory infrastructure to regulate any practice involving radiation sources in medicine, industry, research, agriculture and education. The Safety Guide provides advice on the legislative basis for establishing regulatory bodies, including the effective independence of the regulatory body. It also provides guidance on implementing the functions and activities of regulatory bodies: the development of regulations and guides on radiation safety; implementation of a system for notification and authorization; carrying out regulatory inspections; taking necessary enforcement actions; and investigating accidents and circumstances potentially giving rise to accidents. The various aspects relating to the regulatory control of consumer products are explained, including justification, optimization of exposure, safety assessment and authorization. Guidance is also provided on the organization and staffing of regulatory bodies. Contents: 1. Introduction; 2. Legal framework for a regulatory infrastructure; 3. Principal functions and activities of the regulatory body; 4. Regulatory control of the supply of consumer products; 5. Functions of the regulatory body shared with other governmental agencies; 6. Organization and staffing of the regulatory body; 7. Documentation of the functions and activities of the regulatory body; 8. Support services; 9. Quality management for the regulatory system.

  4. Reaction Control System Thruster Cracking Consultation: NASA Engineering and Safety Center (NESC) Materials Super Problem Resolution Team (SPRT) Findings

    Science.gov (United States)

    MacKay, Rebecca A.; Smith, Stephen W.; Shah, Sandeep R.; Piascik, Robert S.

    2005-01-01

    The shuttle orbiter s reaction control system (RCS) primary thruster serial number 120 was found to contain cracks in the counter bores and relief radius after a chamber repair and rejuvenation was performed in April 2004. Relief radius cracking had been observed in the 1970s and 1980s in seven thrusters prior to flight; however, counter bore cracking had never been seen previously in RCS thrusters. Members of the Materials Super Problem Resolution Team (SPRT) of the NASA Engineering and Safety Center (NESC) conducted a detailed review of the relevant literature and of the documentation from the previous RCS thruster failure analyses. It was concluded that the previous failure analyses lacked sufficient documentation to support the conclusions that stress corrosion cracking or hot-salt cracking was the root cause of the thruster cracking and lacked reliable inspection controls to prevent cracked thrusters from entering the fleet. The NESC team identified and performed new materials characterization and mechanical tests. It was determined that the thruster intergranular cracking was due to hydrogen embrittlement and that the cracking was produced during manufacturing as a result of processing the thrusters with fluoride-containing acids. Testing and characterization demonstrated that appreciable environmental crack propagation does not occur after manufacturing.

  5. OBTAINING FOOD SAFETY BY APPLYING HACCP SYSTEM

    Directory of Open Access Journals (Sweden)

    ION CRIVEANU

    2012-01-01

    Full Text Available In order to increase the confidence of the trading partners and consumers in the products which are sold on the market, enterprises producing food are required to implement the food safety system HACCP,a particularly useful system because the manufacturer is not able to fully control finished products . SR EN ISO 22000:2005 establishes requirements for a food safety management system where an organization in the food chain needs to proove its ability to control food safety hazards in order to ensure that food is safe at the time of human consumption. This paper presents the main steps which ensure food safety using the HACCP system, and SR EN ISO 20000:2005 requirements for food safety.

  6. Aviation Safety Hotline Information System -

    Data.gov (United States)

    Department of Transportation — The Aviation Safety Hotline Information System (ASHIS) collects, stores, and retrieves reports submitted by pilots, mechanics, cabin crew, passengers, or the public...

  7. Traceability of Software Safety Requirements in Legacy Safety Critical Systems

    Science.gov (United States)

    Hill, Janice L.

    2007-01-01

    How can traceability of software safety requirements be created for legacy safety critical systems? Requirements in safety standards are imposed most times during contract negotiations. On the other hand, there are instances where safety standards are levied on legacy safety critical systems, some of which may be considered for reuse for new applications. Safety standards often specify that software development documentation include process-oriented and technical safety requirements, and also require that system and software safety analyses are performed supporting technical safety requirements implementation. So what can be done if the requisite documents for establishing and maintaining safety requirements traceability are not available?

  8. Technical features of ABWR safety systems

    International Nuclear Information System (INIS)

    Sugisaki, Toshihiko; Tominaga, Kenji; Horiuchi, Tetsuo

    1986-01-01

    The engineering safety facilities of ABWRs have been disigned so as to have many excellent characteristics such as safety, reliability and economy, reflecting the merit of adopting new technology such as internal pumps and new control rod driving mechanism, and coupled with the safety peculiar to BWRs. In this paper, about ECCS, containment vessels and others which compose the engineering safety facilities of ABWRs, the characteristics related to the safety owing to the adoption of internal pumps and others, and the evaluation of the performance at the time of various accidents are discussed. As the results of safety evaluation, it was clarified that due to the safety peculiar to ABWRs and the characteristics of the safety facilities, the large increases of safety, reliability and economy have been planned in the ABWRs, and for example, core flooding can be maintained even at the time of a hypothetical loss of coolant accident. BWRs have the simple system constitution, good self controllability, large natural circulation ability, simple operation control method and excellent ability of confining heat and radioactivity. BWRs have three safety functions to stop reactors, to remove heat from reactors, and to confine radioactive substances. These functions of ABWRs were evaluated, and very high safety was confirmed. (Kako, I.)

  9. Compartmentalized safety coolant injection system

    International Nuclear Information System (INIS)

    Johnson, F.T.

    1983-01-01

    A safety coolant injection system for nuclear reactors wherein a core reflood tank is provided to afford more reliable reflooding of the reactor core in the event of a break in one of the reactor coolant supply loops. Each reactor coolant supply loop is arranged in a separate compartment in the containment structure to contain and control the flow of spilled coolant so as to permit its use during emergency core cooling procedures. A spillway allows spilled coolant in the compartment to pass into the emergency water storage tank from where it can be pumped back to the reactor vessel. (author)

  10. INTEGRATED SAFETY MANAGEMENT SYSTEM IN AIR TRAFFIC SERVICES

    Directory of Open Access Journals (Sweden)

    Volodymyr Kharchenko

    2014-06-01

    Full Text Available The article deals with the analysis of the researches conducted in the field of safety management systems.Safety management system framework, methods and tools for safety analysis in Air Traffic Control have been reviewed.Principles of development of Integrated safety management system in Air Traffic Services have been proposed.

  11. Design of an Active Automotive Safety System

    Directory of Open Access Journals (Sweden)

    Y. Wang

    2013-07-01

    Full Text Available With the development of the national economy, the people's standard of living got corresponding improvement, cars has been one of the indispensable traffic tools in many families. An active safety system is proposed, which can real-time detect the vehicle's running status and judge the security status of the vehicle. The system, which takes single-chip microcomputer as the controlling core and combines with millimeter-wave and ultrasonic distance measurement technology, can detect the distance from vehicle to vehicle and judge the security status of the vehicle. The hardware composition of the system and the data acquiring circuit are proposed, the mathematic model for different situation is established, and the controlling algorithm is completed. This system can accurately measure speed and distance between vehicles; the active safety control system can meet the relevant data measurement and transmission requirement; and can meet the functional requirement of the active safety control system

  12. Strategy to safety grade systems replacements

    International Nuclear Information System (INIS)

    Stimler, M.; Sullivan, K.E.; Trebincevic, I.

    1993-01-01

    The introduction of digital instrumentation and control systems in nuclear power plants is characterized by the need to satisfy the requirements of safety, reliability and man-machine ergonomics. Today digital instrumentation and control systems meet these requirements and the trend in Europe is towards full digital based nuclear power plant control systems. This paper describes Siemens (KWU) experience in nuclear power plants and development in trends within Europe. Topics which are the subject of major concern to NPP operators addressed in this paper are: human performance factors - man-machine interface; operating philosophy; safety, availability and reliability. Other aspects addressed are: Siemens open-quotes defense in depthclose quotes concept, description of Siemens digital I ampersand C systems, safety requirements and systems, I ampersand C qualification, control room ergonomics, information systems and retrofitting experience

  13. Enhancing Safety at Airline Operations Control Centre

    Directory of Open Access Journals (Sweden)

    Lukáš Řasa

    2015-04-01

    Full Text Available In recent years a new term of Safety Management System (SMS has been introduced into aviation legislation. This system is being adopted by airline operators. One of the groundbased actors of everyday operations is Operations Control Centre (OCC. The goal of this article has been to identify and assess risks and dangers which occur at OCC and create a template for OCC implementation into SMS.

  14. System safety engineering analysis handbook

    Science.gov (United States)

    Ijams, T. E.

    1972-01-01

    The basic requirements and guidelines for the preparation of System Safety Engineering Analysis are presented. The philosophy of System Safety and the various analytic methods available to the engineering profession are discussed. A text-book description of each of the methods is included.

  15. Safety analysis of control rod drive computers

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  16. Architecture Level Safety Analyses for Safety-Critical Systems

    Directory of Open Access Journals (Sweden)

    K. S. Kushal

    2017-01-01

    Full Text Available The dependency of complex embedded Safety-Critical Systems across Avionics and Aerospace domains on their underlying software and hardware components has gradually increased with progression in time. Such application domain systems are developed based on a complex integrated architecture, which is modular in nature. Engineering practices assured with system safety standards to manage the failure, faulty, and unsafe operational conditions are very much necessary. System safety analyses involve the analysis of complex software architecture of the system, a major aspect in leading to fatal consequences in the behaviour of Safety-Critical Systems, and provide high reliability and dependability factors during their development. In this paper, we propose an architecture fault modeling and the safety analyses approach that will aid in identifying and eliminating the design flaws. The formal foundations of SAE Architecture Analysis & Design Language (AADL augmented with the Error Model Annex (EMV are discussed. The fault propagation, failure behaviour, and the composite behaviour of the design flaws/failures are considered for architecture safety analysis. The illustration of the proposed approach is validated by implementing the Speed Control Unit of Power-Boat Autopilot (PBA system. The Error Model Annex (EMV is guided with the pattern of consideration and inclusion of probable failure scenarios and propagation of fault conditions in the Speed Control Unit of Power-Boat Autopilot (PBA. This helps in validating the system architecture with the detection of the error event in the model and its impact in the operational environment. This also provides an insight of the certification impact that these exceptional conditions pose at various criticality levels and design assurance levels and its implications in verifying and validating the designs.

  17. Application of the defense-in-depth concept to qualify computer-based instrumentation and control systems important to safety

    International Nuclear Information System (INIS)

    Seidel, F.

    1998-01-01

    In parallel to the technological development, the authorities and expert organisations are preparing the application of computer-based I and C to NPPs from the regulatory point of view. Generally the associated world-wide procedure follows steps like identification of safety issues, completion of the regulatory framework particularly regarding the licensing requirements and furthermore, recommendation of an appropriate set of qualification methods to prove that the requirements are met. The paper's intention is to show from the regulatory point of view that the choice as well as the combination of the qualification methods depend on system design features and development strategy. Similar as for the safety system design required, a defense-in-depth qualification concept is suggested to be helpful in order to prove that the computer-based system meets the licensing requirements. (author)

  18. Application of the defense-in-depth concept to qualify computer-based instrumentation and control systems important to safety

    Energy Technology Data Exchange (ETDEWEB)

    Seidel, F [Federal Office for Radiation Protection, Salzgitter (Germany)

    1998-10-01

    In parallel to the technological development, the authorities and expert organisations are preparing the application of computer-based I and C to NPPs from the regulatory point of view. Generally the associated world-wide procedure follows steps like identification of safety issues, completion of the regulatory framework particularly regarding the licensing requirements and furthermore, recommendation of an appropriate set of qualification methods to prove that the requirements are met. The paper`s intention is to show from the regulatory point of view that the choice as well as the combination of the qualification methods depend on system design features and development strategy. Similar as for the safety system design required, a defense-in-depth qualification concept is suggested to be helpful in order to prove that the computer-based system meets the licensing requirements. (author)

  19. Safety balance: Analysis of safety systems

    International Nuclear Information System (INIS)

    Delage, M.; Giroux, C.

    1990-12-01

    Safety analysis, and particularly analysis of exploitation of NPPs is constantly affected by EDF and by the safety authorities and their methodologies. Periodic safety reports ensure that important issues are not missed on daily basis, that incidents are identified and that relevant actions are undertaken. French safety analysis method consists of three principal steps. First type of safety balance is analyzed at the normal start-up phase for each unit including the final safety report. This enables analysis of behaviour of units ten years after their licensing. Second type is periodic operational safety analysis performed during a few years. Finally, the third step consists of safety analysis of the oldest units with the aim to improve the safety standards. The three steps of safety analysis are described in this presentation in detail with the aim to present the objectives and principles. Examples of most recent exercises are included in order to illustrate the importance of such analyses

  20. Safety and interlock system for Tristan

    International Nuclear Information System (INIS)

    Takeda, S.; Kudo, K.; Katoh, T.; Akiyama, A.

    1987-01-01

    This report describes alarm and interlock system of TRISTAN, concentrating on personnel safety. The basis of TRISTAN machine-control system (TMS) is an N-to-N computer network and KEK NODAL which offers high software productivity. TMC achieves high flexibility of operation both for normal operation and for the fast commissioning. However, to assure the safety of personnel and the TRISTAN machine operation, the safety system has to continue functioning during TMC failure as well. A distributed safety and interlock system (DSIS) is used for diversification of risks in TRISTAN system. DSIS is functionally subdivided along local system lines and has a hierarchical structure of 12 programmable sequence controllers (PSCs). Optical fiber links connect the PSCs at subsystem level and a PSC at the supervisory level of TRISTAN central control room (TCCR). The subsystem PSCs provide the interlock functions between their local devices. The local PSCs interact with the central system through a limited number of summarized signals. The central PSC provides the interlock functions between the subsystems and interacts with an operator's panel. Personnel safety is based on a system of electrical interlock keys, emergency push-buttons around the tunnel, at the entrance gates or in the control room

  1. Safety control and minimization of radioactive wastes

    International Nuclear Information System (INIS)

    Wang Jinming; Rong Feng; Li Jinyan; Wang Xin

    2010-01-01

    Compared with the developed countries, the safety control and minimization of the radwastes in China are under-developed. The research of measures for the safety control and minimization of the radwastes is very important for the safety control of the radwastes, and the reduction of the treatment and disposal cost and environment radiation hazards. This paper has systematically discussed the safety control and the minimization of the radwastes produced in the nuclear fuel circulation, nuclear technology applications and the process of decommission of nuclear facilities, and has provided some measures and methods for the safety control and minimization of the radwastes. (authors)

  2. System safety education focused on system management

    Science.gov (United States)

    Grose, V. L.

    1971-01-01

    System safety is defined and characteristics of the system are outlined. Some of the principle characteristics include role of humans in hazard analysis, clear language for input and output, system interdependence, self containment, and parallel analysis of elements.

  3. Software Safety Risk in Legacy Safety-Critical Computer Systems

    Science.gov (United States)

    Hill, Janice L.; Baggs, Rhoda

    2007-01-01

    Safety Standards contain technical and process-oriented safety requirements. Technical requirements are those such as "must work" and "must not work" functions in the system. Process-Oriented requirements are software engineering and safety management process requirements. Address the system perspective and some cover just software in the system > NASA-STD-8719.13B Software Safety Standard is the current standard of interest. NASA programs/projects will have their own set of safety requirements derived from the standard. Safety Cases: a) Documented demonstration that a system complies with the specified safety requirements. b) Evidence is gathered on the integrity of the system and put forward as an argued case. [Gardener (ed.)] c) Problems occur when trying to meet safety standards, and thus make retrospective safety cases, in legacy safety-critical computer systems.

  4. Contribution to the evaluation of safety of software used in command control systems in nuclear plants: application to the SPIN N4

    International Nuclear Information System (INIS)

    Soubies, B.; Boulc'h, J.; Elsensohn, O.; Le Meur, M.; Henry, J.Y.

    1994-06-01

    The licensing procedures process of nuclear plants features compulsory steps which bring about a thorough exam of the commands control system. This analysis accounts for the aspects linked to technologies (integrated circuits, software packages) which have been chosen by the manufacturer for the programmed systems in charge of safety functions. Important innovations have been introduced in terms of design and manufacturing processes of safety systems of 1400 MWe pressurized water reactors, more precisely for the integrated numerical protection system (SPIN). The methodology used by the IPSN for the exam of the software of this system is presented in the communication. This methodology leads the IPSN to carry out studies and developments of tools keeping in sight as their main goal to bring substantial help to analysis. (authors). 2 refs

  5. Bioprotective agents in safety control

    Directory of Open Access Journals (Sweden)

    Dimitrijević-Branković Suzana I.

    2003-01-01

    Full Text Available Food poisoning is the one of the main health hazards even today. More than 200 known diseases are transmitted through food. The causes of foodborne illness include viruses, bacteria, parasites, toxins, metals, and prions and the symptoms of foodborne illness range from mild gastroenteritis to life-threatening neurological, hepatic and renal syndromes.The prevention of food poisonings represents very serious task for food manufacturers. Beside food control according to the concept "from the farm to the table" there is increased need for the development of new technology for longer shelf lifes of food. Food fermented by lactic acid bacteria (LAB and traditionally considered to be safe. There are many substances produced by LAB that affect the shelf life of fermented food, by active suppression of poisoning microorganisms growth. Because of that, the LAB is recently considered as bioprotective agents that have important role in food safety.

  6. Two viewpoints for software failures and their relation in probabilistic safety assessment of digital instrumentation and control systems

    International Nuclear Information System (INIS)

    Kim, Man Cheol

    2015-01-01

    As the use of digital systems in nuclear power plants increases, the reliability of the software becomes one of the important issues in probabilistic safety assessment. In this paper, two viewpoints for a software failure during the operation of a digital system or a statistical software test are identified, and the relation between them is provided. In conventional software reliability analysis, a failure is mainly viewed with respect to the system operation. A new viewpoint with respect to the system input is suggested. The failure probability density functions for the two viewpoints are defined, and the relation between the two failure probability density functions is derived. Each failure probability density function can be derived from the other failure probability density function by applying the derived relation between the two failure probability density functions. The usefulness of the derived relation is demonstrated by applying it to the failure data obtained from the software testing of a real system. The two viewpoints and their relation, as identified in this paper, are expected to help us extend our understanding of the reliability of safety-critical software. (author)

  7. Technical safety requirements control level verification

    International Nuclear Information System (INIS)

    STEWART, J.L.

    1999-01-01

    A Technical Safety Requirement (TSR) control level verification process was developed for the Tank Waste Remediation System (TWRS) TSRs at the Hanford Site in Richland, WA, at the direction of the US. Department of Energy, Richland Operations Office (RL). The objective of the effort was to develop a process to ensure that the TWRS TSR controls are designated and managed at the appropriate levels as Safety Limits (SLs), Limiting Control Settings (LCSs), Limiting Conditions for Operation (LCOs), Administrative Controls (ACs), or Design Features. The TSR control level verification process was developed and implemented by a team of contractor personnel with the participation of Fluor Daniel Hanford, Inc. (FDH), the Project Hanford Management Contract (PHMC) integrating contractor, and RL representatives. The team was composed of individuals with the following experience base: nuclear safety analysis; licensing; nuclear industry and DOE-complex TSR preparation/review experience; tank farm operations; FDH policy and compliance; and RL-TWRS oversight. Each TSR control level designation was completed utilizing TSR control logic diagrams and TSR criteria checklists based on DOE Orders, Standards, Contractor TSR policy, and other guidance. The control logic diagrams and criteria checklists were reviewed and modified by team members during team meetings. The TSR control level verification process was used to systematically evaluate 12 LCOs, 22 AC programs, and approximately 100 program key elements identified in the TWRS TSR document. The verification of each TSR control required a team consensus. Based on the results of the process, refinements were identified and the TWRS TSRs were modified as appropriate. A final report documenting key assumptions and the control level designation for each TSR control was prepared and is maintained on file for future reference. The results of the process were used as a reference in the RL review of the final TWRS TSRs and control suite. RL

  8. Technical safety requirements control level verification; TOPICAL

    International Nuclear Information System (INIS)

    STEWART, J.L.

    1999-01-01

    A Technical Safety Requirement (TSR) control level verification process was developed for the Tank Waste Remediation System (TWRS) TSRs at the Hanford Site in Richland, WA, at the direction of the US. Department of Energy, Richland Operations Office (RL). The objective of the effort was to develop a process to ensure that the TWRS TSR controls are designated and managed at the appropriate levels as Safety Limits (SLs), Limiting Control Settings (LCSs), Limiting Conditions for Operation (LCOs), Administrative Controls (ACs), or Design Features. The TSR control level verification process was developed and implemented by a team of contractor personnel with the participation of Fluor Daniel Hanford, Inc. (FDH), the Project Hanford Management Contract (PHMC) integrating contractor, and RL representatives. The team was composed of individuals with the following experience base: nuclear safety analysis; licensing; nuclear industry and DOE-complex TSR preparation/review experience; tank farm operations; FDH policy and compliance; and RL-TWRS oversight. Each TSR control level designation was completed utilizing TSR control logic diagrams and TSR criteria checklists based on DOE Orders, Standards, Contractor TSR policy, and other guidance. The control logic diagrams and criteria checklists were reviewed and modified by team members during team meetings. The TSR control level verification process was used to systematically evaluate 12 LCOs, 22 AC programs, and approximately 100 program key elements identified in the TWRS TSR document. The verification of each TSR control required a team consensus. Based on the results of the process, refinements were identified and the TWRS TSRs were modified as appropriate. A final report documenting key assumptions and the control level designation for each TSR control was prepared and is maintained on file for future reference. The results of the process were used as a reference in the RL review of the final TWRS TSRs and control suite. RL

  9. Systematic evaluation program review of NRC safety topic VII-2 associated with the electrical, instrumentation and control portions of the ESF system control logic and design for the Dresden Station, Unit II nuclear power plant

    International Nuclear Information System (INIS)

    St Leger-Barter, G.

    1980-11-01

    This report documents the technical evaluation and review of NRC Safety Topic VII-2, associated with the electrical, instrumentation, and control portions of the ESF system control logic and design for the Dresden Station Unit II nuclear power plant, using current licensing criteria

  10. Aviation Safety Reporting System: Process and Procedures

    Science.gov (United States)

    Connell, Linda J.

    1997-01-01

    The Aviation Safety Reporting System (ASRS) was established in 1976 under an agreement between the Federal Aviation Administration (FAA) and the National Aeronautics and Space Administration (NASA). This cooperative safety program invites pilots, air traffic controllers, flight attendants, maintenance personnel, and others to voluntarily report to NASA any aviation incident or safety hazard. The FAA provides most of the program funding. NASA administers the program, sets its policies in consultation with the FAA and aviation community, and receives the reports submitted to the program. The FAA offers those who use the ASRS program two important reporting guarantees: confidentiality and limited immunity. Reports sent to ASRS are held in strict confidence. More than 350,000 reports have been submitted since the program's beginning without a single reporter's identity being revealed. ASRS removes all personal names and other potentially identifying information before entering reports into its database. This system is a very successful, proof-of-concept for gathering safety data in order to provide timely information about safety issues. The ASRS information is crucial to aviation safety efforts both nationally and internationally. It can be utilized as the first step in safety by providing the direction and content to informed policies, procedures, and research, especially human factors. The ASRS process and procedures will be presented as one model of safety reporting feedback systems.

  11. [Prospects in getting accordance between chemical analytic control means and medical technical requirements to safety system concerning chemical weapons destruction].

    Science.gov (United States)

    Rembovskiĭ, V R; Mogilenkova, L A; Savel'eva, E I

    2005-01-01

    The major unit monitoring chemical weapons destruction objects is a system of chemical analyticcontrol over the technologic process procedures and possibility of environment and workplace pollution withtoxicchemicals and their destruction products. At the same time, physical and chemical control means meet sanitary and hygienic requirements incompletely. To provide efficient control, internationally recognized approaches should be adapted to features of Russian system monitoring pollution of chemical weapons destruction objects with toxic chemicals.

  12. Cost benefit analysis of reactor safety systems

    International Nuclear Information System (INIS)

    Maurer, H.A.

    1984-01-01

    Cost/benefit analysis of reactor safety systems is a possibility appropriate to deal with reactor safety. The Commission of the European Communities supported a study on the cost-benefit or cost effectiveness of safety systems installed in modern PWR nuclear power plants. The following systems and their cooperation in emergency cases were in particular investigated in this study: the containment system (double containment), the leakage exhaust and control system, the annulus release exhaust system and the containment spray system. The benefit of a safety system is defined according to its contribution to the reduction of the radiological consequences for the environment after a LOCA. The analysis is so far performed in two different steps: the emergency core cooling system is considered to function properly, failure of the emergency core cooling system is assumed (with the possible consequence of core melt-down) and the results may demonstrate the evidence that striving for cost-effectiveness can produce a safer end result than the philosophy of safety at any cost. (orig.)

  13. FY 1998 annual report on the demonstration tests for establishing load concentration controlling systems. Survey on safety of commercial systems; Fuka shuchu seigyo system kakuritsu jissho shiken 1998 nendo kenkyu hokokusho. Jjitsuyo system anzensei chosa

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1999-03-01

    The demonstration tests are being conducted for establishing load concentration controlling systems, which directly or indirectly control load devices in residential power consumers or the like from a power supplier, as one of the DSM measures. This project is aimed at survey on the systems which support general residential consumers or the like to adequately control loads indirectly, and at clarification of technical essentials the system should have when it is actually put in service and the safety rules to be observed, thereby contributing eventual commercialization of the load concentration controlling systems. The field test results indicate that functions of a monitor set in a domestic consumer can be well operated even by inexperienced persons in handling machines, when they have some experiences. Reliability of a monitoring/controlling device, set in a domestic consumer on a trial basis, can be secured effectively by addition of an LC circuit and changing the modulation mode to FSK. The devices developed on a trial basis are found to be well serviceable for the demonstration tests. The best method for communication with the monitoring/controlling device for electric appliances in a domestic consumer is communication via a power transmission line. (NEDO)

  14. Cardiovascular safety of the oral controlled absorption system (OCAS) formulation of tamsulosin compared to the modified release (MR) formulation

    NARCIS (Netherlands)

    Michel, M. C.; Korstanje, C.; Klauwinkel, W.; Shear, M.; Davies, J.; Quartel, A.

    2005-01-01

    Objective: The potential to interfere with efferent adrenergic drive in the cardiovascular system was tested in elderly healthy subjects for the new oral controlled absorption system (OCAS) 0.4 mg tablet formulation of tamsulosin compared to the modified release (MR) 0.4 mg capsule formulation of

  15. Security for safety critical space borne systems

    Science.gov (United States)

    Legrand, Sue

    1987-01-01

    The Space Station contains safety critical computer software components in systems that can affect life and vital property. These components require a multilevel secure system that provides dynamic access control of the data and processes involved. A study is under way to define requirements for a security model providing access control through level B3 of the Orange Book. The model will be prototyped at NASA-Johnson Space Center.

  16. 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)

  17. Reactor limit control system

    International Nuclear Information System (INIS)

    Rubbel, F.E.

    1982-01-01

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

  18. Radiation (Safety Control) Ordinance 1978

    International Nuclear Information System (INIS)

    1978-01-01

    This Ordinance provides for the control, regulation, possession, use and transport of radioactive substance and irradiating apparatus. The Director of Health is responsible for administration of the Ordinance, which contains detailed provisions concerning the terms and conditions of licences, duties of licensees, medical examinations, maximum radiation doses, precautions to be taken to avoid exceeding such doses. The Ordinance also lays down a system of record-keeping and registration as well as packaging specifications for the transport of radioactive substances. (NEA) [fr

  19. Safety parameter display system for Kalinin NPP

    International Nuclear Information System (INIS)

    Andreev, V.I.; Videneev, E.N.; Tissot, J.C.; Joonekindt, D.; Davidenko, N.N.; Shaftan, G.I.; Dounaev, V.G.; Neboyan, V.T.

    1995-01-01

    The paper discusses the safety parameter display system (SPDS), which is being designed for Kalinin NPP. The assessment of the safety status of the plant is done by the continuous monitoring of six critical safety functions and the corresponding status trees. Besides, a number of additional functions are realized within the scope of KlnNPP, aimed at providing the operator and the safety engineer in the main control room with more detailed information in accidental situation as well as during the normal operation. In particular, these functions are: archiving, data logs and alarm handling, safety actions monitoring, mnemonic diagrams indicating the state of main technological equipment and basic plant parameters, reference data, etc. As compared with the traditional scope of functions of this kind of systems, the functionality of KlnNPP SPDS is significantly expanded due to the inclusion in it the operator support function ''computerized procedures''. The basic SPDS implementation platform is ADACS of SEMA GROUP design. The system architecture includes two workstations in the main control room: one is for reactor operator and the other one for safety engineer. Every station has two CRT screens which ensures computerized procedures implementation and provides for extra services for the operator. Also, the information from the SPDS is transmitted to the local crisis center and to the crisis center of the State utility organization concern ''Rosenergoatom''. (author). 3 refs, 6 figs, 1 tab

  20. Safety assessment of computerized control and protection systems. Report of a technical committee meeting held in Vienna, 12-16 October 1992

    International Nuclear Information System (INIS)

    1994-12-01

    In developing the views expressed in this document, papers were presented by delegates from Member States. A total of 6 papers were presented in all on topics ranging from applications of computerized control and protection systems in older plants and in new advanced reactors to methods for improving software reliability. In addition two informal presentations were provided by a vendor and a licensing authority. These presentations provided valuable insights into the application of computerized control and protection systems and into the concern of software reliability with proposals for diverse 'backup' systems of different types. This was supplemented by utility and vendor presentations on system designs. Following the presentations, three working groups were formed to produce their views on the licensing of software based safety systems on reliability models and techniques for assessment of computerized safety systems, and on systems considered for computerized upgrading (need, criteria, approach, pitfalls and benefits). This document represents these collected views with the papers presented attached as an annex. Refs, figs and tabs

  1. 49 CFR 659.19 - System safety program plan: contents.

    Science.gov (United States)

    2010-10-01

    ... implementation of the system safety program. (j) A description of the process used by the rail transit agency to... the rail transit agency to manage safety issues. (d) The process used to control changes to the system... hazard management program. (n) A description of the process used for facilities and equipment safety...

  2. Quantitative risk assessment of digitalized safety systems

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Sung Min; Lee, Sang Hun; Kang, Hym Gook [KAIST, Daejeon (Korea, Republic of); Lee, Seung Jun [UNIST, Ulasn (Korea, Republic of)

    2016-05-15

    A report published by the U.S. National Research Council indicates that appropriate methods for assessing reliability are key to establishing the acceptability of digital instrumentation and control (I and C) systems in safety-critical plants such as NPPs. Since the release of this issue, the methodology for the probabilistic safety assessment (PSA) of digital I and C systems has been studied. However, there is still no widely accepted method. Kang and Sung found three critical factors for safety assessment of digital systems: detection coverage of fault-tolerant techniques, software reliability quantification, and network communication risk. In reality the various factors composing digitalized I and C systems are not independent of each other but rather closely connected. Thus, from a macro point of view, a method that can integrate risk factors with different characteristics needs to be considered together with the micro approaches to address the challenges facing each factor.

  3. 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.)

  4. An approach to the efficient assessment of safety and usability of computer based control systems, VeNuS 2. Global final report

    International Nuclear Information System (INIS)

    Nelke, T.; Dlugosch, C.; Olaverri Monreal, C.; Sachse, K.; Thuering, M.

    2015-01-01

    Prior to the use of computer-based instrumentation and control the evidence of sufficient safety, development methods and the suitability of man-machine interface must be provided. For this purpose, validation methods must be available, if possible supported by appropriate tools. Based on the multitude of the data which has to be taken into account it is important to generate technical documentation, to realize efficient operation and to prevent human based errors. An approach for computer based generation of user manuals for the operation of technical systems was developed in the VeNuS 2 project. A second goal was to develop an approach to evaluate the usability of safety relevant digital human-machine-interfaces (e.g. for nuclear industries). Therefore a software tool has been developed to assess aspects of usability of user interfaces by considering safety-related priorities. Additionally new or well known methods for provision of evidence of sufficient safety and usability for computer based systems shall be developed in a prototyped way.

  5. Combination therapy with solifenacin and tamsulosin oral controlled absorption system in a single tablet for lower urinary tract symptoms in men: efficacy and safety results from the randomised controlled NEPTUNE trial

    NARCIS (Netherlands)

    van Kerrebroeck, Philip; Chapple, Christopher; Drogendijk, Ted; Klaver, Monique; Sokol, Roman; Speakman, Mark; Traudtner, Klaudia; Drake, Marcus J.; Kiss, G.; Marberger, M.; Strotski, A. V.; Varaksa, A. N.; Vashchula, V.; Dewilde, T.; Braeckman, J.; Roumeguere, T.; Wyndaele, J. J.; Ameye, F.; Everaert, K.; van Cleynenbruegel, B.; de Leval, J.; Vanderkerken, J.; Ackaert, K.; Hiblbauer, J.; Zhanel, P.; Klecka, J.; Lukes, M.; Novak, J.; Lisec, M.; Vrtal, R.; Ondra, D.; Liehne, J.; Tuma, J.; Azzouzi, A.-R.; Wellerand, H.; Jung, J.-L.; Mourey, E.; Colombel, M.; Claude, R.; Ibrahim, H.; Desgrandchamps, F.; Haab, F.; Zerbib, M.; Ruffion, A.; Vincendeau, S.; Haillot, O.; Hentschel, M.; Gerhardt, U.; Hechelmann, W.; de la Rosette, J.

    2013-01-01

    Storage symptoms are particularly bothersome in men with lower urinary tract symptoms (LUTS) but may not be adequately treated by α-blocker monotherapy. To assess the efficacy and safety of a fixed-dose combination (FDC) of solifenacin and an oral controlled absorption system (OCAS) formulation of

  6. The impact of the instrumentation and control systems in the safety of a nuclear plant: a general vision; El impacto de los sistemas de instrumentacion y control en la seguridad de una planta nuclear: una vision general

    Energy Technology Data Exchange (ETDEWEB)

    Celis del Angel, L.; Rivero, T., E-mail: lina.celis@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2011-11-15

    One of the fundamental components so much for the sure operation, like in emergency cases or accident are the equipment s and instrumentation and control systems. The nuclear industry has had some accidents where the instrumentation and control have played and important part: a wrong design, instrumentation lack, faulty systems of safety, etc. At the present time the necessity to modernize the instrumentation and control in a nuclear power plant is before the challenge of finding innovative forms to improve the competitiveness and readiness, reducing operation costs without put ing in risk the safety and reliability of the nuclear power plant. Most of the nuclear power plants require actualizing their instrumentation and control systems, here the digital systems represent a great alternative, improving the performance and the safety, increasing the readiness and reducing the maintenance s. However they require of strict tests that allow assuring their application in critical systems. It is also necessary, the development of modernization programs that allow the programmed substitution of the systems without affecting the readiness of the nuclear power plants. During this whole modernization process will be necessary to put special attention in the cyber-safety because the attacks every time they are more elaborated. Therefore will be necessary to go toward the modernization of the instrumentation and control with the challenge of making without detriment some in the safety of the normal operation and with response reliability in emergency conditions or accident that which represents an effort that should not be postponed in the case of the nuclear power plant of Laguna Verde. (Author)

  7. Mental Workload and Situational Awareness Evaluation of APR1400 Engineered Safety Features- Component Control Activation Systems using Augmented Reality

    International Nuclear Information System (INIS)

    Murungi, Mwongeera; Jung, JaeCheon

    2016-01-01

    In the study, an Augmented Reality procedure guidance support system concept was designed and used as a tool for the measurement of mental workload and Situational awareness of an SRO (Senior Reactor Operator). The EOP was chosen as the scenario for testing because it is the one of the critical plant conditions that requires human intervention and it represents (one of the more) conservative approaches to the test scenarios that are possible. The system is expected to realize an improvement in the level of Situational Awareness and mental workload which have been demonstrated by previous studies to be directly linked with the system response to an emergency situation in the MCR. The planning and design of the project adhered to a Systems Engineering approach in order to provide an optimized framework for ensuring the successful implementation of the system design. Previous study and research into this topic has emphasized the importance of situational awareness in determining the human factor performance issues in the nuclear power plant Control Room operations. This paper broadly defined a technique that successfully used the operator’s mental workload (using NASATLX) and Situational Awareness (using SART) as quantifying measures to evaluate the performance of specific ESF-CCS functions based on human factors. These results show that an improvement of the SA/workload could lead to an improvement of the level of certainty that the emergency situation can be brought under control. It is expected that future development work in this area will yield an actualized Augmented Reality system that could incorporate MCR team control and possibly be implemented in the system validation of other I and C systems

  8. Mental Workload and Situational Awareness Evaluation of APR1400 Engineered Safety Features- Component Control Activation Systems using Augmented Reality

    Energy Technology Data Exchange (ETDEWEB)

    Murungi, Mwongeera; Jung, JaeCheon [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

    2016-10-15

    In the study, an Augmented Reality procedure guidance support system concept was designed and used as a tool for the measurement of mental workload and Situational awareness of an SRO (Senior Reactor Operator). The EOP was chosen as the scenario for testing because it is the one of the critical plant conditions that requires human intervention and it represents (one of the more) conservative approaches to the test scenarios that are possible. The system is expected to realize an improvement in the level of Situational Awareness and mental workload which have been demonstrated by previous studies to be directly linked with the system response to an emergency situation in the MCR. The planning and design of the project adhered to a Systems Engineering approach in order to provide an optimized framework for ensuring the successful implementation of the system design. Previous study and research into this topic has emphasized the importance of situational awareness in determining the human factor performance issues in the nuclear power plant Control Room operations. This paper broadly defined a technique that successfully used the operator’s mental workload (using NASATLX) and Situational Awareness (using SART) as quantifying measures to evaluate the performance of specific ESF-CCS functions based on human factors. These results show that an improvement of the SA/workload could lead to an improvement of the level of certainty that the emergency situation can be brought under control. It is expected that future development work in this area will yield an actualized Augmented Reality system that could incorporate MCR team control and possibly be implemented in the system validation of other I and C systems.

  9. Upgrading safety systems of industrial irradiation facilities

    International Nuclear Information System (INIS)

    Gomes, R.S.; Gomes, J.D.R.L.; Costa, E.L.C.; Costa, M.L.L.; Thomé, Z.D.

    2017-01-01

    The first industrial irradiation facility in operation in Brazil was designed in the 70s. Nowadays, twelve commercial and research facilities are in operation and two already decommissioned. Minor modifications and upgrades, as sensors replacement, have been introduced in these facilities, in order to reduce the technological gap in the control and safety systems. The safety systems are designed in agreement with the codes and standards at the time. Since then, new standards, codes and recommendations, as well as lessons learned from accidents, have been issued by various international committees or regulatory bodies. The rapid advance of the industry makes the safety equipment used in the original construction become obsolete. The decreasing demand for these older products means that they are no longer produced, which can make it impossible or costly to obtain spare parts and the expansion of legacy systems to include new features. This work aims to evaluate existing safety systems at Brazilian irradiation facilities, mainly the oldest facilities, taking into account the recommended IAEA's design requirements. Irrespective of the fact that during its operational period no event with victims have been recorded in Brazilian facilities, and that the regulatory inspections do not present any serious deviations regarding the safety procedures, it is necessary an assessment of safety system with the purpose of bringing their systems to 'the state of the art', avoiding their rapid obsolescence. This study has also taken into account the knowledge, concepts and solutions developed to upgrading safety system in irradiation facilities throughout the world. (author)

  10. Upgrading safety systems of industrial irradiation facilities

    Energy Technology Data Exchange (ETDEWEB)

    Gomes, R.S.; Gomes, J.D.R.L.; Costa, E.L.C.; Costa, M.L.L., E-mail: rogeriog@cnen.gov.br, E-mail: jlopes@cnen.gov.br, E-mail: evaldo@cnen.gov.br, E-mail: mara@cnen.gov.br [Comissão Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil). Diretoria de Radioproteção e Segurança Nuclear; Thomé, Z.D., E-mail: zielithome@gmail.com [Instituto Militar de Engenharia (IME), Rio de Janeiro, RJ (Brazil). Seção de Engenharia Nuclear

    2017-07-01

    The first industrial irradiation facility in operation in Brazil was designed in the 70s. Nowadays, twelve commercial and research facilities are in operation and two already decommissioned. Minor modifications and upgrades, as sensors replacement, have been introduced in these facilities, in order to reduce the technological gap in the control and safety systems. The safety systems are designed in agreement with the codes and standards at the time. Since then, new standards, codes and recommendations, as well as lessons learned from accidents, have been issued by various international committees or regulatory bodies. The rapid advance of the industry makes the safety equipment used in the original construction become obsolete. The decreasing demand for these older products means that they are no longer produced, which can make it impossible or costly to obtain spare parts and the expansion of legacy systems to include new features. This work aims to evaluate existing safety systems at Brazilian irradiation facilities, mainly the oldest facilities, taking into account the recommended IAEA's design requirements. Irrespective of the fact that during its operational period no event with victims have been recorded in Brazilian facilities, and that the regulatory inspections do not present any serious deviations regarding the safety procedures, it is necessary an assessment of safety system with the purpose of bringing their systems to 'the state of the art', avoiding their rapid obsolescence. This study has also taken into account the knowledge, concepts and solutions developed to upgrading safety system in irradiation facilities throughout the world. (author)

  11. Safety in nuclear power systems

    International Nuclear Information System (INIS)

    Myers, L.C.

    1987-05-01

    This paper discusses the issue of safety in complex energy systems and provides brief accounts of some of the most serious reactor accidents that have occurred to date. Details are also provided of Ontario Hydro's problems with Unit 2 at Pickering

  12. Firefighter Safety for PV Systems

    DEFF Research Database (Denmark)

    Mathe, Laszlo; Sera, Dezso; Spataru, Sergiu

    2015-01-01

    An important and highly discussed safety issue for photovoltaic (PV) systems is that as long as the PV panels are illuminated, a high voltage is present at the PV string terminals and cables between the string and inverters that is independent of the state of the inverter's dc disconnection switch...

  13. Safety design guide for safety related systems for CANDU 9

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Duk Su; Chang, Woo Hyun; Lee, Nam Young [Korea Atomic Energy Research Institute, Daeduk (Korea, Republic of); Wright, A.C.D. [Atomic Energy of Canada Ltd., Toronto (Canada)

    1996-03-01

    In general, two types of safety related systems and structures exist in the nuclear plant; The one is a systems and structures which perform safety functions during the normal operation of the plant, and the other is a systems and structures which perform safety functions to mitigate events caused by failure of the normally operating systems or by naturally occurring phenomena. In this safety design guide, these systems are identified in detail, and the major events for which the safety functions are required and the major safety requirements are identified in the list. As the probabilistic safety assessments are completed during the course of the project, additions or deletions to the list may be justified. 3 tabs. (Author) .new.

  14. Safety design guide for safety related systems for CANDU 9

    International Nuclear Information System (INIS)

    Lee, Duk Su; Chang, Woo Hyun; Lee, Nam Young; A. C. D. Wright

    1996-03-01

    In general, two types of safety related systems and structures exist in the nuclear plant; The one is a systems and structures which perform safety functions during the normal operation of the plant, and the other is a systems and structures which perform safety functions to mitigate events caused by failure of the normally operating systems or by naturally occurring phenomena. In this safety design guide, these systems are identified in detail, and the major events for which the safety functions are required and the major safety requirements are identified in the list. As the probabilistic safety assessments are completed during the course of the project, additions or deletions to the list may be justified. 3 tabs. (Author) .new

  15. Stabilization with guaranteed safety using Control Lyapunov–Barrier Function

    NARCIS (Netherlands)

    Romdlony, Muhammad Zakiyullah; Jayawardhana, Bayu

    2016-01-01

    We propose a novel nonlinear control method for solving the problem of stabilization with guaranteed safety for nonlinear systems. The design is based on the merging of the well-known Control Lyapunov Function (CLF) and the recent concept of Control Barrier Function (CBF). The proposed control

  16. The NASA Aviation Safety Reporting System

    Science.gov (United States)

    1983-01-01

    This is the fourteenth in a series of reports based on safety-related incidents submitted to the NASA Aviation Safety Reporting System by pilots, controllers, and, occasionally, other participants in the National Aviation System (refs. 1-13). ASRS operates under a memorandum of agreement between the National Aviation and Space Administration and the Federal Aviation Administration. The report contains, first, a special study prepared by the ASRS Office Staff, of pilot- and controller-submitted reports related to the perceived operation of the ATC system since the 1981 walkout of the controllers' labor organization. Next is a research paper analyzing incidents occurring while single-pilot crews were conducting IFR flights. A third section presents a selection of Alert Bulletins issued by ASRS, with the responses they have elicited from FAA and others concerned. Finally, the report contains a list of publications produced by ASRS with instructions for obtaining them.

  17. Research on Integration of NPP Operational Safety Management Performance Systems

    International Nuclear Information System (INIS)

    Chi, Miao; Shi, Liping

    2014-01-01

    The operational safety management of Nuclear Power Plants demands systematic planning and integrated control. NPPs are following the well-developed safety indicator systems proposed by IAEA Operational Safety Performance Indicator Programme, NRC Reactor Oversight Process or the other institutions. Integration of the systems is proposed to benefiting from the advantages of both systems and avoiding improper application into the real world. The authors analyzed the possibility and necessity for system integration, and propose an indicator system integrating method

  18. Safety critical application of fuzzy control

    International Nuclear Information System (INIS)

    Schildt, G.H.

    1995-01-01

    After an introduction into safety terms a short description of fuzzy logic will be given. Especially, for safety critical applications of fuzzy controllers a possible controller structure will be described. The following items will be discussed: Configuration of fuzzy controllers, design aspects like fuzzfiication, inference strategies, defuzzification and types of membership functions. As an example a typical fuzzy rule set will be presented. Especially, real-time behaviour a fuzzy controllers is mentioned. An example of fuzzy controlling for temperature control purpose within a nuclear reactor together with membership functions and inference strategy of such a fuzzy controller will be presented. (author). 4 refs, 17 figs

  19. System safety education focused on industrial engineering

    Science.gov (United States)

    Johnston, W. L.; Morris, R. S.

    1971-01-01

    An educational program, designed to train students with the specific skills needed to become safety specialists, is described. The discussion concentrates on application, selection, and utilization of various system safety analytical approaches. Emphasis is also placed on the management of a system safety program, its relationship with other disciplines, and new developments and applications of system safety techniques.

  20. Control strategy for power management, efficiency-optimization and operating-safety of a 5-kW solid oxide fuel cell system

    International Nuclear Information System (INIS)

    Zhang, Lin; Jiang, Jianhua; Cheng, Huan; Deng, Zhonghua; Li, Xi

    2015-01-01

    Highlights: • Efficiency optimization associated with simultaneous power and thermal management. • Fast load tracing, fuel starvation, high efficiency and operating safety are considered. • Open loop pre-conditioning current strategy is proposed for load step-up transients. • Feedback control scheme is proposed for load step-up transients. - Abstract: The slow power tracking, operating safety, especially the fuel exhaustion, and high efficiency considerations are the key issues for integrated solid oxide fuel cell (SOFC) systems during power step up transients, resulting in the relatively poor dynamic capabilities and make the transient load following very challenging and must be enhanced. To this end, this paper first focus on addressing the efficiency optimization associated with simultaneous power and thermal management of a 5-kW SOFC system. Particularly, a traverse optimization process including cubic convolution interpolation algorithm are proposed to obtain optimal operating points (OOPs) with the maximum efficiency. Then this paper investigate the current implications on system step-up transient performance, then a two stage pre-conditioning current strategy and a feedback power reference control scheme is proposed for load step-up transients to balance fast load following and fuel starvation, after that safe thermal transient is validated. Simulation results show the efficacy of the control design by demonstrating the fast load following ability while maintaining the safe operation, thus safe; efficient and fast load transition can be achieved

  1. Progress of nuclear safety for symbiosis and sustainability advanced digital instrumentation, control and information systems for nuclear power plants

    CERN Document Server

    Yoshikawa, Hidekazu

    2014-01-01

    This book introduces advanced methods of computational and information systems allowing readers to better understand the state-of-the-art design and implementation technology needed to maintain and enhance the safe operation of nuclear power plants. The subjects dealt with in the book are (i) Full digital instrumentation and control systems and human?machine interface technologies (ii) Risk? monitoring methods for large and? complex? plants (iii) Condition monitors for plant components (iv) Virtual and augmented reality for nuclear power plants and (v) Software reliability verification and val

  2. Performance Testing Methodology for Safety-Critical Programmable Logic Controller

    International Nuclear Information System (INIS)

    Kim, Chang Ho; Oh, Do Young; Kim, Ji Hyeon; Kim, Sung Ho; Sohn, Se Do

    2009-01-01

    The Programmable Logic Controller (PLC) for use in Nuclear Power Plant safety-related applications is being developed and tested first time in Korea. This safety-related PLC is being developed with requirements of regulatory guideline and industry standards for safety system. To test that the quality of the developed PLC is sufficient to be used in safety critical system, document review and various product testings were performed over the development documents for S/W, H/W, and V/V. This paper provides the performance testing methodology and its effectiveness for PLC platform conducted by KOPEC

  3. How could intelligent safety transport systems enhance safety ?

    NARCIS (Netherlands)

    Wiethoff, M. Heijer, T. & Bekiaris, E.

    2017-01-01

    In Europe, many deaths and injured each years are the cost of today's road traffic. Therefore, it is wise to look for possible solutions for enhancing traffic safety. Some Advanced Driver Assistance Systems (ADAS) are expected to increase safety, but they may also evoke new safety hazards. Only

  4. [Establishment of Quality Control System of Nucleic Acid Detection for Ebola Virus in Sierra Leone-China Friendship Biological Safety Laboratory].

    Science.gov (United States)

    Wang, Qin; Zhang, Yong; Nie, Kai; Wang, Huanyu; Du, Haijun; Song, Jingdong; Xiao, Kang; Lei, Wenwen; Guo, Jianqiang; Wei, Hejiang; Cai, Kun; Wang, Yanhai; Wu, Jiang; Gerald, Bangura; Kamara, Idrissa Laybohr; Liang, Mifang; Wu, Guizhen; Dong, Xiaoping

    2016-03-01

    The quality control process throughout the Ebola virus nucleic acid detection in Sierra Leone-China Friendship Biological Safety Laboratory (SLE-CHN Biosafety Lab) was described in detail, in order to comprehensively display the scientific, rigorous, accurate and efficient practice in detection of Ebola virus of first batch detection team in SLE-CHN Biosafety Lab. Firstly, the key points of laboratory quality control system was described, including the managements and organizing, quality control documents and information management, instrument, reagents and supplies, assessment, facilities design and space allocation, laboratory maintenance and biosecurity. Secondly, the application of quality control methods in the whole process of the Ebola virus detection, including before the test, during the test and after the test, was analyzed. The excellent and professional laboratory staffs, the implementation of humanized management are the cornerstone of the success; High-level biological safety protection is the premise for effective quality control and completion of Ebola virus detection tasks. And professional logistics is prerequisite for launching the laboratory diagnosis of Ebola virus. The establishment and running of SLE-CHN Biosafety Lab has landmark significance for the friendship between Sierra Leone and China, and the lab becomes the most important base for Ebola virus laboratory testing in Sierra Leone.

  5. Components for containment enclosures. Part 4: Ventilation and gas-cleaning systems such as filters, traps, safety and regulation valves, control and protection devices

    International Nuclear Information System (INIS)

    2001-01-01

    ISO 11933 consists of the following parts, under the general title Components for containment enclosures: Part 1: Glove/bag ports, bungs for glove/bag ports, enclosure rings and interchangeable units; Part 2: Gloves, welded bags, gaiters for remote-handling tongs and for manipulators; Part 3: Transfer systems such as plain doors, airlock chambers, double door transfer systems, leaktight connections for waste drums; Part 4: Ventilation and gas-cleaning systems such as filters, traps, safety and regulation valves, control and protection devices; Part 5: Penetrations for electrical and fluid circuits. This part of ISO 11933 specifies the design criteria and the characteristics of various components used for ventilation and gas-cleaning in containment enclosures. These components are either directly fixed to the containment enclosure wall, or used in the environment of a shielded or unshielded containment enclosure or line of such enclosures. They can be used alone or in conjunction with other mechanical components, including those specified in ISO 11933-1 and ISO 11933-3. This part of ISO 11933 is applicable to: filtering devices, including high-efficiency particulate air (HEPA) filters and iodine traps; safety valves and pressure regulators; systems ensuring the mechanical protection of containment enclosures; control and pressure-measurement devices

  6. Nuclear power systems: Their safety

    International Nuclear Information System (INIS)

    Myers, L.C.

    1993-01-01

    Mankind utilizes energy in many forms and from a variety of sources. Canada is one of a growing number of countries which have chosen to embrace nuclear-electric generation as a component of their energy systems. As of August 1992 there were 433 power reactors operating in 35 countries and accounting for more than 15% of the world's production of electricity. In 1992, thirteen countries derived at least 25% of their electricity from nuclear units, with France leading at nearly 70%. In the same year, Canada produced about 16% of its electricity from nuclear units. Some 68 power reactors are under construction in 16 countries, enough to expand present generating capacity by close to 20%. No human endeavour carries the guarantee of perfect safety and the question of whether or not nuclear-electric generation represents an 'acceptable' risk to society has long been vigorously debated. Until the events of late April 1986, nuclear safety had indeed been an issue for discussion, for some concern, but not for alarm. The accident at the Chernobyl reactor in the USSR has irrevocably changed all that. This disaster brought the matter of nuclear safety back into the public mind in a dramatic fashion. This paper discusses the issue of safety in complex energy systems and provides brief accounts of some of the most serious reactor accidents which have occurred to date. (author). 7 refs

  7. Radiation safety systems at the NSLS

    International Nuclear Information System (INIS)

    Dickinson, T.

    1987-04-01

    This report describes design principles that were used to establish the radiation safety systems at the National Synchrotron Light Source. The author described existing safety systems and the history of partial system failures. 1 fig

  8. Safety factor profile control in a tokamak

    CERN Document Server

    Bribiesca Argomedo, Federico; Prieur, Christophe

    2014-01-01

    Control of the Safety Factor Profile in a Tokamak uses Lyapunov techniques to address a challenging problem for which even the simplest physically relevant models are represented by nonlinear, time-dependent, partial differential equations (PDEs). This is because of the  spatiotemporal dynamics of transport phenomena (magnetic flux, heat, densities, etc.) in the anisotropic plasma medium. Robustness considerations are ubiquitous in the analysis and control design since direct measurements on the magnetic flux are impossible (its estimation relies on virtual sensors) and large uncertainties remain in the coupling between the plasma particles and the radio-frequency waves (distributed inputs). The Brief begins with a presentation of the reference dynamical model and continues by developing a Lyapunov function for the discretized system (in a polytopic linear-parameter-varying formulation). The limitations of this finite-dimensional approach motivate new developments in the infinite-dimensional framework. The t...

  9. Safety assessment of HLW geological disposal system

    International Nuclear Information System (INIS)

    Naito, Morimasa

    2006-01-01

    In accordance with the Japanese nuclear program, the liquid waste with a high level of radioactivity arising from reprocessing is solidified in a stable glass matrix (vitrification) in stainless steel fabrication containers. The vitrified waste is referred to as high-level radioactive waste (HLW), and is characterized by very high initial radioactivity which, even though it decreases with time, presents a potential long-term risk. It is therefore necessary to thoroughly manage HLW from human and his environment. After vitrification, HLW is stored for a period of 30 to 50 years to allow cooling, and finally disposed of in a stable geological environment at depths greater than 300 m below surface. The deep underground environment, in general, is considered to be stable over geological timescales compared with surface environment. By selecting an appropriate disposal site, therefore, it is considered to be feasible to isolate the waste in the repository from man and his environment until such time as radioactivity levels have decayed to insignificance. The concept of geological disposal in Japan is similar to that in other countries, being based on a multibarrier system which combines the natural geological environment with engineered barriers. It should be noted that geological disposal concept is based on a passive safety system that does not require any institutional control for assuring long term environmental safety. To demonstrate feasibility of safe HLW repository concept in Japan, following technical steps are essential. Selection of a geological environment which is sufficiently stable for disposal (site selection). Design and installation of the engineered barrier system in a stable geological environment (engineering measures). Confirmation of the safety of the constructed geological disposal system (safety assessment). For site selection, particular consideration is given to the long-term stability of the geological environment taking into account the fact

  10. IAEA Safety Standards on Management Systems and Safety Culture

    International Nuclear Information System (INIS)

    Persson, Kerstin Dahlgren

    2007-01-01

    The IAEA has developed a new set of Safety Standard for applying an integrated Management System for facilities and activities. The objective of the new Safety Standards is to define requirements and provide guidance for establishing, implementing, assessing and continually improving a Management System that integrates safety, health, environmental, security, quality and economic related elements to ensure that safety is properly taken into account in all the activities of an organization. With an integrated approach to management system it is also necessary to include the aspect of culture, where the organizational culture and safety culture is seen as crucial elements of the successful implementation of this management system and the attainment of all the goals and particularly the safety goals of the organization. The IAEA has developed a set of service aimed at assisting it's Member States in establishing. Implementing, assessing and continually improving an integrated management system. (author)

  11. Safety monitoring in process and control

    International Nuclear Information System (INIS)

    Esparza, V. Jr.; Sebo, D.E.

    1984-01-01

    Safety Functions provide a method of ensuring the safe operation of any large-scale processing plant. Successful implementation of safety functions requires continuous monitoring of safety function values and trends. Because the volume of information handled by a plant operator occassionally can become overwhelming, attention may be diverted from the primary concern of maintaining plant safety. With this in mind EG and G, Idaho developed various methods and techniques for use in a computerized Safety Function Monitoring System and tested the application of these techniques using a simulated nuclear power plant, the Loss-of-Fluid Test Facility (LOFT) at the Idaho National Engineering Laboratory (INEL). This paper presents the methods used in the development of a Safety Function Monitoring System

  12. Safety-factor profile tailoring by improved electron cyclotron system for sawtooth control and reverse shear scenarios in ITER

    International Nuclear Information System (INIS)

    Zucca, C.; Sauter, O.; Fable, E.; Henderson, M. A.; Polevoi, A.; Farina, D.; Ramponi, G.; Saibene, G.; Zohm, H.

    2008-01-01

    The effect of the predicted local electron cyclotron current driven by the optimized electron cyclotron system on ITER is discussed. A design variant was recently proposed to enlarge the physics program covered by the upper and equatorial launchers. By extending the functionality range of the upper launcher, significant control capabilities of the sawtooth period can be obtained. The upper launcher improvement still allows enough margin to exceed the requirements for neoclassical tearing mode stabilization, for which it was originally designed. The analysis of the sawtooth control is carried on with the ASTRA transport code, coupled with the threshold model by Por-celli, to study the control capabilities of the improved upper launcher on the sawtooth instability. The simulations take into account the significant stabilizing effect of the fusion alpha particles. The sawtooth period can be increased by a factor of 1.5 with co-ECCD outside the q = 1 surface, and decreased by at least 30% with co-ECCD inside q = 1. The present ITER base-line design has the electron cyclotron launchers providing only co-ECCD. The variant for the equatorial launcher proposes the possibility to drive counter-ECCD with 1 of the 3 rows of mirrors: the counter-ECCD can then be balanced with co-ECCD and provide pure ECH with no net driven current. The difference between full co-ECCD off-axis using all 20MW from the equatorial launcher and 20MW co-ECCD driven by 2/3 from the equatorial launcher and 1/3 from the upper launcher is shown to be negligible. Cnt-ECCD also offers greater control of the plasma current density, therefore this analysis addresses the performance of the equatorial launcher to control the central q profile. The equatorial launcher is shown to control very efficiently the value of q 0.2 -q min in advanced scenarios, if one row provides counter-ECCD.

  13. System analysis of vehicle active safety problem

    Science.gov (United States)

    Buznikov, S. E.

    2018-02-01

    The problem of the road transport safety affects the vital interests of the most of the population and is characterized by a global level of significance. The system analysis of problem of creation of competitive active vehicle safety systems is presented as an interrelated complex of tasks of multi-criterion optimization and dynamic stabilization of the state variables of a controlled object. Solving them requires generation of all possible variants of technical solutions within the software and hardware domains and synthesis of the control, which is close to optimum. For implementing the task of the system analysis the Zwicky “morphological box” method is used. Creation of comprehensive active safety systems involves solution of the problem of preventing typical collisions. For solving it, a structured set of collisions is introduced with its elements being generated also using the Zwicky “morphological box” method. The obstacle speed, the longitudinal acceleration of the controlled object and the unpredictable changes in its movement direction due to certain faults, the road surface condition and the control errors are taken as structure variables that characterize the conditions of collisions. The conditions for preventing typical collisions are presented as inequalities for physical variables that define the state vector of the object and its dynamic limits.

  14. ACP Facility Safety Surveillance System Installation

    International Nuclear Information System (INIS)

    You, Gil Sung; Kook, D. H.; Choung, W. M.; Ku, J. H.; Cho, I. J.; You, G. S.; Kwon, K. C.; Lee, W. K.; Lee, E. P.

    2006-10-01

    The Advanced spent fuel Conditioning Process is under development for effective management of spent fuel by converting UO 2 into U-metal. For demonstration of this process, α-γ type new hotcell was built in the IMEF basement. All facilities which treat radioactive materials must manage CCTV system which is under control of Health Physics department. Three main points (including hotcell rear door area) have each camera, but operators who are in charge of facility management need to check the safety of the facility immediately through the network in his office. This needs introduce additional network cameras installation and this new surveillance system is expected to update the whole safety control ability with existing system

  15. Reactor safety: the Nova computer system

    International Nuclear Information System (INIS)

    Eisgruber, H.; Stadelmann, W.

    1991-01-01

    After instances of maloperation, the causes of defects, the effectiveness of the measures taken to control the situation, and possibilities to avoid future recurrences need to be investigated above all before the plant is restarted. The most important aspect in all these efforts is to check the sequence in time, and the completeness, of the control measures initiated automatically. For this verification, a computer system is used instead of time-consuming manual analytical techniques, which produces the necessary information almost in real time. The results are available within minutes after completion of the measures initiated automatically. As all short-term safety functions are initiated by automatic systems, their consistent and comprehensive verification results in a clearly higher level of safety. The report covers the development of the computer system, and its implementation, in the Gundremmingen nuclear power station. Similar plans are being pursued in Biblis and Muelheim-Kaerlich. (orig.) [de

  16. Role of computers in CANDU safety systems

    International Nuclear Information System (INIS)

    Hepburn, G.A.; Gilbert, R.S.; Ichiyen, N.M.

    1985-01-01

    Small digital computers are playing an expanding role in the safety systems of CANDU nuclear generating stations, both as active components in the trip logic, and as monitoring and testing systems. The paper describes three recent applications: (i) A programmable controller was retro-fitted to Bruce ''A'' Nuclear Generating Station to handle trip setpoint modification as a function of booster rod insertion. (ii) A centralized monitoring computer to monitor both shutdown systems and the Emergency Coolant Injection system, is currently being retro-fitted to Bruce ''A''. (iii) The implementation of process trips on the CANDU 600 design using microcomputers. While not truly a retrofit, this feature was added very late in the design cycle to increase the margin against spurious trips, and has now seen about 4 unit-years of service at three separate sites. Committed future applications of computers in special safety systems are also described. (author)

  17. Environmental safety to decomposer invertebrates of azadirachtin (neem) as a systemic insecticide in trees to control emerald ash borer.

    Science.gov (United States)

    Kreutzweiser, David; Thompson, Dean; Grimalt, Susana; Chartrand, Derek; Good, Kevin; Scarr, Taylor

    2011-09-01

    The non-target effects of an azadirachtin-based systemic insecticide used for control of wood-boring insect pests in trees were assessed on litter-dwelling earthworms, leaf-shredding aquatic insects, and microbial communities in terrestrial and aquatic microcosms. The insecticide was injected into the trunks of ash trees at a rate of 0.2 gazadirachtin cm(-1) tree diameter in early summer. At the time of senescence, foliar concentrations in most (65%) leaves where at or below detection (azadirachtin) and the average concentration among leaves overall at senescence was 0.19 mg kg(-1). Leaves from the azadirachtin-treated trees at senescence were added to microcosms and responses by test organisms were compared to those in microcosms containing leaves from non-treated ash trees (controls). No significant reductions were detected among earthworm survival, leaf consumption rates, growth rates, or cocoon production, aquatic insect survival and leaf consumption rates, and among terrestrial and aquatic microbial decomposition of leaf material in comparison to controls. In a further set of microcosm tests containing leaves from intentional high-dose trees, the only significant, adverse effect detected was a reduction in microbial decomposition of leaf material, and only at the highest test concentration (∼6 mg kg(-1)). Results indicated no significant adverse effects on litter-dwelling earthworms or leaf-shredding aquatic insects at concentrations up to at least 30 × the expected field concentrations at operational rates, and at 6 × expected field concentrations for adverse effects on microbial decomposition. We conclude that when azadirachtin is used as a systemic insecticide in trees for control of insect pests such as the invasive wood-boring beetle, emerald ash borer, resultant foliar concentrations in senescent leaf material are likely to pose little risk of harm to decomposer invertebrates. Crown Copyright © 2011. Published by Elsevier Inc. All rights reserved.

  18. Luxury cruise? The safety potential of advanced cruise control.

    NARCIS (Netherlands)

    Oei, H.L.

    2003-01-01

    The principles of advanced cruise control (ACC) are outlined and the requirements for an ACC system are described. An intelligent cruise control system fitted in a Nissan Primera was tested on the road over a 2-week period by 10 drivers, eight of which were experts in road safety. Most test-drives

  19. 75 FR 67450 - Pipeline Safety: Control Room Management Implementation Workshop

    Science.gov (United States)

    2010-11-02

    ... regulations to address human factors and other aspects of control room management for certain pipelines where controllers use supervisory control and data acquisition (SCADA) systems. Under the final rule, pipeline... Washington, DC on October 22, 2010. Jeffrey D. Wiese, Associate Administrator for Pipeline Safety. [FR Doc...

  20. A study on LAN applications in nuclear safety systems

    International Nuclear Information System (INIS)

    Kim, Sung; Lee, Young Ryul; Koo, Jun Mo; Han, Jai Bok

    1995-01-01

    It is a general tendency to digitalize the conventional relay based I and C systems in nuclear power plant. But, the digitalisation of nuclear safety systems has many a difficulty to surmount. The typical one thing of many difficulties is the data communication problem between local controllers and systems. The network architecture built with LAN (Local Area Network) in digital systems of the other industries are general. But in case of nuclear safety systems many considerations in point of safety and license are required to implement it in the field. In this parer, some considerations for applying LAN in nuclear safety systems were reviewed

  1. DESIGN PACKAGE 1E SYSTEM SAFETY ANALYSIS

    Energy Technology Data Exchange (ETDEWEB)

    M. Salem

    1995-06-23

    The purpose of this analysis is to systematically identify and evaluate hazards related to the Yucca Mountain Project Exploratory Studies Facility (ESF) Design Package 1E, Surface Facilities, (for a list of design items included in the package 1E system safety analysis see section 3). This process is an integral part of the systems engineering process; whereby safety is considered during planning, design, testing, and construction. A largely qualitative approach was used since a radiological System Safety Analysis is not required. The risk assessment in this analysis characterizes the accident scenarios associated with the Design Package 1E structures/systems/components(S/S/Cs) in terms of relative risk and includes recommendations for mitigating all identified risks. The priority for recommending and implementing mitigation control features is: (1) Incorporate measures to reduce risks and hazards into the structure/system/component design, (2) add safety devices and capabilities to the designs that reduce risk, (3) provide devices that detect and warn personnel of hazardous conditions, and (4) develop procedures and conduct training to increase worker awareness of potential hazards, on methods to reduce exposure to hazards, and on the actions required to avoid accidents or correct hazardous conditions.

  2. International conference on the safety and security of radioactive sources: Towards a global system for the continuous control of sources throughout their life cycle. Contributed papers

    International Nuclear Information System (INIS)

    2005-01-01

    The objective of the conference is to promote a wide exchange of information on key issues relating to the safety and security of radioactive sources, including: drawing up an inventory; finding a solution without delay to situations resulting from past activities; preparing for the future by defining a global cooperative approach to the continuous control of radioactive sources during their life cycle. It is expected that the conference will foster a better understanding of the risks posed by these sources from the point of view of radiation safety and the threat associated with some of them in the event of malevolent use, and will help in finding ways of reducing the likelihood of the occurrence of a radiological incident or accident, or of a malevolent act. It is also expected to identify the preparedness and response measures that are necessary and to facilitate a common understanding on the feasibility of creating a sustainable global system for ensuring the safety and security of radioactive sources

  3. International conference on the safety and security of radioactive sources: Towards a global system for the continuous control of sources throughout their life cycle. Contributed papers

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2005-07-01

    The objective of the conference is to promote a wide exchange of information on key issues relating to the safety and security of radioactive sources, including: drawing up an inventory; finding a solution without delay to situations resulting from past activities; preparing for the future by defining a global cooperative approach to the continuous control of radioactive sources during their life cycle. It is expected that the conference will foster a better understanding of the risks posed by these sources from the point of view of radiation safety and the threat associated with some of them in the event of malevolent use, and will help in finding ways of reducing the likelihood of the occurrence of a radiological incident or accident, or of a malevolent act. It is also expected to identify the preparedness and response measures that are necessary and to facilitate a common understanding on the feasibility of creating a sustainable global system for ensuring the safety and security of radioactive sources.

  4. Software for computer based systems important to safety in nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2004-01-01

    Computer based systems are of increasing importance to safety in nuclear power plants as their use in both new and older plants is rapidly increasing. They are used both in safety related applications, such as some functions of the process control and monitoring systems, as well as in safety critical applications, such as reactor protection or actuation of safety features. The dependability of computer based systems important to safety is therefore of prime interest and should be ensured. With current technology, it is possible in principle to develop computer based instrumentation and control systems for systems important to safety that have the potential for improving the level of safety and reliability with sufficient dependability. However, their dependability can be predicted and demonstrated only if a systematic, fully documented and reviewable engineering process is followed. Although a number of national and international standards dealing with quality assurance for computer based systems important to safety have been or are being prepared, internationally agreed criteria for demonstrating the safety of such systems are not generally available. It is recognized that there may be other ways of providing the necessary safety demonstration than those recommended here. The basic requirements for the design of safety systems for nuclear power plants are provided in the Requirements for Design issued in the IAEA Safety Standards Series.The IAEA has issued a Technical Report to assist Member States in ensuring that computer based systems important to safety in nuclear power plants are safe and properly licensed. The report provides information on current software engineering practices and, together with relevant standards, forms a technical basis for this Safety Guide. The objective of this Safety Guide is to provide guidance on the collection of evidence and preparation of documentation to be used in the safety demonstration for the software for computer based

  5. Software for computer based systems important to safety in nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2005-01-01

    Computer based systems are of increasing importance to safety in nuclear power plants as their use in both new and older plants is rapidly increasing. They are used both in safety related applications, such as some functions of the process control and monitoring systems, as well as in safety critical applications, such as reactor protection or actuation of safety features. The dependability of computer based systems important to safety is therefore of prime interest and should be ensured. With current technology, it is possible in principle to develop computer based instrumentation and control systems for systems important to safety that have the potential for improving the level of safety and reliability with sufficient dependability. However, their dependability can be predicted and demonstrated only if a systematic, fully documented and reviewable engineering process is followed. Although a number of national and international standards dealing with quality assurance for computer based systems important to safety have been or are being prepared, internationally agreed criteria for demonstrating the safety of such systems are not generally available. It is recognized that there may be other ways of providing the necessary safety demonstration than those recommended here. The basic requirements for the design of safety systems for nuclear power plants are provided in the Requirements for Design issued in the IAEA Safety Standards Series.The IAEA has issued a Technical Report to assist Member States in ensuring that computer based systems important to safety in nuclear power plants are safe and properly licensed. The report provides information on current software engineering practices and, together with relevant standards, forms a technical basis for this Safety Guide. The objective of this Safety Guide is to provide guidance on the collection of evidence and preparation of documentation to be used in the safety demonstration for the software for computer based

  6. Software for computer based systems important to safety in nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2000-01-01

    Computer based systems are of increasing importance to safety in nuclear power plants as their use in both new and older plants is rapidly increasing. They are used both in safety related applications, such as some functions of the process control and monitoring systems, as well as in safety critical applications, such as reactor protection or actuation of safety features. The dependability of computer based systems important to safety is therefore of prime interest and should be ensured. With current technology, it is possible in principle to develop computer based instrumentation and control systems for systems important to safety that have the potential for improving the level of safety and reliability with sufficient dependability. However, their dependability can be predicted and demonstrated only if a systematic, fully documented and reviewable engineering process is followed. Although a number of national and international standards dealing with quality assurance for computer based systems important to safety have been or are being prepared, internationally agreed criteria for demonstrating the safety of such systems are not generally available. It is recognized that there may be other ways of providing the necessary safety demonstration than those recommended here. The basic requirements for the design of safety systems for nuclear power plants are provided in the Requirements for Design issued in the IAEA Safety Standards Series.The IAEA has issued a Technical Report to assist Member States in ensuring that computer based systems important to safety in nuclear power plants are safe and properly licensed. The report provides information on current software engineering practices and, together with relevant standards, forms a technical basis for this Safety Guide. The objective of this Safety Guide is to provide guidance on the collection of evidence and preparation of documentation to be used in the safety demonstration for the software for computer based

  7. Comparison of the Safety Critical Software V and V Requirements for the Research Reactor Instrumentation and Control System

    Energy Technology Data Exchange (ETDEWEB)

    Joo, Sungmoon; Suh, Yong-Suk; Park, Cheol [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    This study was motivated by a research reactor project where the owner of the project and the equipment vendors are from two different standards frameworks. This paper reviews two major standards frameworks - NRC-IEEE and IAEA-IEC - and the software classification schemes as a background, then discuss the V and V issue. The purpose of this paper is by no means to solve the cross-standards-framework qualification issue, but, rather, is to remind the stakeholders of research reactor projects. V and V are also essential for the approval from regulatory bodies. As standards define or recommend consolidated engineering practices, methods, or criteria, V and V activities for software qualification are not exceptional. Within a standards framework, usually, the processes for the qualification of safety-critical software are well-established such that the safety is maximized while minimizing the compromises in software quality, safety, and reliability. When, however, multiple standards frameworks are involved in a research reactor project, it is difficult for equipment vendors to implement appropriate V and V activities as there is no unified view on this cross-standards-framework qualification issue yet. There are two major standards frameworks for safety-critical software development in nuclear industry. Unfortunately different safety classifications for software and thus different requirements for qualification are in place. What makes things worse is that (i) there are ambiguities in the standards and rooms for each stakeholders’ interpretation, and (ii) there is no one-to-one mapping between the associated V and V methods and activities. These may put the stakeholders of research reactor projects in trouble.

  8. Comparison of the Safety Critical Software V and V Requirements for the Research Reactor Instrumentation and Control System

    International Nuclear Information System (INIS)

    Joo, Sungmoon; Suh, Yong-Suk; Park, Cheol

    2016-01-01

    This study was motivated by a research reactor project where the owner of the project and the equipment vendors are from two different standards frameworks. This paper reviews two major standards frameworks - NRC-IEEE and IAEA-IEC - and the software classification schemes as a background, then discuss the V and V issue. The purpose of this paper is by no means to solve the cross-standards-framework qualification issue, but, rather, is to remind the stakeholders of research reactor projects. V and V are also essential for the approval from regulatory bodies. As standards define or recommend consolidated engineering practices, methods, or criteria, V and V activities for software qualification are not exceptional. Within a standards framework, usually, the processes for the qualification of safety-critical software are well-established such that the safety is maximized while minimizing the compromises in software quality, safety, and reliability. When, however, multiple standards frameworks are involved in a research reactor project, it is difficult for equipment vendors to implement appropriate V and V activities as there is no unified view on this cross-standards-framework qualification issue yet. There are two major standards frameworks for safety-critical software development in nuclear industry. Unfortunately different safety classifications for software and thus different requirements for qualification are in place. What makes things worse is that (i) there are ambiguities in the standards and rooms for each stakeholders’ interpretation, and (ii) there is no one-to-one mapping between the associated V and V methods and activities. These may put the stakeholders of research reactor projects in trouble

  9. Safety performance monitoring of autonomous marine systems

    International Nuclear Information System (INIS)

    Thieme, Christoph A.; Utne, Ingrid B.

    2017-01-01

    The marine environment is vast, harsh, and challenging. Unanticipated faults and events might lead to loss of vessels, transported goods, collected scientific data, and business reputation. Hence, systems have to be in place that monitor the safety performance of operation and indicate if it drifts into an intolerable safety level. This article proposes a process for developing safety indicators for the operation of autonomous marine systems (AMS). The condition of safety barriers and resilience engineering form the basis for the development of safety indicators, synthesizing and further adjusting the dual assurance and the resilience based early warning indicator (REWI) approaches. The article locates the process for developing safety indicators in the system life cycle emphasizing a timely implementation of the safety indicators. The resulting safety indicators reflect safety in AMS operation and can assist in planning of operations, in daily operational decision-making, and identification of improvements. Operation of an autonomous underwater vehicle (AUV) exemplifies the process for developing safety indicators and their implementation. The case study shows that the proposed process leads to a comprehensive set of safety indicators. It is expected that application of the resulting safety indicators consequently will contribute to safer operation of current and future AMS. - Highlights: • Process for developing safety indicators for autonomous marine systems. • Safety indicators based on safety barriers and resilience thinking. • Location of the development process in the system lifecycle. • Case study on AUV demonstrating applicability of the process.

  10. Safety precautions in atomic pile control (1962)

    International Nuclear Information System (INIS)

    Furet, J.

    1962-01-01

    We have been led to study the problem of safety in atomic pile control as a result of our participation on the one hand in the planning of C.E.A. atomic piles, and on the other hand in the pile safety sub omission considering atomic pile safety of operational or planned C.E.A. piles. We have thus had to consider the wishes occurring in piles during their operation and also their behaviour in the dynamic state The present work deals mainly with the importance of intrinsic safety devices, with the influence of reactivity variations on the power fluctuations during accidental operation, and with the development of robust and reliable safety appliances. The starting p accident has been especially studied both for low-flux piles where a compromise is necessary between the response time of the safety appliances and the statistical fluctuations and for high lux piles where xenon poisoning has an effect on the lower limit of the velocity of reactivity liberation. The desirability has been stressed of automation as a safety factor in atomic pile control. The details required for an understanding of the diagrams of the apparatus are given. (author) [fr

  11. Quality control guarantees the safety of radiotherapy

    International Nuclear Information System (INIS)

    Aaltonen, P.

    1994-01-01

    While radiotherapy equipment has seen some decisive improvements in the last few decades, the technology has also become more complicated. The advanced equipment produces increasingly good treatment results, but the condition of the equipment must be controlled efficiently so as to eliminate any defects that might jeopardise patient safety. The quality assurance measures that are taken to show that certain equipment functions as required are known as quality control. The advanced equipment and stricter requirements set for the precision of radiotherapy have meant that more attention must be paid to quality control. The present radiation legislation stipulates that radiotherapy equipment must undergo regular quality control. The implementation of the quality control is supervised by the Finnish Centre for Radiation and Nuclear Safety (STUK). Hospitals carry out quality control in accordance with a programme approved by STUK, and STUK inspectors periodically visit hospitals to check the results of quality control. (orig.)

  12. Safety climate and culture: Integrating psychological and systems perspectives.

    Science.gov (United States)

    Casey, Tristan; Griffin, Mark A; Flatau Harrison, Huw; Neal, Andrew

    2017-07-01

    Safety climate research has reached a mature stage of development, with a number of meta-analyses demonstrating the link between safety climate and safety outcomes. More recently, there has been interest from systems theorists in integrating the concept of safety culture and to a lesser extent, safety climate into systems-based models of organizational safety. Such models represent a theoretical and practical development of the safety climate concept by positioning climate as part of a dynamic work system in which perceptions of safety act to constrain and shape employee behavior. We propose safety climate and safety culture constitute part of the enabling capitals through which organizations build safety capability. We discuss how organizations can deploy different configurations of enabling capital to exert control over work systems and maintain safe and productive performance. We outline 4 key strategies through which organizations to reconcile the system control problems of promotion versus prevention, and stability versus flexibility. (PsycINFO Database Record (c) 2017 APA, all rights reserved).

  13. Emerging standards with application to accelerator safety systems

    International Nuclear Information System (INIS)

    Mahoney, K.L.; Robertson, H.P.

    1997-01-01

    This paper addresses international standards which can be applied to the requirements for accelerator personnel safety systems. Particular emphasis is given to standards which specify requirements for safety interlock systems which employ programmable electronic subsystems. The work draws on methodologies currently under development for the medical, process control, and nuclear industries

  14. Design for safety: theoretical framework of the safety aspect of BIM system to determine the safety index

    Directory of Open Access Journals (Sweden)

    Ai Lin Evelyn Teo

    2016-12-01

    Full Text Available Despite the safety improvement drive that has been implemented in the construction industry in Singapore for many years, the industry continues to report the highest number of workplace fatalities, compared to other industries. The purpose of this paper is to discuss the theoretical framework of the safety aspect of a proposed BIM System to determine a Safety Index. An online questionnaire survey was conducted to ascertain the current workplace safety and health situation in the construction industry and explore how BIM can be used to improve safety performance in the industry. A safety hazard library was developed based on the main contributors to fatal accidents in the construction industry, determined from the formal records and existing literature, and a series of discussions with representatives from the Workplace Safety and Health Institute (WSH Institute in Singapore. The results from the survey suggested that the majority of the firms have implemented the necessary policies, programmes and procedures on Workplace Safety and Health (WSH practices. However, BIM is still not widely applied or explored beyond the mandatory requirement that building plans should be submitted to the authorities for approval in BIM format. This paper presents a discussion of the safety aspect of the Intelligent Productivity and Safety System (IPASS developed in the study. IPASS is an intelligent system incorporating the buildable design concept, theory on the detection, prevention and control of hazards, and the Construction Safety Audit Scoring System (ConSASS. The system is based on the premise that safety should be considered at the design stage, and BIM can be an effective tool to facilitate the efforts to enhance safety performance. IPASS allows users to analyse and monitor key aspects of the safety performance of the project before the project starts and as the project progresses.

  15. Systems engineered health and safety criteria for safety analysis reports

    International Nuclear Information System (INIS)

    Beitel, G.A.; Morcos, N.

    1993-01-01

    The world of safety analysis is filled with ambiguous words: codes and standards, consequences and risks, hazard and accident, and health and safety. These words have been subject to disparate interpretations by safety analysis report (SAR) writers, readers, and users. open-quotes Principal health and safety criteriaclose quotes has been one of the most frequently misused phrases; rarely is it used consistently or effectively. This paper offers an easily understood definition for open-quotes principal health and safety criteriaclose quotes and uses systems engineering to convert an otherwise mysterious topic into the primary means of producing an integrated SAR. This paper is based on SARs being written for environmental restoration and waste management activities for the U.S. Department of Energy (DOE). Requirements for these SARs are prescribed in DOE Order 5480-23, open-quotes Nuclear Safety Analysis Reports.close quotes

  16. Development of a safety parameter supervision system for Angra-1

    International Nuclear Information System (INIS)

    Silva, R.A. da; Thome Filho, Z.D.; Schirru, R.; Martinez, A.S.; Oliveira, L.F.S. de

    1986-01-01

    The Safety Parameter Supervision System (SSPS) which is a computerized system for monitoring essential parameters in real time, determining the safety status and emergency procedures for returning normal reactor operation, in case of an anomaly occurrence, is presented. The SSPS consists of three sub-systems: Integrated parameter monitoring system which gives to operators an integrated vision of values of a parameter set, able to detect any deviation of normal reactor operation; safety critical function system which evaluates safety status in terms of a safety critical function set appointed in advance, and in case of violation of any critical function, it initiates the adequate emergency procedure to return normal operation; and safety parameter computer system which carries out the arquirement of analogic and digital control signals of nuclear power plant. (M.C.K.) [pt

  17. An analysis of safety control effectiveness

    International Nuclear Information System (INIS)

    Son, K.S.; Melchers, R.E.; Kal, W.M.

    2000-01-01

    The cost of injuries and 'accidents' to an organisation is very important in establishing how much it should spend on safety control. Despite the usefulness of information about the cost of a company's accidents, it is not customary accounting practice to make these data available. Of the two kinds of costs incurred by a company through occupational injuries and accidents, direct costs and indirect costs; the direct costs are much easier to estimate. However, the uninsured costs are usually more critical and should be estimated by each company. The authors investigate a general model to estimate the above costs and hence to establish efficient safety control. One construction company has been a pilot for this study. By analysing actual company data for three years, it is found that the efficient safety control cost should be 1.2-1.3% of total contract costs

  18. 77 FR 70409 - System Safety Program

    Science.gov (United States)

    2012-11-26

    ...-0060, Notice No. 2] 2130-AC31 System Safety Program AGENCY: Federal Railroad Administration (FRA... rulemaking (NPRM) published on September 7, 2012, FRA proposed regulations to require commuter and intercity passenger railroads to develop and implement a system safety program (SSP) to improve the safety of their...

  19. Preliminary safety evaluation for CSR1000 with passive safety system

    International Nuclear Information System (INIS)

    Wu, Pan; Gou, Junli; Shan, Jianqiang; Zhang, Bo; Li, Xiang

    2014-01-01

    Highlights: • The basic information of a Chinese SCWR concept CSR1000 is introduced. • An innovative passive safety system is proposed for CSR1000. • 6 Transients and 3 accidents are analysed with system code SCTRAN. • The passive safety systems greatly mitigate the consequences of these incidents. • The inherent safety of CSR1000 is enhanced. - Abstract: This paper describes the preliminary safety analysis of the Chinese Supercritical water cooled Reactor (CSR1000), which is proposed by Nuclear Power Institute of China (NPIC). The two-pass core design applied to CSR1000 decreases the fuel cladding temperature and flattens the power distribution of the core at normal operation condition. Each fuel assembly is made up of four sub-assemblies with downward-flow water rods, which is favorable to the core cooling during abnormal conditions due to the large water inventory of the water rods. Additionally, a passive safety system is proposed for CSR1000 to increase the safety reliability at abnormal conditions. In this paper, accidents of “pump seizure”, “loss of coolant flow accidents (LOFA)”, “core depressurization”, as well as some typical transients are analysed with code SCTRAN, which is a one-dimensional safety analysis code for SCWRs. The results indicate that the maximum cladding surface temperatures (MCST), which is the most important safety criterion, of the both passes in the mentioned incidents are all below the safety criterion by a large margin. The sensitivity analyses of the delay time of RCPs trip in “loss of offsite power” and the delay time of RMT actuation in “loss of coolant flowrate” were also included in this paper. The analyses have shown that the core design of CSR1000 is feasible and the proposed passive safety system is capable of mitigating the consequences of the selected abnormalities

  20. Comprehensive Lifecycle for Assuring System Safety

    Science.gov (United States)

    Knight, John C.; Rowanhill, Jonathan C.

    2017-01-01

    CLASS is a novel approach to the enhancement of system safety in which the system safety case becomes the focus of safety engineering throughout the system lifecycle. CLASS also expands the role of the safety case across all phases of the system's lifetime, from concept formation to decommissioning. As CLASS has been developed, the concept has been generalized to a more comprehensive notion of assurance becoming the driving goal, where safety is an important special case. This report summarizes major aspects of CLASS and contains a bibliography of papers that provide additional details.

  1. Safety aspects of core power distribution surveillance and control

    International Nuclear Information System (INIS)

    Beraha, D.; Grumbach, R.; Hoeld, A.; Werner, W.

    1978-01-01

    The incentives for improved core surveillance and core control systems are outlined. An efficient code for evaluating the power distribution is indispensable for designing and testing such a system. The characteristics of the core simulator QUABOX/CUBBOX and the features required for off-line and on-line applications are described. The important role of the simulator for the safety assessment of a digital core control system is underlined. With regard to the safety aspects of core control, possible disturbances are classified. Simulation results are given concerning the failure of a control actuator. It is shown that means can be devised to prevent unstable behaviour of the control system and, furthermore, to contribute to a safe reactor operation by accounting for process disturbances. (author)

  2. Aircraft Loss-of-Control: Analysis and Requirements for Future Safety-Critical Systems and Their Validation

    Science.gov (United States)

    Belcastro, Christine M.

    2011-01-01

    Loss of control remains one of the largest contributors to fatal aircraft accidents worldwide. Aircraft loss-of-control accidents are complex, resulting from numerous causal and contributing factors acting alone or more often in combination. Hence, there is no single intervention strategy to prevent these accidents. This paper summarizes recent analysis results in identifying worst-case combinations of loss-of-control accident precursors and their time sequences, a holistic approach to preventing loss-of-control accidents in the future, and key requirements for validating the associated technologies.

  3. Documents pertaining to safety control of nuclear facilities

    International Nuclear Information System (INIS)

    1998-01-01

    The Finnish Radiation and Nuclear Safety Authority (STUK) controls the safety of nuclear facilities in Finland. This control encompasses on one hand the evaluation of plant safety on the basis of plans and analyses pertaining to the plant and on the other hand the inspection of plant structures, systems and components as well as of operational activity. STUK also monitors plants operational experience feedback and technical developments in the field, as well as the development of safety research and takes the necessary measures on their basis. Guide YVL 1.1 describes how STUK controls the design, construction and operation of nuclear power plants. The documents to be submitted to STUK are described in the nuclear energy legislation and YVL guides. This guide presents the mode of delivery, quality, contents and number of documents to be submitted to STUK

  4. Application range affected by software failures in safety relevant instrumentation and control systems of nuclear power plants; Auswirkungsbereiche von Softwarefehlern in sicherheitstechnisch wichtigen Einrichtungen von Kernkraftwerken

    Energy Technology Data Exchange (ETDEWEB)

    Jopen, Manuela; Mbonjo, Herve; Sommer, Dagmar; Ulrich, Birte

    2017-03-15

    This report presents results that have been developed within a BMUB-funded research project (Promotion Code 3614R01304). The overall objective of this project was to broaden the knowledge base of GRS regarding software failures and their impact in software-based instrumentation and control (I and C) systems. To this end, relevant definitions and terms in standards and publications (DIN, IEEE standards, IAEA standards, NUREG publications) as well as in the German safety requirements for nuclear power plants were analyzed first. In particular, it was found that the term ''software fault'' is defined differently and partly contradictory in the considered literature sources. For this reason, a definition of software fault was developed on the basis of the software life cycle of software-based I and C systems within the framework of this project, which takes into account the various aspects relevant to software faults and their related effects. It turns out that software failures result from latent faults in a software-based control system, which can lead to a non-compliant behavior of a software-based I and C system. Hereby a distinction should be made between programming faults and specification faults. In a further step, operational experience with software failures in software-based I and C systems in nuclear facilities and in nonnuclear sector was investigated. The identified events were analyzed with regard to their cause and impacts and the analysis results were summarized. Based on the developed definition of software failure and on the COMPSIS-classification scheme for events related to software based I and C systems, the COCS-classification scheme was developed to classify events from operating experience with software failures, in which the events are classified according to the criteria ''cause'', ''affected system'', ''impact'' and ''CCF potential''. This

  5. METHODS OF CONTROL DIPHTHERIA VACCINE SAFETY

    Directory of Open Access Journals (Sweden)

    Isayenko Ye. Yu

    2016-12-01

    of toxin it's examined the WHO's proposal to use an intradermal method on guinea pigs and tests on cell cultures. Attention is drawn to the fact that the determination of specific toxicity in cell culture can be carried out at presence of the approval of this method of a national control authority and sensitivity rates no less than in experiments on guinea pigs. The determining of specific toxicity of ready vaccine by subcutaneous method is described. The publication gave a test for elevated toxicity of the final product by intraperitoneal infection of mice and guinea pigs. It’s cited the WHO recommendations aimed at removing the possibility of recovery of the refined toxin toxicity. Checking vaccines toxicity, pyrogenicity, sterility, allergenicity, teratogenicity, mutagenicity and immunogenicity mainly performed on laboratory animals. The review examined the unreliability of animal experiments and the need to find alternative methods for determining the toxicity without their use particularly in light of the “3R”concept. Methods for determining diphtherial toxin using cell cultures is considered, namely, colony overlay test (COT, tests using a monolayer of HeLa cell culture, a culture of Vero cells (kidney cells of african green monkeys , a culture of CHO cells (cells of Chinese hamster ovary, which are based on the toxin cytopathic effect on sensitive cell culture. Their advantages and disadvantages are listed. An alternative method for the quantitative detection of C. diphtheriae toxin using the polystyrene plate coated with monoclonal antibody to the part of the diphtheria toxin which defines its binding to the cell, is described. It’s regarded the cytotoxic effect of diphtheria toxin on cells of the immune system of mice and guinea pigs: splenocytes, adhesive phagocytes i B- lymphocytes.

  6. ESSAA: Embedded system safety analysis assistant

    Science.gov (United States)

    Wallace, Peter; Holzer, Joseph; Guarro, Sergio; Hyatt, Larry

    1987-01-01

    The Embedded System Safety Analysis Assistant (ESSAA) is a knowledge-based tool that can assist in identifying disaster scenarios. Imbedded software issues hazardous control commands to the surrounding hardware. ESSAA is intended to work from outputs to inputs, as a complement to simulation and verification methods. Rather than treating the software in isolation, it examines the context in which the software is to be deployed. Given a specified disasterous outcome, ESSAA works from a qualitative, abstract model of the complete system to infer sets of environmental conditions and/or failures that could cause a disasterous outcome. The scenarios can then be examined in depth for plausibility using existing techniques.

  7. Review report: safety and reliability issues on digital instrumentation and control systems in nuclear power plants and United States Nuclear Regulatory Commission`s dispositions

    Energy Technology Data Exchange (ETDEWEB)

    Watanabe, Norio; Suzudo, Tomoaki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-09-01

    Recently, digital instrumentation and control (I and C) systems have been applied to nuclear power plants (NPPs) in various countries. Introduction of digital I and C systems, however, raises special issues on design, implementation, safety and licensing. Since FY 1997, the Japan Atomic Energy Research Institute (JAERI) has been carrying out a project, Study on Reliability of Digital I and C Systems, which includes extensive reviews of design approaches, technical standards, regulatory processes, especially, in the United States. This report summarizes the results from the study of National Research Council (NRC) and the U.S. Nuclear Regulatory Commission`s (USNRC`s) responses to the recommendations made by the NRC`s study. That study identified six technical key issues (system aspects of digital I and C technology, software quality assurance, common-mode software failure potential, safety and reliability assessment methods, human factors and man-machine interface, dedication of commercial off-the-shelf hardware and software) and two strategic key issues (case-by-case licensing process, adequacy of technical infrastructure) that arise from the introduction of digital I and C technology and then, made recommendations to the USNRC for coping with digital I and C applications. The USNRC responded to each recommendation and showed their own dispositions in which the USNRC agreed with most of the recommendations. In Japan, it is expected that introduction of digital I and C technology is inevitable in NPPs because the vendors are gradually discontinuing support and stocking of analog components. To cope with such situations, there is a need to develop and update the standards and guidelines applicable to digital I and C technology. The key issues and the USNRC`s dispositions provided in this report is believed to be useful for developing and updating them. (J.P.N.)

  8. Review report: safety and reliability issues on digital instrumentation and control systems in nuclear power plants and United States Nuclear Regulatory Commission's dispositions

    International Nuclear Information System (INIS)

    Watanabe, Norio; Suzudo, Tomoaki

    1998-09-01

    Recently, digital instrumentation and control (I and C) systems have been applied to nuclear power plants (NPPs) in various countries. Introduction of digital I and C systems, however, raises special issues on design, implementation, safety and licensing. Since FY 1997, the Japan Atomic Energy Research Institute (JAERI) has been carrying out a project, Study on Reliability of Digital I and C Systems, which includes extensive reviews of design approaches, technical standards, regulatory processes, especially, in the United States. This report summarizes the results from the study of National Research Council (NRC) and the U.S. Nuclear Regulatory Commission's (USNRC's) responses to the recommendations made by the NRC's study. That study identified six technical key issues (system aspects of digital I and C technology, software quality assurance, common-mode software failure potential, safety and reliability assessment methods, human factors and man-machine interface, dedication of commercial off-the-shelf hardware and software) and two strategic key issues (case-by-case licensing process, adequacy of technical infrastructure) that arise from the introduction of digital I and C technology and then, made recommendations to the USNRC for coping with digital I and C applications. The USNRC responded to each recommendation and showed their own dispositions in which the USNRC agreed with most of the recommendations. In Japan, it is expected that introduction of digital I and C technology is inevitable in NPPs because the vendors are gradually discontinuing support and stocking of analog components. To cope with such situations, there is a need to develop and update the standards and guidelines applicable to digital I and C technology. The key issues and the USNRC's dispositions provided in this report is believed to be useful for developing and updating them. (J.P.N.)

  9. Remote mobile communication in safety support system

    International Nuclear Information System (INIS)

    Inagaki, Kanji; Kobayashi, Hiroyuki; Hatanaka, Takahiro; Sakuma, Akira; Fukumoto, Akira; Ikeda, Jun

    1999-01-01

    Safety Support System (SSS) is a computerized operator support system for nuclear power plants, which is now under development. The concept of SSS covers 1) earlier detection of failure symptom and prediction of its influence to the plant operation, 2) improved transparency and robustness of plant control systems, 3) advanced human-machine interface and communication. The authors have been working on the third concept and proposed a remote mobile communication system called Plant Communication System (PCS). PCS aims to realize convenient communication between main control room and other areas such as plant local areas and site offices, using Personal Handyphone System (PHS) and wireless LAN (Local Area Network). PCS can transmit not only data but also graphic displays and dynamic video displays between the main control room and plant local areas. MPEG4 (Moving Picture Experts Group 4) technology is utilized in video data compression and decompression. The authors have developed the special multiplexing unit that connects PHS Cell Stations (CSs) and exiting coaxial cables. Voice recognition and announcement capability is also realized in the system, which enables verbal retrieval of information in the computer systems in the main control room from local areas. (author)

  10. Logical safety system for triggering off the protection action of a safety actuator

    International Nuclear Information System (INIS)

    Plaige, Yves.

    1982-01-01

    This invention applies in particular to the emergency triggering of safety actuators controlling the shutdown of a nuclear reactor. This logical safety system includes four redundant lines each composed, inter alia, of a logical circuit for controlling the triggering of a protection action, a logical alarm circuit connected to the control circuit and a logical inhibiting circuit making it impossible to inhibit several alarm circuits simultaneously [fr

  11. Risk Level Based Management System: a control banding model for occupational health and safety risk management in a highly regulated environment

    Energy Technology Data Exchange (ETDEWEB)

    Zalk, D; Kamerzell, R; Paik, S; Kapp, J; Harrington, D; Swuste, P

    2009-05-27

    The Risk Level Based Management System (RLBMS) is an occupational risk management (ORM) model that focuses occupational safety, hygeiene, and health (OSHH) resources on the highest risk procedures at work. This article demonstrates the model's simplicity through an implementation within a heavily regulated research institution. The model utilizes control banding strategies with a stratification of four risk levels (RLs) for many commonly performed maintenance and support activities, characterizing risk consistently for comparable tasks. RLBMS creates an auditable tracking of activities, maximizes OSHH professional field time, and standardizes documentation and control commensurate to a given task's RL. Validation of RLs and their exposure control effectiveness is collected in a traditional quantitative collection regime for regulatory auditing. However, qualitative risk assessment methods are also used within this validation process. Participatory approaches are used throughout the RLBMS process. Workers are involved in all phases of building, maintaining, and improving this model. This work participation also improves the implementation of established controls.

  12. A controlled, randomized, head-to-head comparison of the Prolieve thermodilatation System versus the Targis System for benign prostatic hyperplasia: safety, procedural tolerability, and clinical results.

    Science.gov (United States)

    Shore, Neal D; Sethi, Parminder S

    2010-09-01

    Compare safety and tolerability of the Prolieve(®) System with the Targis(®) System using objective and subjective measures. Thirty-four men with symptomatic benign prostatic hyperplasia (BPH) were randomized to a single treatment with either the Prolieve or Targis system; 30 were treated and then followed for 6 months. After post-treatment bladder fill with ≥200 mL saline, patients were catheterized if they could not void after 2 hours or had a postvoid residual >150 mL. Catheter use, visual analog scale (VAS) tolerability scores, American Urological Association scores, and safety were assessed. After treatment, 15/16 (94%) Prolieve patients remained catheter-free compared with 3/14 (21%) Targis patients (P = 0.0001). Foley catheter indwelling time was 58.8 hours for the one Prolieve patient compared with 103.9 hrs (range 54-270 h) for the Targis patients (n = 9). Targis patients' catheterization requirements were: Seven Foley only, two intermittent self-catheters only, and two needing both. Intermittent self-catheterization continued for 1 month in two Targis patients. VAS tolerability scores were 24% to 50% lower during Prolieve treatment vs Targis (P 0.05). Overall, the incidence of device-related adverse events was 31% (Prolieve) compared with 64% (Targis) (P > 0.05)-most prevalently, urinary retention, dysuria, and hematuria. No device-related serious adverse events occurred. Prolieve provided enhanced near-term therapeutic benefit over Targis as assessed by catheterization, tolerability, and symptom relief, which may assist physician and patient decision-making when selecting an office-based transurethral microwave therapy option for patients.

  13. System Design and the Safety Basis

    International Nuclear Information System (INIS)

    Ellingson, Darrel

    2008-01-01

    The objective of this paper is to present the Bechtel Jacobs Company, LLC (BJC) Lessons Learned for system design as it relates to safety basis documentation. BJC has had to reconcile incomplete or outdated system description information with current facility safety basis for a number of situations in recent months. This paper has relevance in multiple topical areas including documented safety analysis, decontamination and decommissioning (D and D), safety basis (SB) implementation, safety and design integration, potential inadequacy of the safety analysis (PISA), technical safety requirements (TSR), and unreviewed safety questions. BJC learned that nuclear safety compliance relies on adequate and well documented system design information. A number of PIS As and TSR violations occurred due to inadequate or erroneous system design information. As a corrective action, BJC assessed the occurrences caused by systems design-safety basis interface problems. Safety systems reviewed included the Molten Salt Reactor Experiment (MSRE) Fluorination System, K-1065 fire alarm system, and the K-25 Radiation Criticality Accident Alarm System. The conclusion was that an inadequate knowledge of system design could result in continuous non-compliance issues relating to nuclear safety. This was especially true with older facilities that lacked current as-built drawings coupled with the loss of 'historical knowledge' as personnel retired or moved on in their careers. Walkdown of systems and the updating of drawings are imperative for nuclear safety compliance. System design integration with safety basis has relevance in the Department of Energy (DOE) complex. This paper presents the BJC Lessons Learned in this area. It will be of benefit to DOE contractors that manage and operate an aging population of nuclear facilities

  14. NASA System Safety Handbook. Volume 1; System Safety Framework and Concepts for Implementation

    Science.gov (United States)

    Dezfuli, Homayoon; Benjamin, Allan; Everett, Christopher; Smith, Curtis; Stamatelatos, Michael; Youngblood, Robert

    2011-01-01

    System safety assessment is defined in NPR 8715.3C, NASA General Safety Program Requirements as a disciplined, systematic approach to the analysis of risks resulting from hazards that can affect humans, the environment, and mission assets. Achievement of the highest practicable degree of system safety is one of NASA's highest priorities. Traditionally, system safety assessment at NASA and elsewhere has focused on the application of a set of safety analysis tools to identify safety risks and formulate effective controls.1 Familiar tools used for this purpose include various forms of hazard analyses, failure modes and effects analyses, and probabilistic safety assessment (commonly also referred to as probabilistic risk assessment (PRA)). In the past, it has been assumed that to show that a system is safe, it is sufficient to provide assurance that the process for identifying the hazards has been as comprehensive as possible and that each identified hazard has one or more associated controls. The NASA Aerospace Safety Advisory Panel (ASAP) has made several statements in its annual reports supporting a more holistic approach. In 2006, it recommended that "... a comprehensive risk assessment, communication and acceptance process be implemented to ensure that overall launch risk is considered in an integrated and consistent manner." In 2009, it advocated for "... a process for using a risk-informed design approach to produce a design that is optimally and sufficiently safe." As a rationale for the latter advocacy, it stated that "... the ASAP applauds switching to a performance-based approach because it emphasizes early risk identification to guide designs, thus enabling creative design approaches that might be more efficient, safer, or both." For purposes of this preface, it is worth mentioning three areas where the handbook emphasizes a more holistic type of thinking. First, the handbook takes the position that it is important to not just focus on risk on an individual

  15. Assessment of Quadrivalent Human Papillomavirus Vaccine Safety Using the Self-Controlled Tree-Temporal Scan Statistic Signal-Detection Method in the Sentinel System.

    Science.gov (United States)

    Yih, W Katherine; Maro, Judith C; Nguyen, Michael; Baker, Meghan A; Balsbaugh, Carolyn; Cole, David V; Dashevsky, Inna; Mba-Jonas, Adamma; Kulldorff, Martin

    2018-06-01

    The self-controlled tree-temporal scan statistic-a new signal-detection method-can evaluate whether any of a wide variety of health outcomes are temporally associated with receipt of a specific vaccine, while adjusting for multiple testing. Neither health outcomes nor postvaccination potential periods of increased risk need be prespecified. Using US medical claims data in the Food and Drug Administration's Sentinel system, we employed the method to evaluate adverse events occurring after receipt of quadrivalent human papillomavirus vaccine (4vHPV). Incident outcomes recorded in emergency department or inpatient settings within 56 days after first doses of 4vHPV received by 9- through 26.9-year-olds in 2006-2014 were identified using International Classification of Diseases, Ninth Revision, diagnosis codes and analyzed by pairing the new method with a standard hierarchical classification of diagnoses. On scanning diagnoses of 1.9 million 4vHPV recipients, 2 statistically significant categories of adverse events were found: cellulitis on days 2-3 after vaccination and "other complications of surgical and medical procedures" on days 1-3 after vaccination. Cellulitis is a known adverse event. Clinically informed investigation of electronic claims records of the patients with "other complications" did not suggest any previously unknown vaccine safety problem. Considering that thousands of potential short-term adverse events and hundreds of potential risk intervals were evaluated, these findings add significantly to the growing safety record of 4vHPV.

  16. Safety assessment for Generation IV nuclear systems

    International Nuclear Information System (INIS)

    Leahy, T.J.

    2012-01-01

    The Generation IV International Forum (GIF) Risk and Safety Working Group (RSWG) was created to develop an effective approach for the safety of Generation IV advanced nuclear energy systems. Recent RSWG work has focused on the definition of an integrated safety assessment methodology (ISAM) for evaluating the safety of Generation IV systems. ISAM is an integrated 'tool-kit' consisting of 5 analytical techniques that are available and matched to appropriate stages of Generation IV system concept development: 1) qualitative safety features review - QSR, 2) phenomena identification and ranking table - PIRT, 3) objective provision tree - OPT, 4) deterministic and phenomenological analyses - DPA, and 5) probabilistic safety analysis - PSA. The integrated methodology is intended to yield safety-related insights that help actively drive the evolving design throughout the technology development cycle, potentially resulting in enhanced safety, reduced costs, and shortened development time

  17. Watershed safety and quality control by safety threshold method

    Science.gov (United States)

    Da-Wei Tsai, David; Mengjung Chou, Caroline; Ramaraj, Rameshprabu; Liu, Wen-Cheng; Honglay Chen, Paris

    2014-05-01

    Taiwan was warned as one of the most dangerous countries by IPCC and the World Bank. In such an exceptional and perilous island, we would like to launch the strategic research of land-use management on the catastrophe prevention and environmental protection. This study used the watershed management by "Safety Threshold Method" to restore and to prevent the disasters and pollution on island. For the deluge prevention, this study applied the restoration strategy to reduce total runoff which was equilibrium to 59.4% of the infiltration each year. For the sediment management, safety threshold management could reduce the sediment below the equilibrium of the natural sediment cycle. In the water quality issues, the best strategies exhibited the significant total load reductions of 10% in carbon (BOD5), 15% in nitrogen (nitrate) and 9% in phosphorus (TP). We found out the water quality could meet the BOD target by the 50% peak reduction with management. All the simulations demonstrated the safety threshold method was helpful to control the loadings within the safe range of disasters and environmental quality. Moreover, from the historical data of whole island, the past deforestation policy and the mistake economic projects were the prime culprits. Consequently, this study showed a practical method to manage both the disasters and pollution in a watershed scale by the land-use management.

  18. Systematic evaluation program review of NRC Safety Topic VI-10.A associated with the electrical, instrumentation and control portions of the testing of reactor trip system and engineered safety features, including response time for the Dresden station, Unit II nuclear power plant

    International Nuclear Information System (INIS)

    St Leger-Barter, G.

    1980-11-01

    This report documents the technical evaluation and review of NRC Safety Topic VI-10.A, associated with the electrical, instrumentation, and control portions of the testing of reactor trip systems and engineered safety features including response time for the Dresden II nuclear power plant, using current licensing criteria

  19. Nuclear safety risk control in the outage of CANDU unit

    International Nuclear Information System (INIS)

    Wu Mingliang; Zheng Jianhua

    2014-01-01

    Nuclear fuel remains in the core during the outage of CANDU unit, but there are still nuclear safety risks such as reactor accidental criticality, fuel element failure due to inability to properly remove residual heat. Furthermore, these risks are aggravated by the weakening plant system configuration and multiple cross operations during the outage. This paper analyzes the phases where there are potential nuclear safety risks on the basis of the typical critical path arrangement of the outage of Qinshan NPP 3 and introduces a series of CANDU-specific risk control measures taken during the past plant outages to ensure nuclear safety during the unit outage. (authors)

  20. 49 CFR 193.2619 - Control systems.

    Science.gov (United States)

    2010-10-01

    ... ADMINISTRATION, DEPARTMENT OF TRANSPORTATION (CONTINUED) PIPELINE SAFETY LIQUEFIED NATURAL GAS FACILITIES..., and control systems for internal shutoff valves for bottom penetration tanks must be inspected and...

  1. Autonomous system for launch vehicle range safety

    Science.gov (United States)

    Ferrell, Bob; Haley, Sam

    2001-02-01

    The Autonomous Flight Safety System (AFSS) is a launch vehicle subsystem whose ultimate goal is an autonomous capability to assure range safety (people and valuable resources), flight personnel safety, flight assets safety (recovery of valuable vehicles and cargo), and global coverage with a dramatic simplification of range infrastructure. The AFSS is capable of determining current vehicle position and predicting the impact point with respect to flight restriction zones. Additionally, it is able to discern whether or not the launch vehicle is an immediate threat to public safety, and initiate the appropriate range safety response. These features provide for a dramatic cost reduction in range operations and improved reliability of mission success. .

  2. Progress report: 1996 Radiation Safety Systems Division

    International Nuclear Information System (INIS)

    Bhagwat, A.M.; Sharma, D.N.; Abani, M.C.; Mehta, S.K.

    1997-01-01

    The activities of Radiation Safety Systems Division include (i) development of specialised monitoring systems and radiation safety information network, (ii) radiation hazards control at the nuclear fuel cycle facilities, the radioisotope programmes at Bhabha Atomic Research Centre (BARC) and for the accelerators programme at BARC and Centre for Advanced Technology (CAT), Indore. The systems on which development and upgradation work was carried out during the year included aerial gamma spectrometer, automated environment monitor using railway network, radioisotope package monitor and air monitors for tritium and alpha active aerosols. Other R and D efforts at the division included assessment of risk for radiation exposures and evaluation of ICRP 60 recommendations in the Indian context, shielding evaluation and dosimetry for the new upcoming accelerator facilities and solid state nuclear track detector techniques for neutron measurements. The expertise of the divisional members was provided for 36 safety committees of BARC and Atomic Energy Regulatory Board (AERB). Twenty three publications were brought out during the year 1996. (author)

  3. Development and application of digital safety system in NPPs

    International Nuclear Information System (INIS)

    Kwon, Keechoon; Kim, Changhwoi; Lee, Dongyoung

    2012-01-01

    This paper describes the development of digital safety system in NPPs based on safety- grade programmable logic controller (PLC) platform and its application to real NPP construction. The digital safety system consists of a reactor protection system and an engineered safety feature-component control system. The safety-grade PLC platform was developed so that it meets the requirements of the regulation. The PLC consists of various modules such as a power module, a processor module, communication modules, digital input/output modules, analog input/output modules, a LOCA bus extension module, and a high-speed pulse counter module. The reactor protection system is designed with a redundant 4-channel architecture, and every channel is implemented with the same architecture. A single channel consists of a redundant bi-stable processor, a redundant coincidence processor, an automatic test and interface processor, and a cabinet operator module. The engineered safety feature-component control system is designed with four redundant divisions, and implemented with the PLC platform. The principal components of an individual division are fault tolerant group controllers, loop controllers, a test and interface processor, a cabinet operator module and a control channel gateway. The topical report is submitted to the regulatory body, and got safety evaluation report from the regulatory body. Also, the developed system is tested in the integrated performance validation facility. It is decided that the digital safety system applied to Shin-Uljin unit 1 and 2 after a topical report approval and validation test. Design changes occur in the digital safety system that is applied to an actual nuclear power plant construction, and the PLC has also been upgraded

  4. Development and application of digital safety system in NPPs

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Keechoon; Kim, Changhwoi; Lee, Dongyoung [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-03-15

    This paper describes the development of digital safety system in NPPs based on safety- grade programmable logic controller (PLC) platform and its application to real NPP construction. The digital safety system consists of a reactor protection system and an engineered safety feature-component control system. The safety-grade PLC platform was developed so that it meets the requirements of the regulation. The PLC consists of various modules such as a power module, a processor module, communication modules, digital input/output modules, analog input/output modules, a LOCA bus extension module, and a high-speed pulse counter module. The reactor protection system is designed with a redundant 4-channel architecture, and every channel is implemented with the same architecture. A single channel consists of a redundant bi-stable processor, a redundant coincidence processor, an automatic test and interface processor, and a cabinet operator module. The engineered safety feature-component control system is designed with four redundant divisions, and implemented with the PLC platform. The principal components of an individual division are fault tolerant group controllers, loop controllers, a test and interface processor, a cabinet operator module and a control channel gateway. The topical report is submitted to the regulatory body, and got safety evaluation report from the regulatory body. Also, the developed system is tested in the integrated performance validation facility. It is decided that the digital safety system applied to Shin-Uljin unit 1 and 2 after a topical report approval and validation test. Design changes occur in the digital safety system that is applied to an actual nuclear power plant construction, and the PLC has also been upgraded.

  5. Safety design requirements for safety systems and components of JSFR

    International Nuclear Information System (INIS)

    Kubo, Shigenobu; Shimakawa, Yoshio; Yamano, Hidemasa; Kotake, Shoji

    2011-01-01

    Safety design requirements for JSFR were summarized taking the development targets of the FaCT project and design feature of JSFR into account. The related safety principle and requirements for Monju, CRBRP, PRISM, SPX, LWRs, IAEA standards, goals of GIF, basic principle of INPRO etc. were also taken into account so that the safety design requirements can be a next-generation global standard. The development targets for safety and reliability are set based on those of FaCT, namely, ensuring safety and reliability equal to future LWR and related fuel cycle facilities. In order to achieve these targets, the defence-in-depth concept is used as the basic safety design principle. General features of the safety design requirements are 1) Achievement of higher reliability, 2) Achievement of higher inspectability and maintainability, 3) Introduction of passive safety features, 4) Reduction of operator action needs, 5) Design consideration against Beyond Design Basis Events, 6) In-Vessel Retention of degraded core materials, 7) Prevention and mitigation against sodium chemical reactions, and 8) Design against external events. The current specific requirements for each system and component are summarized taking the basic design concept of JSFR into account, which is an advanced loop-type large-output power plant with a mixed-oxide-fuelled core. (author)

  6. The PIANC Safety Factor System for Breakwaters

    DEFF Research Database (Denmark)

    Burcharth, H. F.

    2000-01-01

    The paper presents a summary of the recommendations for implementation of safety in breakwater designs given by the PIANC PTC IT Working Group No 12 on Analysis of Rubble Mound Breakwaters with Vertical and Inclined Concrete Walls. The working groups developed for the most important failure modes...... a system of partial safety factors which facilitate design to any target safety level....

  7. Safety case development with SBVR-based controlled language

    NARCIS (Netherlands)

    Luo, Y.; van den Brand, M.G.J.; Kiburse, A.; Desfray, P.; Philipe, J.; Hammoudi, S.; Pires, L.F.

    2015-01-01

    Safety case development is highly recommended by some safety standards to justify the safety of a system. The Goal Structuring Notation (GSN) is a popular approach to construct a safety case. However, the content of the safety case elements, such as safety claims, is in natural language. Therefore,

  8. Qualification of FPGA-Based Safety-Related PRM System

    International Nuclear Information System (INIS)

    Miyazaki, Tadashi; Oda, Naotaka; Goto, Yasushi; Hayashi, Toshifumi

    2011-01-01

    Toshiba has developed Non-rewritable (NRW) Field Programmable Gate Array (FPGA)-based safety-related Instrumentation and Control (I and C) system. Considering application to safety-related systems, nonvolatile and non-rewritable FPGA which is impossible to be changed after once manufactured has been adopted in Toshiba FPGA-based system. FPGA is a device which consists only of basic logic circuits, and FPGA performs defined processing which is configured by connecting the basic logic circuit inside the FPGA. FPGA-based system solves issues existing both in the conventional systems operated by analog circuits (analog-based system) and the systems operated by central processing unit (CPU-based system). The advantages of applying FPGA are to keep the long-life supply of products, improving testability (verification), and to reduce the drift which may occur in analog-based system. The system which Toshiba developed this time is Power Range Neutron Monitor (PRM). Toshiba is planning to expand application of FPGA-based technology by adopting this development process to the other safety-related systems such as RPS from now on. Toshiba developed a special design process for NRW-FPGA-based safety-related I and C systems. The design process resolves issues for many years regarding testability of the digital system for nuclear safety application. Thus, Toshiba NRW-FPGA-based safety-related I and C systems has much advantage to be a would standard of the digital systems for nuclear safety application. (author)

  9. USAEC Controls for Nuclear Criticality Safety

    Energy Technology Data Exchange (ETDEWEB)

    McCluggage, W. C. [Division of Operational Safety, United States Atomic Energy Commission Washington, DC (United States)

    1966-05-15

    This is a paper written to provide a broad general view of the United States Atomic Energy Commission's controls for nuclear criticality safety within its own facilities. Included also is a brief' discussion of the USAEC's methods of obtaining assurance that the controls are being applied. The body of the document contains three sections. The first two describe the functions of the USAEC; the third deals with the contractors. The provisions of the Atomic Energy Act applicable to health and safety are discussed in relation to nuclear criticality safety. The use of United States Atomic Energy Commission manual chapters and Federal regulations is described. The functions of the USAEC Headquarters' offices and the operations offices are briefly outlined. Comments regarding the USAEC's inspection, auditing and appraisal programmes are included. Also briefly mentioned are the basic qualifications which must be met to become a contractor to possess and process or use fissionable materials. On the plant, factory or facility level the duties and responsibilities of industrial management are briefly outlined. The fundamental standards and their origin, together with the principal documents and guides are mentioned. The chief methods of control used by contractors operating large USAEC facilities and plants are described and compared. These include diagrams of how a typical nuclear criticality safety problem is handled from inception, design, construction and finally plant operation. Also included is a brief discussion of the contractors' methods of assuring strict employee compliance with the operating rules and limits. (author)

  10. A philosophy for space nuclear systems safety

    International Nuclear Information System (INIS)

    Marshall, A.C.

    1992-01-01

    The unique requirements and contraints of space nuclear systems require careful consideration in the development of a safety policy. The Nuclear Safety Policy Working Group (NSPWG) for the Space Exploration Initiative has proposed a hierarchical approach with safety policy at the top of the hierarchy. This policy allows safety requirements to be tailored to specific applications while still providing reassurance to regulators and the general public that the necessary measures have been taken to assure safe application of space nuclear systems. The safety policy used by the NSPWG is recommended for all space nuclear programs and missions

  11. Selecting Optimal Control Portfolios to Improve Army Aviation Safety

    National Research Council Canada - National Science Library

    Shelton, Sarah

    2001-01-01

    .... The Safety Center chartered the Aviation Safety Investment Strategy Team to evaluate accidents to determine their hazards, or contributing conditions, and their controls, or reduction measures...

  12. Automation for System Safety Analysis

    Science.gov (United States)

    Malin, Jane T.; Fleming, Land; Throop, David; Thronesbery, Carroll; Flores, Joshua; Bennett, Ted; Wennberg, Paul

    2009-01-01

    This presentation describes work to integrate a set of tools to support early model-based analysis of failures and hazards due to system-software interactions. The tools perform and assist analysts in the following tasks: 1) extract model parts from text for architecture and safety/hazard models; 2) combine the parts with library information to develop the models for visualization and analysis; 3) perform graph analysis and simulation to identify and evaluate possible paths from hazard sources to vulnerable entities and functions, in nominal and anomalous system-software configurations and scenarios; and 4) identify resulting candidate scenarios for software integration testing. There has been significant technical progress in model extraction from Orion program text sources, architecture model derivation (components and connections) and documentation of extraction sources. Models have been derived from Internal Interface Requirements Documents (IIRDs) and FMEA documents. Linguistic text processing is used to extract model parts and relationships, and the Aerospace Ontology also aids automated model development from the extracted information. Visualizations of these models assist analysts in requirements overview and in checking consistency and completeness.

  13. Probabilistic safety criteria at the safety function/system level

    International Nuclear Information System (INIS)

    1989-09-01

    A Technical Committee Meeting was held in Vienna, Austria, from 26-30 January 1987. The objectives of the meeting were: to review the national developments of PSC at the level of safety functions/systems including future trends; to analyse basic principles, assumptions, and objectives; to compare numerical values and the rationale for choosing them; to compile the experience with use of such PSC; to analyse the role of uncertainties in particular regarding procedures for showing compliance. The general objective of establishing PSC at the level of safety functions/systems is to provide a pragmatic tool to evaluate plant safety which is placing emphasis on the prevention principle. Such criteria could thus lead to a better understanding of the importance to safety of the various functions which have to be performed to ensure the safety of the plant, and the engineering means of performing these functions. They would reflect the state-of-the-art in modern PSAs and could contribute to a balance in system design. This report, prepared by the participants of the meeting, reviews the current status and future trends in the field and should assist Member States in developing their national approaches. The draft of this document was also submitted to INSAG to be considered in its work to prepare a document on safety principles for nuclear power plants. Five papers presented at the meeting are also included in this publication. A separate abstract was prepared for each of these papers. Refs, figs and tabs

  14. The Effect on Teenage Risky Driving of Feedback From a Safety Monitoring System: A Randomized Controlled Trial

    Science.gov (United States)

    Bingham, C. Raymond; Ouimet, Marie Claude; Pradhan, Anuj; Chen, Rusan; Barretto, Andrea; Shope, Jean

    2012-01-01

    Purpose Teenage risky driving may be due to teenagers not knowing what is risky, preferring risk, or the lack of consequences. Elevated gravitational-force (g-force) events, caused mainly by hard braking and sharp turns, provide a valid measure of risky driving and are the target of interventions using in-vehicle data recording and feedback devices. The effect of two forms of feedback about risky driving events to teenagers only or to teenagers and their parents was tested in a randomized controlled trial. Methods Ninety parent-teen dyads were randomized to one of two groups: (1) immediate feedback to teens (Lights Only); or (2) immediate feedback to teens plus family access to event videos and ranking of the teen relative to other teenage drivers (Lights Plus). Participants’ vehicles were instrumented with data recording devices and events exceeding 0.5 g were assessed for two weeks of baseline and 13 weeks of feedback. Results Growth analysis with random slopes yielded a significant decrease in event rates for the Lights Plus group (slope = −.11, p teenagers did not. Implications and Contribution Reducing elevated g-force events due to hard stops and sharp turns could reduce crash rates among novice teenage drivers. Using materials from the DriveCam For Families Program we found that feedback to both teens and parents significantly reduced rates, while feedback only to teens did not. PMID:23375825

  15. Adoption of digital safety protection system in Japan

    International Nuclear Information System (INIS)

    Ogiso, Z.

    1998-01-01

    The application of micro-processor-based digital controllers has been widely propagated among various industries in recent years. While in the nuclear power plant industry, the application of them has also been expanding gradually starting from non-safety related systems, taking advantage of their reliability and maintainability over the conventional analog devices. Based on the careful study of the feasibility of digital controllers to the safety protection system, the Tokyo Electric Power Company proposed on May 1989 the adoption of digital controllers to the safety protection system in the Application for Permission of Establishment of Kashiwazaki-Kariwa units 6 and 7 (ABWR-1350Mwe each). MITI, Ministry of International Trade and Industry, the Japanese regulatory body for electric power generating facilities, had approved this application after careful review. This paper describes a series of supporting activities leading to the MITI's approval of the digital safety protection system and the MITI's licensing activities. (author)

  16. Design requirements of communication architecture of SMART safety system

    International Nuclear Information System (INIS)

    Park, H. Y.; Kim, D. H.; Sin, Y. C.; Lee, J. Y.

    2001-01-01

    To develop the communication network architecture of safety system of SMART, the evaluation elements for reliability and performance factors are extracted from commercial networks and classified the required-level by importance. A predictable determinacy, status and fixed based architecture, separation and isolation from other systems, high reliability, verification and validation are introduced as the essential requirements of safety system communication network. Based on the suggested requirements, optical cable, star topology, synchronous transmission, point-to-point physical link, connection-oriented logical link, MAC (medium access control) with fixed allocation are selected as the design elements. The proposed architecture will be applied as basic communication network architecture of SMART safety system

  17. Expert systems and nuclear safety

    International Nuclear Information System (INIS)

    Beltracchi, L.

    1990-01-01

    The US Nuclear Regulatory Commission (NRC) and the Electric Power Research Institute have initiated a broad-based exploration of means to evaluate the potential applications of expert systems in the nuclear industry. This exploratory effort will assess the use of expert systems to augment the diagnostic and decision-making capabilities of personnel with the goal of enhancing productivity, reliability, and performance. The initial research effort is the development and documentation of guidelines for verifying and validating (V and V) expert systems. An initial application of expert systems in the nuclear industry is to aid operations and maintenance personnel in decision-making tasks. The scope of the decision aiding covers all types of cognitive behavior consisting of skill, rule, and knowledge-based behavior. For example, procedure trackers were designed and tested to support rule-based behavior. Further, these systems automate many of the tedious, error-prone human monitoring tasks, thereby reducing the potential for human error. The paper version of the procedure contains the knowledge base and the rules and thus serves as the basis of the design verification of the procedure tracker. Person-in-the-loop tests serve as the basis for the validation of a procedure tracker. When conducting validation tests, it is important to ascertain that the human retains the locus of control in the use of the expert system

  18. Safety Review related to Commercial Grade Digital Equipment in Safety System

    International Nuclear Information System (INIS)

    Yu, Yeongjin; Park, Hyunshin; Yu, Yeongjin; Lee, Jaeheung

    2013-01-01

    The upgrades or replacement of I and C systems on safety system typically involve digital equipment developed in accordance with non-nuclear standards. However, the use of commercial grade digital equipment could include the vulnerability for software common-mode failure, electromagnetic interference and unanticipated problems. Although guidelines and standards for dedication methods of commercial grade digital equipment are provided, there are some difficulties to apply the methods to commercial grade digital equipment for safety system. This paper focuses on regulatory guidelines and relevant documents for commercial grade digital equipment and presents safety review experiences related to commercial grade digital equipment in safety system. This paper focuses on KINS regulatory guides and relevant documents for dedication of commercial grade digital equipment and presents safety review experiences related to commercial grade digital equipment in safety system. Dedication including critical characteristics is required to use the commercial grade digital equipment on safety system in accordance with KEPIC ENB 6370 and EPRI TR-106439. The dedication process should be controlled in a configuration management process. Appropriate methods, criteria and evaluation result should be provided to verify acceptability of the commercial digital equipment used for safety function

  19. Regulatory Control of Radiation Sources. Safety Guide (Arabic Edition)

    International Nuclear Information System (INIS)

    2012-01-01

    This Safety Guide is intended to assist States in implementing the requirements established in Safety Standards Series No. GS-R-1, Legal and Governmental Infrastructure for Nuclear, Radiation, Radioactive Waste and Transport Safety, for a national regulatory infrastructure to regulate any practice involving radiation sources in medicine, industry, research, agriculture and education. The Safety Guide provides advice on the legislative basis for establishing regulatory bodies, including the effective independence of the regulatory body. It also provides guidance on implementing the functions and activities of regulatory bodies: the development of regulations and guides on radiation safety; implementation of a system for notification and authorization; carrying out regulatory inspections; taking necessary enforcement actions; and investigating accidents and circumstances potentially giving rise to accidents. The various aspects relating to the regulatory control of consumer products are explained, including justification, optimization of exposure, safety assessment and authorization. Guidance is also provided on the organization and staffing of regulatory bodies. Contents: 1. Introduction; 2. Legal framework for a regulatory infrastructure; 3. Principal functions and activities of the regulatory body; 4. Regulatory control of the supply of consumer products; 5. Functions of the regulatory body shared with other governmental agencies; 6. Organization and staffing of the regulatory body; 7. Documentation of the functions and activities of the regulatory body; 8. Support services; 9. Quality management for the regulatory system.

  20. Meeting the maglev system's safety requirements

    Energy Technology Data Exchange (ETDEWEB)

    Pierick, K

    1983-12-01

    The author shows how the safety requirements of the maglev track system derive from the general legal conditions for the safety of tracked transport. It is described how their compliance beyond the so-called ''development-accompanying'' and ''acceptance-preparatory'' safety work can be assured for the Transrapid test layout (TVE) now building in Emsland and also for later application as public transport system in Germany within the meaning of the General Railway Act.

  1. Control of Nuclear Materials and Special Equipment (Nuclear Safety Regulations)

    International Nuclear Information System (INIS)

    Cizmek, A.; Prah, M.; Medakovic, S.; Ilijas, B.

    2008-01-01

    Based on Nuclear Safety Act (OG 173/03) the State Office for Nuclear Safety (SONS) in 2008 adopted beside Ordinance on performing nuclear activities (OG 74/06) and Ordinance on special conditions for individual activities to be performed by expert organizations which perform activities in the area of nuclear safety (OG 74/06) the new Ordinance on the control of nuclear material and special equipment (OG 15/08). Ordinance on the control of nuclear material and special equipment lays down the list of nuclear materials and special equipment as well as of nuclear activities covered by the system of control of production of special equipment and non-nuclear material, the procedure for notifying the intention to and filing the application for a license to carry out nuclear activities, and the format and contents of the forms for doing so. This Ordinance also lays down the manner in which nuclear material records have to be kept, the procedure for notifying the State administration organization (regulatory body) responsible for nuclear safety by the nuclear material user, and the keeping of registers of nuclear activities, nuclear material and special equipment by the State administration organization (regulatory body) responsible for nuclear safety, as well as the form and content of official nuclear safety inspector identification card and badge.(author)

  2. The Evolution of System Safety at NASA

    Science.gov (United States)

    Dezfuli, Homayoon; Everett, Chris; Groen, Frank

    2014-01-01

    The NASA system safety framework is in the process of change, motivated by the desire to promote an objectives-driven approach to system safety that explicitly focuses system safety efforts on system-level safety performance, and serves to unify, in a purposeful manner, safety-related activities that otherwise might be done in a way that results in gaps, redundancies, or unnecessary work. An objectives-driven approach to system safety affords more flexibility to determine, on a system-specific basis, the means by which adequate safety is achieved and verified. Such flexibility and efficiency is becoming increasingly important in the face of evolving engineering modalities and acquisition models, where, for example, NASA will increasingly rely on commercial providers for transportation services to low-earth orbit. A key element of this objectives-driven approach is the use of the risk-informed safety case (RISC): a structured argument, supported by a body of evidence, that provides a compelling, comprehensible and valid case that a system is or will be adequately safe for a given application in a given environment. The RISC addresses each of the objectives defined for the system, providing a rational basis for making informed risk acceptance decisions at relevant decision points in the system life cycle.

  3. Multilayer robust control for safety enhancement of reactor operations

    International Nuclear Information System (INIS)

    Edwards, R.M.; Lee, K.Y.; Ray, A.

    1991-01-01

    A novel concept of reactor power and temperature control has been recently reported in which a conventional output feedback controller is embedded within a state feedback setting. The embedded output feedback controller at the inner layer largely compensates for plant modeling uncertainties and external disturbances, and the outer layer generates an optimal control signal via feedback of the estimated plant states. A major advantage of this embedded architecture is the robustness of the control system relative to parametric and nonparametric uncertainties and thus the opportunity for designing fault-accommodating control algorithms to improve reactor operations and plant safety. The paper illustrates the architecture of the state-feedback-assisted classical (SFAC) control, which utilizes an embedded output feedback controller designed via classical techniques. It demonstrates the difference between the performance of conventional state feedback control and SFAC by examining the sensitivity of the dominant eigenvalues of the individual closed-loop systems

  4. Ergonomics in the context of system safety

    International Nuclear Information System (INIS)

    Donnelly, K.E.

    1984-01-01

    In a complex industrial environment, ergonomics must be combined with management science and systems analysis to produce a program which can create effective change and improve safety performance. We give an overview of such an approach, namely System Safety, so that its ergonomic content may be seen

  5. Reconstruction of instrumentation and control system (SKR)

    International Nuclear Information System (INIS)

    Wiening, K.-H.

    2001-01-01

    For the first time extensive upgrades have been performed in all safety related areas of units with WWER 440/230 reactors. One of the most important actions was the replacement of the safety and safety related instrumentation and control. The state of the art digital safety instrumentation and control system TELEPERM XS has been implemented in units 1 and 2 of the Bohunice V1 power plant. The requirements as deduced from safety assessments conducted by commissions of international experts have been fulfilled, so that Bohunice V1 after this gradual reconstruction has been upgraded to an internationally accepted safety level for the remainder of its service life. (author)

  6. Systems Thinking and Patient Safety

    National Research Council Canada - National Science Library

    Schyve, Paul M

    2005-01-01

    Patient safety is a prominent theme in health care delivery today. This should come as no surprise, given that "first, do no harm" has been the ethical watchword throughout the history of medicine, nursing, and pharmacy...

  7. Neutron flux control systems validation

    International Nuclear Information System (INIS)

    Hascik, R.

    2003-01-01

    In nuclear installations main requirement is to obtain corresponding nuclear safety in all operation conditions. From the nuclear safety point of view is commissioning and start-up after reactor refuelling appropriate period for safety systems verification. In this paper, methodology, performance and results of neutron flux measurements systems validation is presented. Standard neutron flux measuring chains incorporated into the reactor protection and control system are used. Standard neutron flux measuring chain contains detector, preamplifier, wiring to data acquisition unit, data acquisition unit, wiring to display at control room and display at control room. During reactor outage only data acquisition unit and wiring and displaying at reactor control room is verified. It is impossible to verify detector, preamplifier and wiring to data acquisition recording unit during reactor refuelling according to low power. Adjustment and accurate functionality of these chains is confirmed by start-up rate (SUR) measurement during start-up tests after refuelling of the reactors. This measurement has direct impact to nuclear safety and increase operational nuclear safety level. Briefly description of each measuring system is given. Results are illustrated on measurements performed at Bohunice NPP during reactor start-up tests. Main failures and their elimination are described (Authors)

  8. 76 FR 49532 - Federal Motor Vehicle Safety Standards; Electronic Stability Control; Technical Report on the...

    Science.gov (United States)

    2011-08-10

    ...-0112] Federal Motor Vehicle Safety Standards; Electronic Stability Control; Technical Report on the Effectiveness of Electronic Stability Control Systems for Cars and LTVs AGENCY: National Highway Traffic Safety..., Electronic Stability Control Systems. The report's title is: Crash Prevention Effectiveness in Light-Vehicle...

  9. 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)

  10. Safety management systems and their role in achieving high standards of operational safety

    International Nuclear Information System (INIS)

    Coulston, D.J.; Baylis, C.C.

    2000-01-01

    Achieving high standards of operational safety requires a robust management framework that is visible to all personnel with responsibility for its implementation. The structure of the management framework must ensure that all processes used to manage safety interlink in a logical and coherent manner, that is, they form a management system that leads to continuous improvement in safety performance. This Paper describes BNFL's safety management system (SMS). The SMS has management processes grouped within 5 main elements: 1. Policy, 2. Organisation, 3. Planning and Implementation, 4. Measuring and Reviewing Performance, 5. Audit. These elements reflect the overall process of setting safety objective (from Policy), measuring success and reviewing the performance. Effective implementation of the SMS requires senior managers to demonstrate leadership through their commitment and accountability. However, the SMS as a whole reflects that every employee at every level within BNFL is responsible for safety of operations under their control. The SMS therefore promotes a proactive safety culture and safe operations. The system is formally documented in the Company's Environmental, Health and Safety (EHS) Manual. Within in BNFL Group, the Company structures enables the Manual to provide overall SMS guidance and co-ordination to its range of nuclear businesses. Each business develops the SMS to be appropriate at all levels of its organisation, but ensuring that each level is consistent with the higher level. The Paper concludes with a summary of BNFL's safety performance. (author)

  11. Safety-critical Java for embedded systems

    DEFF Research Database (Denmark)

    Schoeberl, Martin; Dalsgaard, Andreas Engelbredt; Hansen, René Rydhof

    2016-01-01

    This paper presents the motivation for and outcomes of an engineering research project on certifiable Javafor embedded systems. The project supports the upcoming standard for safety-critical Java, which defines asubset of Java and libraries aiming for development of high criticality systems....... The outcome of this projectinclude prototype safety-critical Java implementations, a time-predictable Java processor, analysis tools formemory safety, and example applications to explore the usability of safety-critical Java for this applicationarea. The text summarizes developments and key contributions...

  12. SBO simulations for Integrated Passive Safety System (IPSS) using MARS

    International Nuclear Information System (INIS)

    Kim, Sang Ho; Jeong, Sung Yeop; Chang, Soon Heung

    2012-01-01

    The current nuclear power plants have lots of active safety systems with some passive safety systems. The safety of current and future nuclear power plants can be enhanced by the application of additional passive safety systems for the ultimate safety. It is helpful to install the passive safety systems on current nuclear power plants without the design change for the licensibility. For solving the problem about the system complexity shown in the Fukushima accidents, the current nuclear power plants are needed to be enhanced by an additional integrated and simplified system. As a previous research, the integrated passive safety system (IPSS) was proposed to solve the safety issues related with the decay heat removal, containment integrity and radiation release. It could be operated by natural phenomena like gravity, natural circulation and pressure difference without AC power. The five main functions of IPSS are: (a) Passive decay heat removal, (b) Passive emergency core cooling, (c) Passive containment cooling, (d) Passive in vessel retention and ex-vessel cooling, and (e) Filtered venting and pressure control. The purpose of this research is to analyze the performances of each function by using MARS code. The simulated accident scenarios were station black out (SBO) and the additional accidents accompanied by SBO

  13. Operation safety of complex industrial systems

    International Nuclear Information System (INIS)

    Zwingelstein, G.

    1999-01-01

    Zero fault or zero risk is an unreachable goal in industrial activities like nuclear activities. However, methods and techniques exist to reduce the risks to the lowest possible and acceptable level. The operation safety consists in the recognition, evaluation, prediction, measurement and mastery of technological and human faults. This paper analyses each of these points successively: 1 - evolution of operation safety; 2 - definitions and basic concepts: failure, missions and functions of a system and of its components, basic concepts and operation safety; 3 - forecasting analysis of operation safety: reliability data, data-banks, precautions for the use of experience feedback data; realization of an operation safety study: management of operation safety, quality assurance, critical review and audit of operation safety studies; 6 - conclusions. (J.S.)

  14. UML-based testcase- and system model for railway process control- and safety engineering; UML-basierte Testfall- und Systemmodelle fuer die Eisenbahnleit- und -sicherungstechnik

    Energy Technology Data Exchange (ETDEWEB)

    Knollmann, V.

    2007-12-15

    The continuous introduction of more sophisticated and more complex automation systems is of high relevance for operations- and control devices for railway applications. One example of this trend is the European Train Control System (ETCS). I is harmonized for all European member states and currently deployed for commercial use on selected lines. Applications like ETCS take responsibility for passengers, environment and the rolling stock material. For this reason, the development of those applications is characterized by high demands on testing, documentation and quality assurance. Unfortunately, real projects have proven to achieve these high demands only with tremendous efforts as long as they stick to their traditional development processes and methods. Therefore, a wide spread exploration for new concepts regarding planning, implementation and - most of all - testing of safety relevant systems is under its way. This thesis contributes to the search for new design methods for such systems. It combines recent achievements in the environment of the Unified Modeling Language (UML) to an integrated approach for a consistent testcase- and system description in a single UML model. The model comprises the UML profiles SysML (System Modeling Language) and U2TP (UML 2 Testing Profile), which are officially specified and released by the Object Management Group (OMG). This thesis shows how based on these two profiles - a single, combined system- and testcase model can be build - relationships between requirements, implementation- and testcase can be established - requirement relationships can be evaluated automatically to ensure model consistency and to determine implementation and testcase coverage - C++-code can be generated directly from the model - TTCN-3-code (Testing and Test Control Notation, Version 3) can be derived automatically from the model - a toolchain can be set up which supports all mentioned functions. In order to prove the feasibility of this approach, a

  15. Systematic evaluation program review of NRC Safety Topic VI-7.3 associated with the electrical, instrumentation and control portions of the ECCS actuation system for the Dresden II Nuclear Power Plant

    International Nuclear Information System (INIS)

    St Leger-Barter, G.

    1980-11-01

    This report documents the technical evaluation and review of NRC Safety Topic VI-7.A.3, associated with the electrical, instrumentation, and control portions of the classification of the ECCS actuation system for the Dresden II nuclear power plant, using current licensing criteria

  16. Safety implications of computerized process control in nuclear power plants

    International Nuclear Information System (INIS)

    1991-02-01

    Modern nuclear power plants are making increasing use of computerized process control because of the number of potential benefits that accrue. This practice not only applies to new plants but also to those in operation. Here, the replacement of both conventional process control systems and outdated computerized systems is seen to be of benefit. Whilst this contribution is obviously of great importance to the viability of nuclear electricity generation, it must be recognized that there are major safety concerns in taking this route. However, there is the potential for enhancing the safety of nuclear power plants if the full power of microcomputers and the associated electronics is applied correctly through well designed, engineered, installed and maintained systems. It is essential that areas where safety can be improved be identified and that the pitfalls are clearly marked so that they can be avoided. The deliberations of this Technical Committee Meeting are a step on the road to this goal of improved safety through computerized process control. This report also contains the papers presented at the technical committee meeting by participants. A separate abstract was prepared for each of these 15 presentations. Refs, figs and tabs

  17. Manufacture of Platform Prototype for Digital Safety System

    International Nuclear Information System (INIS)

    Lee, S. Y.; Kim, J. S.; Kim, J. M.

    2010-01-01

    Unit controller is a basic unit of digital safety system platform prototype. The typical unit controller is comprised of CPB(CPU board), CMB(communication board), AIB(Analog input board), AOB(Analog output board), CIB(contact input board), COB(contact output board), and a subrack. It is developed according to H/W development procedure and S/W development life cycle. A digital safety system(for example, plant protection system) is the assemblies of unit controllers. CPB performs the function of each system. DSP(digital signal processor) is built in CPB. CMB is responsible for communication between unit controllers. NSD(Network Switching Device) exchanges data between the unit controllers. Each unit controller of the platform are connected to NSD through CMB. Reliability analyses on unit controller and NSD are performed. These reliability data are used as input of technical validation

  18. Safety features of subcritical fluid fueled systems

    International Nuclear Information System (INIS)

    Bell, C.R.

    1995-01-01

    Accelerator-driven transmutation technology has been under study at Los Alamos for several years for application to nuclear waste treatment, tritium production, energy generation, and recently, to the disposition of excess weapons plutonium. Studies and evaluations performed to date at Los Alamos have led to a current focus on a fluid-fuel, fission system operating in a neutron source-supported subcritical mode, using molten salt reactor technology and accelerator-driven proton-neutron spallation. In this paper, the safety features and characteristics of such systems are explored from the perspective of the fundamental nuclear safety objectives that any reactor-type system should address. This exploration is qualitative in nature and uses current vintage solid-fueled reactors as a baseline for comparison. Based on the safety perspectives presented, such systems should be capable of meeting the fundamental nuclear safety objectives. In addition, they should be able to provide the safety robustness desired for advanced reactors. However, the manner in which safety objectives and robustness are achieved is very different from that associated with conventional reactors. Also, there are a number of safety design and operational challenges that will have to be addressed for the safety potential of such systems to be credible

  19. Safety features of subcritical fluid fueled systems

    International Nuclear Information System (INIS)

    Bell, C.R.

    1994-01-01

    Accelerator-driven transmutation technology has been under study at Los Alamos for several years for application to nuclear waste treatment, tritium production, energy generation, and recently, to the disposition of excess weapons plutonium. Studies and evaluations performed to date at Los Alamos have led to a current focus on a fluid-fuel, fission system operating in a neutron source-supported subcritical mode, using molten salt reactor technology and accelerator-driven proton-neutron spallation. In this paper, the safety features and characteristics of such systems are explored from the perspective of the fundamental nuclear safety objectives that any reactor-type system should address. This exploration is qualitative in nature and uses current vintage solid-fueled reactors as a baseline for comparison. Based on the safety perspectives presented, such systems should be capable of meeting the fundamental nuclear safety objectives. In addition, they should be able to provide the safety robustness desired for advanced reactors. However, the manner in which safety objectives and robustness are achieved in very different from that associated with conventional reactors. Also, there are a number of safety design and operational challenges that will have to be addressed for the safety potential of such systems to be credible

  20. Safety features of subcritical fluid fueled systems

    Energy Technology Data Exchange (ETDEWEB)

    Bell, C.R. [Los Alamos National Laboratory, NM (United States)

    1995-10-01

    Accelerator-driven transmutation technology has been under study at Los Alamos for several years for application to nuclear waste treatment, tritium production, energy generation, and recently, to the disposition of excess weapons plutonium. Studies and evaluations performed to date at Los Alamos have led to a current focus on a fluid-fuel, fission system operating in a neutron source-supported subcritical mode, using molten salt reactor technology and accelerator-driven proton-neutron spallation. In this paper, the safety features and characteristics of such systems are explored from the perspective of the fundamental nuclear safety objectives that any reactor-type system should address. This exploration is qualitative in nature and uses current vintage solid-fueled reactors as a baseline for comparison. Based on the safety perspectives presented, such systems should be capable of meeting the fundamental nuclear safety objectives. In addition, they should be able to provide the safety robustness desired for advanced reactors. However, the manner in which safety objectives and robustness are achieved is very different from that associated with conventional reactors. Also, there are a number of safety design and operational challenges that will have to be addressed for the safety potential of such systems to be credible.

  1. Integrating system safety into the basic systems engineering process

    Science.gov (United States)

    Griswold, J. W.

    1971-01-01

    The basic elements of a systems engineering process are given along with a detailed description of what the safety system requires from the systems engineering process. Also discussed is the safety that the system provides to other subfunctions of systems engineering.

  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. Efficacy and safety of solifenacin plus tamsulosin oral controlled absorption system in men with lower urinary tract symptoms: a meta-analysis

    Directory of Open Access Journals (Sweden)

    Ming-Chao Li

    2015-02-01

    Full Text Available We performed a meta-analysis to compare treatment with a combination of solifenacin plus tamsulosin oral controlled absorption system (TOCAS with placebo or TOCAS monotherapy. The aim of the meta-analysis was to clarify the efficacy and safety of the combination treatments method for lower urinary tract symptoms (LUTS. We searched for trials of men with LUTS that were randomized to combination treatment compared with TOCAS monotherapy or placebo. We pooled data from three placebo-controlled trials meeting inclusion criteria. Primary outcomes of interest included changes in International Prostate Symptom Score (IPSS and urinary frequency. We also assessed postvoid residual, maximum urinary flow rate, incidence of urinary retention (UR, adverse events. Data were pooled using random or fixed effect models for continuous outcomes and the Mantel-Haenszel method to generate risk ratio. Reductions in IPSS storage subscore and total urgency and frequency score (TUFS were observed with solifenacin 6 mg plus TOCAS compared with placebo (P< 0.0001 and P< 0.0001, respectively. Reductions in IPSS storage subscore and TUFS were observed with solifenacin 9 mg plus TOCAS compared with placebo (P = 0.003 and P= 0.0006, respectively. Reductions in TUFS was observed with solifenacin 6 mg plus TOCAS compared with TOCAS (P = 0.01. Both combination treatments were well tolerated, with low incidence of UR. Solifenacin 6 mg plus TOCAS significantly improved total IPSS, storage and voiding symptoms compared with placebo. Solifenacin 6 mg plus TOCAS also improved storage symptoms compared with TOCAS alone. There was no additional benefit of solifenacin 9 mg compared with 6 mg when used in combination with TOCAS.

  4. FDR (drive-dynamics-control) - a new driving safety system with active control of brake and drive forces in the dynamic fringe range; FDR, ein neues Fahrsicherheitssystem mit aktiver Regelung der Brems- und Antriebskraefte im fahrdynamischen Grenzbereich

    Energy Technology Data Exchange (ETDEWEB)

    Erhardt, R. [Bosch (R.) GmbH, Stuttgart (Germany); Zanten, A.T. van [Bosch (R.) GmbH, Stuttgart (Germany)

    1995-12-31

    BOSCH is going to introduce a new driving safety system in 1995, the FDR (drive-dynamics-control). Using the measured and estimated dynamic magnitudes as a basis, the system calculates inhowfar the actual vehicle motion differs from the desired stable trace- and direction-consistent handling properties. Depending on the driving situation and driver`s wishes the braking and driving forces at the wheels are adjusted with a considerable divergence in order to achieve the desired handling properties. The system improves the driving stability in all operating states as soon as the dynamic limiting range is reached. It even reduces the risk of skidding in case of extreme steering manoeuvres and also enables the safe control of the vehicle in critical traffic situations. Furthermore the system offers improved basic anti-skid braking system and anti-slip control functions. Due to these advantages it can be expected that the FDR is going to make an important contribution to avoiding accidents and reducing damage. (orig.) [Deutsch] Mit FDR (Fahr-Dynamik-Regelung) wird BOSCH 1995 ein neues Fahrsicherheitssystem einfuehren. Das System berechnet auf der Basis gemessener und geschaetzter fahrdynamischer Groessen, wie stark die tatsaechliche Fahrzeugbewegung von einem gewuenschten stabilen, spur- und richtungstreuen Fahrverhalten abweicht. Die Brems- und Antriebskraefte an den Raedern werden bei deutlicher Abweichung abhaengig von Fahrsituation und Fahrerwunsch so eingestellt, dass die Abweichung minimiert und das gewuenschte Fahrverhalten weitgehend erreicht wird. Das System verbessert die Fahrstabilitaet in allen Betriebszustaenden, sobald der fahrdynamische Grenzbereich erreicht wird. Es reduziert selbst bei extremen Lenkmanoevern die Schleudergefahr drastisch und ermoeglicht auch in kritischen Verkehrssituationen die sicherere Beherrschung des Fahrzeugs. Darueberhinaus bietet das System verbesserte ABS- und ASR-Grundfunktionen. Diese Vorteile lassen erwarten, dass FDR einen

  5. Analysis of Aviation Safety Reporting System Incident Data Associated With the Technical Challenges of the Vehicle Systems Safety Technology Project

    Science.gov (United States)

    Withrow, Colleen A.; Reveley, Mary S.

    2014-01-01

    This analysis was conducted to support the Vehicle Systems Safety Technology (VSST) Project of the Aviation Safety Program (AVsP) milestone VSST4.2.1.01, "Identification of VSST-Related Trends." In particular, this is a review of incident data from the NASA Aviation Safety Reporting System (ASRS). The following three VSST-related technical challenges (TCs) were the focus of the incidents searched in the ASRS database: (1) Vechicle health assurance, (2) Effective crew-system interactions and decisions in all conditions; and (3) Aircraft loss of control prevention, mitigation, and recovery.

  6. Safety management system needs assessment.

    Science.gov (United States)

    2016-04-01

    The safety of the traveling public is critical as each year there are approximately 200 highway fatalities in Nebraska and numerous crash injuries. The objective of this research was to conduct a needs assessment to identify the requirements of a sta...

  7. Safety standards of IAEA for management systems

    International Nuclear Information System (INIS)

    Vincze, P.

    2005-01-01

    IAEA has developed a new series of safety standards which are assigned for constitution of the conditions and which give the instruction for setting up the management systems that integrate the aims of safety, health, life environment and quality. The new standard shall replace IAEA 50-C-Q - Requirements for security of the quality for safety in nuclear power plants and other nuclear facilities as well as 14 related safety instructions mentioned in the Safety series No. 50-C/SG-Q (1996). When developing of this complex, integrated set of requirements for management systems, the IAEA requirements 50-C-Q (1996) were taken into consideration as well as the publications developed within the International organisation for standardization (ISO) ISO 9001:2000 and ISO14001: 1996. The experience of European Union member states during the development, implementation and improvement of the management systems were also taken into consideration

  8. Reliability analysis of Angra I safety systems

    International Nuclear Information System (INIS)

    Oliveira, L.F.S. de; Soto, J.B.; Maciel, C.C.; Gibelli, S.M.O.; Fleming, P.V.; Arrieta, L.A.

    1980-07-01

    An extensive reliability analysis of some safety systems of Angra I, are presented. The fault tree technique, which has been successfully used in most reliability studies of nuclear safety systems performed to date is employed. Results of a quantitative determination of the unvailability of the accumulator and the containment spray injection systems are presented. These results are also compared to those reported in WASH-1400. (E.G.) [pt

  9. The detector safety system for LHC experiments

    CERN Document Server

    Schmeling, Sascha; Lüders, S; Morpurgo, Giulio

    2004-01-01

    The Detector Safety System (DSS), currently being developed at CERN under the auspices of the Joint Controls Project (JCOP), will be responsible for assuring the protection of equipment for the four Large Hadron Collider (LHC)**1 experiments. Thus, the DSS will require a high degree of both availability and reliability. After evaluation of various possible solutions, a prototype is being built based on a redundant Siemens PLC**2 front-end, to which the safety- critical part of the DSS task is delegated. This is then supervised by a PVSS**3 SCADA**4 system via an OPC**5 server. The PLC front-end is capable of running autonomously and of automatically taking predefined protective actions whenever required. The supervisory layer provides the operator with a status display and with limited online reconfiguration capabilities. Configuration of the code running in the PLCs will be completely data driven via the contents of a "configuration database." Thus, the DSS can easily adapt to the different and constantly ev...

  10. NASA Aviation Safety Reporting System (ASRS)

    Science.gov (United States)

    Connell, Linda J.

    2017-01-01

    The NASA Aviation Safety Reporting System (ASRS) collects, analyzes, and distributes de-identified safety information provided through confidentially submitted reports from frontline aviation personnel. Since its inception in 1976, the ASRS has collected over 1.4 million reports and has never breached the identity of the people sharing their information about events or safety issues. From this volume of data, the ASRS has released over 6,000 aviation safety alerts concerning potential hazards and safety concerns. The ASRS processes these reports, evaluates the information, and provides selected de-identified report information through the online ASRS Database at http:asrs.arc.nasa.gov. The NASA ASRS is also a founding member of the International Confidential Aviation Safety Systems (ICASS) group which is a collection of other national aviation reporting systems throughout the world. The ASRS model has also been replicated for application to improving safety in railroad, medical, fire fighting, and other domains. This presentation will discuss confidential, voluntary, and non-punitive reporting systems and their advantages in providing information for safety improvements.

  11. Safety classification of nuclear power plant systems, structures and components

    International Nuclear Information System (INIS)

    1992-01-01

    The Safety Classification principles used for the systems, structures and components of a nuclear power plant are detailed in the guide. For classification, the nuclear power plant is divided into structural and operational units called systems. Every structure and component under control is included into some system. The Safety Classes are 1, 2 and 3 and the Class EYT (non-nuclear). Instructions how to assign each system, structure and component to an appropriate safety class are given in the guide. The guide applies to new nuclear power plants and to the safety classification of systems, structures and components designed for the refitting of old nuclear power plants. The classification principles and procedures applying to the classification document are also given

  12. Survey of electronic safety systems in accelerator applications

    International Nuclear Information System (INIS)

    Mahoney, K.

    1997-01-01

    This paper presents the preliminary results and analysis of a comprehensive survey of the implementation of accelerator safety interlock systems from over 30 international labs. At the present time there is not a self consistent means to evaluate both the experiences and level of protection provided by electronic safety interlock systems. This research is intended to analyze the strength and weaknesses of several different types of interlock system implementation methodologies. Research, medical, and industrial accelerators are compared. Thomas Jefferson National Accelerator Facility (TJNAF) was one of the first large particle accelerators to implement a safety interlock system using programmable logic controllers. Since that time all of the major new U.S. accelerator construction projects plan to use some form of programmable electronics as part of a safety interlock system in some capacity

  13. Access safety systems - New concepts from the LHC experience

    International Nuclear Information System (INIS)

    Ladzinski, T.; Delamare, C.; Luca, S. di; Hakulinen, T.; Hammouti, L.; Havart, F.; Juget, J.F.; Ninin, P.; Nunes, R.; Riesco, T.; Sanchez-Corral Mena, E.; Valentini, F.

    2012-01-01

    The LHC Access Safety System has introduced a number of new concepts into the domain of personnel protection at CERN. These can be grouped into several categories: organisational, architectural and concerning the end-user experience. By anchoring the project on the solid foundations of the IEC 61508/61511 methodology, the CERN team and its contractors managed to design, develop, test and commission on time a SIL3 safety system. The system uses a successful combination of the latest Siemens redundant safety programmable logic controllers with a traditional relay logic hard wired loop. The external envelope barriers used in the LHC include personnel and material access devices, which are interlocked door-booths introducing increased automation of individual access control, thus removing the strain from the operators. These devices ensure the inviolability of the controlled zones by users not holding the required credentials. To this end they are equipped with personnel presence detectors and the access control includes a state of the art bio-metry check. Building on the LHC experience, new projects targeting the refurbishment of the existing access safety infrastructure in the injector chain have started. This paper summarises the new concepts introduced in the LHC access control and safety systems, discusses the return of experience and outlines the main guiding principles for the renewal stage of the personnel protection systems in the LHC injector chain in a homogeneous manner. (authors)

  14. Safety status system for operating room devices.

    Science.gov (United States)

    Guédon, Annetje C P; Wauben, Linda S G L; Overvelde, Marlies; Blok, Joleen H; van der Elst, Maarten; Dankelman, Jenny; van den Dobbelsteen, John J

    2014-01-01

    Since the increase of the number of technological aids in the operating room (OR), equipment-related incidents have come to be a common kind of adverse events. This underlines the importance of adequate equipment management to improve the safety in the OR. A system was developed to monitor the safety status (periodic maintenance and registered malfunctions) of OR devices and to facilitate the notification of malfunctions. The objective was to assess whether the system is suitable for use in an busy OR setting and to analyse its effect on the notification of malfunctions. The system checks automatically the safety status of OR devices through constant communication with the technical facility management system, informs the OR staff real-time and facilitates notification of malfunctions. The system was tested for a pilot period of six months in four ORs of a Dutch teaching hospital and 17 users were interviewed on the usability of the system. The users provided positive feedback on the usability. For 86.6% of total time, the localisation of OR devices was accurate. 62 malfunctions of OR devices were reported, an increase of 12 notifications compared to the previous year. The safety status system was suitable for an OR complex, both from a usability and technical point of view, and an increase of reported malfunctions was observed. The system eases monitoring the safety status of equipment and is a promising tool to improve the safety related to OR devices.

  15. Traction Control System for Motorcycles

    Directory of Open Access Journals (Sweden)

    Cardinale Pascal

    2009-01-01

    Full Text Available Traction control is a widely used control system to increase stability and safety of four wheel vehicles. Automatic stability control is used in the BMW K1200R motorcycle and in motoGP competition, but not in other motorcycles. This paper presents an algorithm and a low-cost real-time hardware implementation for motorcycles. A prototype has been developed, applied on a commercial motorcycle, and tested in a real track. The control system that can be tuned by the driver during the race has been appreciated by the test driver.

  16. A formal safety analysis for PLC software-based safety critical system using Z

    International Nuclear Information System (INIS)

    Koh, Jung Soo; Seong, Poong Hyun

    1997-01-01

    This paper describes a formal safety analysis technique which is demonstrated by performing empirical formal safety analysis with the case study of beamline hutch door Interlock system that is developed by using PLC (Programmable Logic Controller) systems at the Pohang Accelerator Laboratory. In order to perform formed safety analysis, we have built the Z formal specifications representation from user requirement written in ambiguous natural language and target PLC ladder logic, respectively. We have also studied the effective method to express typical PLC timer component by using specific Z formal notation which is supported by temporal history. We present a formal proof technique specifying and verifying that the hazardous states are not introduced into ladder logic in the PLC-based safety critical system

  17. Automatic control systems engineering

    International Nuclear Information System (INIS)

    Shin, Yun Gi

    2004-01-01

    This book gives descriptions of automatic control for electrical electronics, which indicates history of automatic control, Laplace transform, block diagram and signal flow diagram, electrometer, linearization of system, space of situation, state space analysis of electric system, sensor, hydro controlling system, stability, time response of linear dynamic system, conception of root locus, procedure to draw root locus, frequency response, and design of control system.

  18. Proactive Management of Aviation System Safety Risk

    Data.gov (United States)

    National Aeronautics and Space Administration — Aviation safety systems have undergone dramatic changes over the past fifty years. If you take a look at the early technology in this area, you'll see that there was...

  19. Safety considerations for compressed hydrogen storage systems

    International Nuclear Information System (INIS)

    Gleason, D.

    2006-01-01

    An overview of the safety considerations for various hydrogen storage options, including stationary, vehicle storage, and mobile refueling technologies. Indications of some of the challenges facing the industry as the demand for hydrogen fuel storage systems increases. (author)

  20. Controls and Machine Protection Systems

    CERN Document Server

    Carrone, E.

    2016-01-01

    Machine protection, as part of accelerator control systems, can be managed with a 'functional safety' approach, which takes into account product life cycle, processes, quality, industrial standards and cybersafety. This paper will discuss strategies to manage such complexity and the related risks, with particular attention to fail-safe design and safety integrity levels, software and hardware standards, testing, and verification philosophy. It will also discuss an implementation of a machine protection system at the SLAC National Accelerator Laboratory's Linac Coherent Light Source (LCLS).

  1. COMPRESS - a computerized reactor safety system

    International Nuclear Information System (INIS)

    Vegh, E.

    1986-01-01

    The computerized reactor safety system, called COMPRESS, provides the following services: scram initiation; safety interlockings; event recording. The paper describes the architecture of the system and deals with reliability problems. A self-testing unit checks permanently the correct operation of the independent decision units. Moreover the decision units are tested by short pulses whether they can initiate a scram. The self-testing is described in detail

  2. Safety Justification and Safety Case for Safety-critical Software in Digital Reactor Protection System

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Kee-Choon; Lee, Jang-Soo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Jee, Eunkyoung [KAIST, Daejeon (Korea, Republic of)

    2016-10-15

    Nuclear safety-critical software is under strict regulatory requirements and these regulatory requirements are essential for ensuring the safety of nuclear power plants. The verification & validation (V and V) and hazard analysis of the safety-critical software are required to follow regulatory requirements through the entire software life cycle. In order to obtain a license from the regulatory body through the development and validation of safety-critical software, it is essential to meet the standards which are required by the regulatory body throughout the software development process. Generally, large amounts of documents, which demonstrate safety justification including standard compliance, V and V, hazard analysis, and vulnerability assessment activities, are submitted to the regulatory body during the licensing process. It is not easy to accurately read and evaluate the whole documentation for the development activities, implementation technology, and validation activities. The safety case methodology has been kwon a promising approach to evaluate the level and depth of the development and validation results. A safety case is a structured argument, supported by a body of evidence that provides a compelling, comprehensible, and valid case that a system is safe for a given application in a given operating environment. It is suggested to evaluate the level and depth of the results of development and validation by applying safety case methodology to achieve software safety demonstration. A lot of documents provided as evidence are connected to claim that corresponds to the topic for safety demonstration. We demonstrated a case study in which more systematic safety demonstration for the target system software is performed via safety case construction than simply listing the documents.

  3. Safety Justification and Safety Case for Safety-critical Software in Digital Reactor Protection System

    International Nuclear Information System (INIS)

    Kwon, Kee-Choon; Lee, Jang-Soo; Jee, Eunkyoung

    2016-01-01

    Nuclear safety-critical software is under strict regulatory requirements and these regulatory requirements are essential for ensuring the safety of nuclear power plants. The verification & validation (V and V) and hazard analysis of the safety-critical software are required to follow regulatory requirements through the entire software life cycle. In order to obtain a license from the regulatory body through the development and validation of safety-critical software, it is essential to meet the standards which are required by the regulatory body throughout the software development process. Generally, large amounts of documents, which demonstrate safety justification including standard compliance, V and V, hazard analysis, and vulnerability assessment activities, are submitted to the regulatory body during the licensing process. It is not easy to accurately read and evaluate the whole documentation for the development activities, implementation technology, and validation activities. The safety case methodology has been kwon a promising approach to evaluate the level and depth of the development and validation results. A safety case is a structured argument, supported by a body of evidence that provides a compelling, comprehensible, and valid case that a system is safe for a given application in a given operating environment. It is suggested to evaluate the level and depth of the results of development and validation by applying safety case methodology to achieve software safety demonstration. A lot of documents provided as evidence are connected to claim that corresponds to the topic for safety demonstration. We demonstrated a case study in which more systematic safety demonstration for the target system software is performed via safety case construction than simply listing the documents

  4. Control of Industrial Safety Based on Dynamic Characteristics of a Safety Budget-Industrial Accident Rate Model in Republic of Korea.

    Science.gov (United States)

    Choi, Gi Heung; Loh, Byoung Gook

    2017-06-01

    Despite the recent efforts to prevent industrial accidents in the Republic of Korea, the industrial accident rate has not improved much. Industrial safety policies and safety management are also known to be inefficient. This study focused on dynamic characteristics of industrial safety systems and their effects on safety performance in the Republic of Korea. Such dynamic characteristics are particularly important for restructuring of the industrial safety system. The effects of damping and elastic characteristics of the industrial safety system model on safety performance were examined and feedback control performance was explained in view of cost and benefit. The implications on safety policies of restructuring the industrial safety system were also explored. A strong correlation between the safety budget and the industrial accident rate enabled modeling of an industrial safety system with these variables as the input and the output, respectively. A more effective and efficient industrial safety system could be realized by having weaker elastic characteristics and stronger damping characteristics in it. A substantial decrease in total social cost is expected as the industrial safety system is restructured accordingly. A simple feedback control with proportional-integral action is effective in prevention of industrial accidents. Securing a lower level of elastic industrial accident-driving energy appears to have dominant effects on the control performance compared with the damping effort to dissipate such energy. More attention needs to be directed towards physical and social feedbacks that have prolonged cumulative effects. Suggestions for further improvement of the safety system including physical and social feedbacks are also made.

  5. A formal safety analysis for PLC software-based safety critical system using Z

    International Nuclear Information System (INIS)

    Koh, Jung Soo

    1997-02-01

    This paper describes a formal safety analysis technique which is demonstrated by performing empirical formal safety analysis with the case study of beamline hutch door Interlock system that is developed by using PLC (Programmable Logic Controller) systems at the Pohang Accelerator Laboratory. In order to perform formal safety analysis, we have built the Z formal specifications representation from user requirement written in ambiguous natural language and target PLC ladder logic, respectively. We have also studied the effective method to express typical PLC timer component by using specific Z formal notation which is supported by temporal history. We present a formal proof technique specifying and verifying that the hazardous states are not introduced into ladder logic in the PLC-based safety critical system. And also, we have found that some errors or mismatches in user requirement and final implemented PLC ladder logic while analyzing the process of the consistency and completeness of Z translated formal specifications. In the case of relatively small systems like Beamline hutch door interlock system, a formal safety analysis including explicit proof is highly recommended so that the safety of PLC-based critical system may be enhanced and guaranteed. It also provides a helpful benefits enough to comprehend user requirement expressed by ambiguous natural language

  6. Improvement of risk informed surveillance test interval for the safety related instrument and control system of Ulchin units 3 and 4

    International Nuclear Information System (INIS)

    Jang, Seung Cheol; Lee, Yun Hwan; Lee, Seung Joon; Han, Sang Hoon

    2012-05-01

    The purpose of this research is the development of various methodologies necessary for the licensing of the risk informed surveillance test interval(STI) improvement for the safety related I and C systems in UCN 3 and 4, for instance, reactor protection system (RPS), engineered safety features actuation system (ESFAS), ESF auxiliary relay cabinet (ARC), and core protection calculator (CPC). The technical adequacy of the methodology was sufficiently verified through the application to the following STI changes. o CPC channel functional test (change from 1 month to 3 months including safety channel and log power test) o RPS channel functional test (change from 1 month to 3 months) o RPS logic and trip channel test (change from 1 month to 3 months. 1 month for RPS manual actuation test) o ESFAS channel functional test (change from 1 month to 3 months) o ESFAS logic and trip channel test (change from 1 month to 3 months) o ESF auxiliary relay test (change from 1 month to 3 months with staggered test. Manual actuation at the ESF ARC is added as a backup of ESF actuation signals during emergency operation

  7. Improvement of risk informed surveillance test interval for the safety related instrumentation and control system of Yonggwang units 3 and 4

    International Nuclear Information System (INIS)

    Jang, Seung Cheol; Lee, Yun Hwan; Lee, Seung Joon; Han, Sang Hoon

    2012-05-01

    The purpose of this research is the development of various methodologies necessary for the licensing of the risk informed surveillance test interval(STI) improvement for the safety related I and C systems in YGN 3 and 4, for instance, reactor protection system (RPS), engineered safety features actuation system (ESFAS), ESF auxiliary relay cabinet (ARC), and core protection calculator (CPC). The technical adequacy of the methodology was sufficiently verified through the application to the following STI changes. o CPC channel functional test (change from 1 month to 3 months including safety channel and log power test) o RPS channel functional test (change from 1 month to 3 months) o RPS logic and trip channel test (change from 1 month to 3 months. 1 month for RPS manual actuation test) o ESFAS channel functional test (change from 1 month to 3 months) o ESFAS logic and trip channel test (change from 1 month to 3 months) o ESF auxiliary relay test (change from 1 month to 3 months with staggered test. Manual actuation at the ESF ARC is added as a backup of ESF actuation signals during emergency operation

  8. European protection principles against external hazards by means of Emergency Power Supply and Control Safety System Building in Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Gallinat, Dipl Ing [Max Aicher Engineering GmbH, Freilassing (Germany)

    2016-10-15

    One of the most important nuclear power plant safety requirements is a redundant and independent power system. This requires such a design of emergency power systems that failure of one will not adversely impact the other. External hazards of natural origin or linked to human activity could potentially affect plant safety. The general objective of the design provisions is to ensure that the safety functions of the systems and components required to return the plant to a safe shutdown state and to prevent and limit radioactive release are not adversely affected. As external hazards are site dependent, Technical Guidelines specify that 'it is not necessary to take all of the hazards in a standardized design; such external hazards as external flooding, drought, ice formation and toxic, corrosive or combustible gases may be dealt with only for a specific plant, on a plant specific basis'. In accordance with the Technical Guidelines, external hazards are taken into consideration at the design stage consistently with internal events or hazards. The basic design principle is to protect against external hazards in accordance with the Technical Guidelines using a 'load case' procedure.

  9. SAFETY: STRICTER CONTROLS IN CONTROLLED AREAS IN THE PS

    CERN Multimedia

    G. Daems

    2001-01-01

    The PS accelerators will soon stop for several months. Work will take place in controlled areas in the PS and will involve many people who are not always aware of the risks associated with the work sites. To guarentee the safety of these workers, the following two measures will be applied: everyone working in a controlled zone - Linacs, PSB, and PS machines tunnels, and transfer lines - must wear, visibly, his CERN access card and his film badge. the CERN access card and the film badge will only be issued after following a basic safety course. Regular checks will be carried out during the shutdown. Anyone without these two items on their person will be obliged to leave the area immediately.

  10. Artificial intelligence enhancements to safety parameter display systems

    International Nuclear Information System (INIS)

    Hajek, B.K.; Hashemi, S.; Sharma, D.; Chandrasekaran, B.; Miller, D.W.

    1986-01-01

    Two prototype knowledge based systems have been developed at The Ohio State University to be the basis of an operator aid that can be attached to an existing nuclear power plant Safety Parameter Display System. The first system uses improved sensor validation techniques to provide input to a fault diagnosis process. The second system would use the diagnostic system output to synthesize corrective procedures to aid the control room licensed operator in plant recovery

  11. Enhanced Maritime Safety through Diagnosis and Fault Tolerant Control

    DEFF Research Database (Denmark)

    Blanke, Mogens

    2001-01-01

    Faults in steering, navigation instruments or propulsion machinery are serious on a marine vessel since the consequence could be loss of maneuvering ability, and imply risk of damage to vessel personnel or environment. Early diagnosis and accomodation of faults could enhance safety. Fault...... of properties of a falty system; means to determine remedial actions. The paper illustrates the techniques by two marine examples, sensor fusion for automatic steering and control of the main engine....

  12. The effect of using road safety equipment and systems and ...

    African Journals Online (AJOL)

    The effect of using road safety equipment and systems and determine their role on ... traffic control equipment situation and by multi-criteria weighting systems AHP ... The results have shown that indices median, lighting and panel type and the ...

  13. Replacement of the complete control system of the NPP Oskarshamn 1 by digital distributed control system

    International Nuclear Information System (INIS)

    Berger, E.

    1998-01-01

    As part of an ongoing modernization program, the I and C system and the control room of Oskarshamn 1 will be upgraded by ABB using its 'Advant Power' range of digital, programmable process control system. Besides ensuring the higher level of safety that is demanded today, the new equipment provides the plant with an integrated system which will improve operator interaction with the plant and reduce the risk of human error. The newly installed DCS system will serve also as a platform for further improvements of the control room. This paper discusses Oskarshamn 1 exchange of the complete control system of a nuclear power plant, the technical solution and the time schedule. Oskarshamn 1 is the first nuclear power plant in Sweden. It is a boiling water reactor built between 1966 and 1971 by ABB ATOM in Sweden. According to the plant age the control system is relay-based, while instrumentation and analogue control is semiconductor-based. This makes maintenance expensive and even worse, makes extensions nearly impossible. According to the safety standards of the 1960s, there is no separation between safety and non safety control and no seismic qualification. To extend the life of this plant the owner has decided to improve the safety system as well as to replace the reactor protection system, the safety related control and the non safety related control by a state-of-the-art digital distributed control system from ABB. In March 1997, ABB got the order to replace the reactor protection system, the safety control system and to start the replacement of all control systems. The old control room has to be replaced by a new ergonomically design. Together with the exchange of the control system the safety features of the plant and the emergency power supply has to be extended. (author)

  14. Access Safety Systems – New Concepts from the LHC Experience

    CERN Document Server

    Ladzinski, T; di Luca, S; Hakulinen, T; Hammouti, L; Riesco, T; Nunes, R; Ninin, P; Juget, J-F; Havart, F; Valentini, F; Sanchez-Corral Mena, E

    2011-01-01

    The LHC Access Safety System has introduced a number of new concepts into the domain of personnel protection at CERN. These can be grouped into several categories: organisational, architectural and concerning the end-user experience. By anchoring the project on the solid foundations of the IEC 61508/61511 methodology, the CERN team and its contractors managed to design, develop, test and commission on time a SIL3 safety system. The system uses a successful combination of the latest Siemens redundant safety programmable logic controllers with a traditional relay logic hardwired loop. The external envelope barriers used in the LHC include personnel and material access devices, which are interlocked door-booths introducing increased automation of individual access control, thus removing the strain from the operators. These devices ensure the inviolability of the controlled zones by users not holding the required credentials. To this end they are equipped with personnel presence detectors and th...

  15. Precision digital control systems

    Science.gov (United States)

    Vyskub, V. G.; Rozov, B. S.; Savelev, V. I.

    This book is concerned with the characteristics of digital control systems of great accuracy. A classification of such systems is considered along with aspects of stabilization, programmable control applications, digital tracking systems and servomechanisms, and precision systems for the control of a scanning laser beam. Other topics explored are related to systems of proportional control, linear devices and methods for increasing precision, approaches for further decreasing the response time in the case of high-speed operation, possibilities for the implementation of a logical control law, and methods for the study of precision digital control systems. A description is presented of precision automatic control systems which make use of electronic computers, taking into account the existing possibilities for an employment of computers in automatic control systems, approaches and studies required for including a computer in such control systems, and an analysis of the structure of automatic control systems with computers. Attention is also given to functional blocks in the considered systems.

  16. BSF control system

    International Nuclear Information System (INIS)

    Irie, Y.; Ishii, K.; Ninomiya, S.; Sasaki, H.; Sakai, I.

    1982-08-01

    The booster synchrotron utilization facility (BSF) is a facility which utilizes the four fifths of available beam pulses from the KEK booster synchrotron. The BSF control system includes the beam line control, interactions with the PS central control room and the experimental facilities, and the access control system. A brief description of the various components in the control system is given. (author)

  17. The passive safety systems of the Swr 1000

    International Nuclear Information System (INIS)

    Neumann, D.

    2001-01-01

    In recent years, a new boiling water reactor (BWR) plant called the SWR 1000 has been developed by Siemens on behalf of Germany's electric utilities. This new plant design concept incorporates the wide range of operating experience gained with German BWRs. The main objective behind developing the SWR 1000 was to design a plant with a rated electric output of approximately 1000 MW which would not only have a lower capital cost and lower power generating costs but would also provide a much higher level of nuclear safety compared to plants currently in operation. This safety-related goal has been met through, for example, the use of passive safety equipment. Passive systems make a significant contribution towards increasing the over-all level of plant safety due to the way in which they operate. They function solely accord-ing to basic laws of nature, such as gravity, and perform their designated functions with-out any need for electric power or other sources of external energy, or signals from instrumentation and control (I and C) equipment. The passive safety systems have been designed such that design basis accidents can be controlled using just these systems alone. However, the design concept of the SWR 1000 is nevertheless still based on the provision of active safety systems in addition to passive systems. (author)

  18. [Infection control and safety culture in German hospitals].

    Science.gov (United States)

    Hansen, Sonja; Schwab, Frank; Gropmann, Alexander; Behnke, Michael; Gastmeier, Petra

    2016-07-01

    Healthcare-associated infections (HAI) are the most frequent adverse events in the healthcare setting and their prevention is an important contribution to patient safety in hospitals. To analyse to what extent safety cultural aspects with relevance to infection control are implemented in German hospitals. Safety cultural aspects of infection control were surveyed with an online questionnaire; data were analysed descriptively. Data from 543 hospitals with a median of [IQR] 275 [157; 453] beds were analysed. Almost all hospitals (96.6 %) had internal guidelines for infection control (IC) in place; 82 % defined IC objectives, most often regarding hand hygiene (HH) (93 %) and multidrug resistant organisms (72 %) and less frequently for antibiotic stewardship (48 %) or prevention of specific HAI. In 94 % of hospitals, a reporting system for adverse events was in place, which was also used to report low compliance with HH, outbreaks and Clostridium difficile-associated infections. Members of the IC team were most often seen to hold daily responsibility for IC in the hospital, but rarely other hospital staff (94 versus 19 %). Safety cultural aspects are not fully implemented in German hospitals. IC should be more strongly implemented in healthcare workers' daily routine and more visibly supported by hospital management.

  19. Safety aspect of digital reactor protection system in Japan

    International Nuclear Information System (INIS)

    Ogiso, Zen-Ichi

    1998-01-01

    It was early in 1980's that the digital controllers were first applied to nuclear power plant in japan. After that, their application area had been expanding gradually, reaching to the overall integrated digital system including the safety system in Kashiwazaki-Kariwa units 6 and 7. The software for computer-based systems has been produced using the graphical language ''POL'' in Japanese nuclear power plants. It is the fundamental principle that the reliability of the software should be assured through the properly managed quality assurance. The POL-based system is fitted to this principle. In applying POL-based systems to safety system, the MITI, Ministry of International Trade and Industry, identified the licensing issues as the regulatory body, while the utilities had developed the digital technology feasible to the safety application. Through the activities, a specific industrial design guide for the software important to safety was established and the adequacy of the technology was certified through the demonstration tests of the integrated system. In the safety examination of the digital reactor protection system of K-6/7, the application of POL were approved. The POL-based systems in nuclear power plants were successful design and production process of the POL-based systems. This paper describes the activities in licensing and maintaining the computer-based systems by the utilities and manufacturers as well as the MITI. (author)

  20. Modeling for safety in a synthesis-centric systems engineering framework

    NARCIS (Netherlands)

    Markovski, J.; Mortel - Fronczak, van de J.M.; Ortmeier, F.; Daniel, P.

    2012-01-01

    The ever-increasing complexity of safety-critical systems puts high demands on safety assurance and certification. We focus on the development of control software, where safety) requirements engineering plays a crucial and delicate role. Nowadays, most of the safety features are ensured by the

  1. Field Programmable Gate Array-based I and C Safety System

    International Nuclear Information System (INIS)

    Kim, Hyun Jeong; Kim, Koh Eun; Kim, Young Geul; Kwon, Jong Soo

    2014-01-01

    Programmable Logic Controller (PLC)-based I and C safety system used in the operating nuclear power plants has the disadvantages of the Common Cause Failure (CCF), high maintenance costs and quick obsolescence, and then it is necessary to develop the other platform to replace the PLC. The Field Programmable Gate Array (FPGA)-based Instrument and Control (I and C) safety system is safer and more economical than Programmable Logic Controller (PLC)-based I and C safety system. Therefore, in the future, FPGA-based I and C safety system will be able to replace the PLC-based I and C safety system in the operating and the new nuclear power plants to get benefited from its safety and economic advantage. FPGA-based I and C safety system shall be implemented and verified by applying the related requirements to perform the safety function

  2. Field Programmable Gate Array-based I and C Safety System

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyun Jeong; Kim, Koh Eun; Kim, Young Geul; Kwon, Jong Soo [KEPCO, Daejeon (Korea, Republic of)

    2014-08-15

    Programmable Logic Controller (PLC)-based I and C safety system used in the operating nuclear power plants has the disadvantages of the Common Cause Failure (CCF), high maintenance costs and quick obsolescence, and then it is necessary to develop the other platform to replace the PLC. The Field Programmable Gate Array (FPGA)-based Instrument and Control (I and C) safety system is safer and more economical than Programmable Logic Controller (PLC)-based I and C safety system. Therefore, in the future, FPGA-based I and C safety system will be able to replace the PLC-based I and C safety system in the operating and the new nuclear power plants to get benefited from its safety and economic advantage. FPGA-based I and C safety system shall be implemented and verified by applying the related requirements to perform the safety function.

  3. Safety climate and self-reported injury: assessing the mediating role of employee safety control.

    Science.gov (United States)

    Huang, Yueng-Hsiang; Ho, Michael; Smith, Gordon S; Chen, Peter Y

    2006-05-01

    To further reduce injuries in the workplace, companies have begun focusing on organizational factors which may contribute to workplace safety. Safety climate is an organizational factor commonly cited as a predictor of injury occurrence. Characterized by the shared perceptions of employees, safety climate can be viewed as a snapshot of the prevailing state of safety in the organization at a discrete point in time. However, few studies have elaborated plausible mechanisms through which safety climate likely influences injury occurrence. A mediating model is proposed to link safety climate (i.e., management commitment to safety, return-to-work policies, post-injury administration, and safety training) with self-reported injury through employees' perceived control on safety. Factorial evidence substantiated that management commitment to safety, return-to-work policies, post-injury administration, and safety training are important dimensions of safety climate. In addition, the data support that safety climate is a critical factor predicting the history of a self-reported occupational injury, and that employee safety control mediates the relationship between safety climate and occupational injury. These findings highlight the importance of incorporating organizational factors and workers' characteristics in efforts to improve organizational safety performance.

  4. System containing a safety disk

    International Nuclear Information System (INIS)

    Schupp, W.

    1975-01-01

    The safety element is not overdimensioned at pressures between 2 and 150 atmospheric excess pressure. Therefore the flat bursting disc is mounted within a supporting and stopping holding and the rated breaking point is covered by a supporting body. Its outer diameter sufficiently overlaps the recesses on both sides of the rated breaking point. It absorbs the total load given by the operating pressure. Only a release mechanism with slide wedge, eccentric disc, magnet, and rocker arm releases the supporting body, e.g. if the blow-down pressure is reached, so that the operating pressure may work on the bursting disc. An insulated copper wire layed in the breaking region within the bursting disc in case of shearing off signalizes the instant of failing of the breaking point because of current interruption. (DG) [de

  5. NASA System Safety Handbook. Volume 2: System Safety Concepts, Guidelines, and Implementation Examples

    Science.gov (United States)

    Dezfuli, Homayoon; Benjamin, Allan; Everett, Christopher; Feather, Martin; Rutledge, Peter; Sen, Dev; Youngblood, Robert

    2015-01-01

    This is the second of two volumes that collectively comprise the NASA System Safety Handbook. Volume 1 (NASASP-210-580) was prepared for the purpose of presenting the overall framework for System Safety and for providing the general concepts needed to implement the framework. Volume 2 provides guidance for implementing these concepts as an integral part of systems engineering and risk management. This guidance addresses the following functional areas: 1.The development of objectives that collectively define adequate safety for a system, and the safety requirements derived from these objectives that are levied on the system. 2.The conduct of system safety activities, performed to meet the safety requirements, with specific emphasis on the conduct of integrated safety analysis (ISA) as a fundamental means by which systems engineering and risk management decisions are risk-informed. 3.The development of a risk-informed safety case (RISC) at major milestone reviews to argue that the systems safety objectives are satisfied (and therefore that the system is adequately safe). 4.The evaluation of the RISC (including supporting evidence) using a defined set of evaluation criteria, to assess the veracity of the claims made therein in order to support risk acceptance decisions.

  6. 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

  7. Safety of emerging nuclear energy systems

    International Nuclear Information System (INIS)

    Novikov, V.M.; Slesarev, I.S.

    1989-01-01

    The first stage of world nuclear power development based on light water fission reactors has demonstrated not only rather high rate but at the same time too optimistic attitude to safety problems. Large accidents at Three Mile Island and Chernobyl essentially affects the concept of NP development. As a result the safety and social acceptance of NP became of absolute priority among other problems. That's why emerging nuclear power systems should be first of all estimated from this point of view. In the paper some quantitative criteria of safety derived from estimations of social risk and economic-ecological damage from hypothetical accidents are formulated. On the base of these criteria we define two stages of possible way to meet safety demands: first--development of high safety fission reactors and second--that of asymptotic high safety ENEs. The limits of tolorated expenses for safety are regarded. The basis physical factors determining hazards of NES accidents are considered. This permits to classify the ways of safety demands fulfillment due to physical principals used

  8. K West integrated water treatment system subproject safety analysis document

    International Nuclear Information System (INIS)

    SEMMENS, L.S.

    1999-01-01

    This Accident Analysis evaluates unmitigated accident scenarios, and identifies Safety Significant and Safety Class structures, systems, and components for the K West Integrated Water Treatment System

  9. K West integrated water treatment system subproject safety analysis document

    Energy Technology Data Exchange (ETDEWEB)

    SEMMENS, L.S.

    1999-02-24

    This Accident Analysis evaluates unmitigated accident scenarios, and identifies Safety Significant and Safety Class structures, systems, and components for the K West Integrated Water Treatment System.

  10. Classification of Aeronautics System Health and Safety Documents

    Data.gov (United States)

    National Aeronautics and Space Administration — Most complex aerospace systems have many text reports on safety, maintenance, and associated issues. The Aviation Safety Reporting System (ASRS) spans several...

  11. Control system design method

    Science.gov (United States)

    Wilson, David G [Tijeras, NM; Robinett, III, Rush D.

    2012-02-21

    A control system design method and concomitant control system comprising representing a physical apparatus to be controlled as a Hamiltonian system, determining elements of the Hamiltonian system representation which are power generators, power dissipators, and power storage devices, analyzing stability and performance of the Hamiltonian system based on the results of the determining step and determining necessary and sufficient conditions for stability of the Hamiltonian system, creating a stable control system based on the results of the analyzing step, and employing the resulting control system to control the physical apparatus.

  12. Safety design integrated in the building delivery system

    DEFF Research Database (Denmark)

    Jørgensen, Kirsten

    2013-01-01

    . The purpose of this article is to demonstrate how safety and health can be integrated in the design phases integrated in the management delivery systems within construction, The method for the research was to go through the building delivery system step by step and create a normative description of what, when......In construction, it is important to view safety and health as an integrated part of the way that “designers” are working. The designers cowers architects, constructors, engineers and others who carry out their consulting services in the design phase of a construction project. The philosophy...... and how to fully integrate safety in each part of the process. The result is a concept and guideline including control forms for how to integrate safety design in the Building Delivery System plus what to do and when. The concept has been tested in an educational context. The practical value...

  13. FFTF control system experience

    International Nuclear Information System (INIS)

    Warrick, R.P.

    1981-01-01

    The FFTF control systems provide control equipment for safe and efficient operation of the plant. For convenience, these systems will be divided into three parts for discussions: (1) Plant Protection System (PPS); (2) Plant Control System (PCS); and (3) General Observations. Performance of each of these systems is discussed

  14. System Safety in an IT Service Organization

    Science.gov (United States)

    Parsons, Mike; Scutt, Simon

    Within Logica UK, over 30 IT service projects are considered safetyrelated. These include operational IT services for airports, railway infrastructure asset management, nationwide radiation monitoring and hospital medical records services. A recent internal audit examined the processes and documents used to manage system safety on these services and made a series of recommendations for improvement. This paper looks at the changes and the challenges to introducing them, especially where the service is provided by multiple units supporting both safety and non-safety related services from multiple locations around the world. The recommendations include improvements to service agreements, improved process definitions, routine safety assessment of changes, enhanced call logging, improved staff competency and training, and increased safety awareness. Progress is reported as of today, together with a road map for implementation of the improvements to the service safety management system. A proposal for service assurance levels (SALs) is discussed as a way forward to cover the wide variety of services and associated safety risks.

  15. Safety assessment of high consequence robotics system

    International Nuclear Information System (INIS)

    Robinson, D.G.; Atcitty, C.B.

    1996-01-01

    This paper outlines the use of a failure modes and effects analysis for the safety assessment of a robotic system being developed at Sandia National Laboratories. The robotic system, the weigh and leak check system, is to replace a manual process for weight and leakage of nuclear materials at the DOE Pantex facility. Failure modes and effects analyses were completed for the robotics process to ensure that safety goals for the systems have been met. Due to the flexible nature of the robot configuration, traditional failure modes and effects analysis (FMEA) were not applicable. In addition, the primary focus of safety assessments of robotics systems has been the protection of personnel in the immediate area. In this application, the safety analysis must account for the sensitivities of the payload as well as traditional issues. A unique variation on the classical FMEA was developed that permits an organized and quite effective tool to be used to assure that safety was adequately considered during the development of the robotic system. The fundamental aspects of the approach are outlined in the paper

  16. Operating safety requirements for the intermediate level liquid waste system

    International Nuclear Information System (INIS)

    1980-07-01

    The operation of the Intermediate Level Liquid Waste (ILW) System, which is described in the Final Safety Analysis, consists of two types of operations, namely: (1) the operation of a tank farm which involves the storage and transportation through pipelines of various radioactive liquids; and (2) concentration of the radioactive liquids by evaporation including rejection of the decontaminated condensate to the Waste Treatment Plant and retention of the concentrate. The following safety requirements in regard to these operations are presented: safety limits and limiting control settings; limiting conditions for operation; and surveillance requirements. Staffing requirements, reporting requirements, and steps to be taken in the event of an abnormal occurrence are also described

  17. Regulatory control of nuclear safety in Finland. Annual report 1997

    International Nuclear Information System (INIS)

    Tossavainen, K.

    1998-08-01

    The report describes regulatory control of the use of nuclear energy by the Radiation and Nuclear Safety Authority (STUK) in Finland in 1997. Nuclear regulatory control ascertained that the operation of Finnish NPPs was in compliance with the conditions set out in operating licences and current regulations. In addition to NPP normal operation, STUK oversaw projects at the plant units relating to power uprating and safety improvements. STUK prepared statements for the Ministry of Trade and Industry about the applications for renewing the operating licenses of Loviisa and Olkiluoto NPPs. The most important items of supervision in nuclear waste management were studies relating to the final disposal of spent fuel from NPPs and the review of the licence application for a repository for low- and intermediate-level reactor waste from Loviisa NPP. Preparation of general safety regulations for the final disposal of spent nuclear fuel, to be published in the form of a Council of State Decision, was started. By safeguards control, the use of nuclear materials was verified to be in compliance with current regulations and that the whereabouts of every batch of nuclear material were always known. Nuclear material safeguards were stepped up to prevent illicit trafficking of nuclear materials and other radioactive materials. In co-operation with the Ministry for Foreign Affairs and the Institute of Seismology (University of Helsinki), preparations were undertaken to implement the Comprehensive Nuclear Test Ban Treaty (CTBT). For enforcement of the Treaty and as part of the international regulatory approach, STUK is currently developing laboratory analyses relating to airborne radioactivity measurements. The focus of co-operation funded by external sources was as follows: improvement of the safety of Kola and Leningrad NPPs, improvement of nuclear waste management in North-West Russia, development of the organizations of nuclear safety authorities in Eastern Europe and development

  18. Safety device and machine system of nuclear power plant

    International Nuclear Information System (INIS)

    1978-10-01

    It introduces principle and kinds of heat power including heat balance and nuclear power. It explains a lot of technical terms about the nuclear power system, which are primary loop, reactor, steam generator, primary coolant pump and pressurizer in PWR, chemical and volume control system, component cooling system, safety injection system, and spent fuel cooling and storage system in auxiliary system, liquid solid and gaseous waste disposal system in radwaste disposal, gland sealing system, turbine instrumentation, turning gear, hydrogen cooling system, condenser, feedwater heater, degenerate heater, auxiliary heat exchanger, centrifugal pump, rotary reciprocating and tank and pressure vessel.

  19. Regulatory control of radiation sources. Safety guide

    International Nuclear Information System (INIS)

    2004-01-01

    The basic requirements for the protection of persons against exposure to ionizing radiation and for the safety of radiation sources were established in the International Basic Safety Standards for Protection against Ionizing Radiation and for the Safety of Radiation Sources (the Basic Safety Standards), jointly sponsored by the Food and Agriculture Organization of the United Nations (FAO), the International Atomic Energy Agency (IAEA), the International Labour Organization (ILO), the OECD Nuclear Energy Agency (OECD/ NEA), the Pan American Health Organization (PAHO) and the World Health Organization (WHO) (the Sponsoring Organizations). The application of the Basic Safety Standards is based on the presumption that national infrastructures are in place to enable governments to discharge their responsibilities for radiation protection and safety. Requirements relating to the legal and governmental infrastructure for the safety of nuclear facilities and sources of ionizing radiation, radiation protection, the safe management of radioactive waste and the safe transport of radioactive material are established in the Safety Requirements on Legal and Governmental Infrastructure for Nuclear, Radiation, Radioactive Waste and Transport Safety, Safety Standards Series No. GS-R-1. This Safety Guide, which is jointly sponsored by the FAO, the IAEA, the International Labour Office, the PAHO and the WHO, gives detailed guidance on the key elements for the organization and operation of a national regulatory infrastructure for radiation safety, with particular reference to the functions of the national regulatory body that are necessary to ensure the implementation of the Basic Safety Standards. The Safety Guide is based technically on material first published in IAEA-TECDOC-10671, which was jointly sponsored by the FAO, the IAEA, the OECD/NEA, the PAHO and the WHO. The requirements established in GS-R-1 have been taken into account. The Safety Guide is oriented towards national

  20. Understanding Nuclear Safety Culture: A Systemic Approach

    International Nuclear Information System (INIS)

    Afghan, A.N.

    2016-01-01

    The Fukushima accident was a systemic failure (Report by Director General IAEA on the Fukushima Daiichi Accident). Systemic failure is a failure at system level unlike the currently understood notion which regards it as the failure of component and equipment. Systemic failures are due to the interdependence, complexity and unpredictability within systems and that is why these systems are called complex adaptive systems (CAS), in which “attractors” play an important role. If we want to understand the systemic failures we need to understand CAS and the role of these attractors. The intent of this paper is to identify some typical attractors (including stakeholders) and their role within complex adaptive system. Attractors can be stakeholders, individuals, processes, rules and regulations, SOPs etc., towards which other agents and individuals are attracted. This paper will try to identify attractors in nuclear safety culture and influence of their assumptions on safety culture behavior by taking examples from nuclear industry in Pakistan. For example, if the nuclear regulator is an attractor within nuclear safety culture CAS then how basic assumptions of nuclear plant operators and shift in-charges about “regulator” affect their own safety behavior?

  1. New Paradigm in Nuclear Safety from Quality Assurance to Safety Management System

    International Nuclear Information System (INIS)

    Lim, Nam-Jin; Park, Chan-Gook; Nam, Ji-Hee; Kim, Kwan-Hyun; Kwon, Hyuk-il; Lee, Young-Gun Lee

    2006-01-01

    The initial concept of Quality Control (QC) controlling the quality of products is now evolving toward the Management System (MS) achieving safety, through Quality Assurance (QA) ensuring the quality of products and Quality Management (QM) managing the quality by a systematic approach. Nuclear safety can be achieved through an integrated MS that ensures the health, environmental, security, quality and economic requirements being considered together with nuclear safety requirements. MS approach is developed through realizing that most of nuclear accidents had occurred not by the malfunction of hardware or equipment, but by the human error. The MS is a set of inter-related or interacting elements (system) that establishes policies and objectives and which enables those objectives to be achieved in an efficient and effective way

  2. Use of feedback control to address flight safety issues

    Science.gov (United States)

    Ganguli, Subhabrata

    This thesis addresses three control problems related to flight safety. The first problem relates to the scope of improvement in performance of conventional flight control laws. In particular, aircraft longitudinal axis control based on the Total Energy Control System (TECS) is studied. The research draws attention to a potentially sluggish and undesirable aircraft response when the engine dynamics is slow (typically the case). The proposed design method uses a theoretically well-developed modern design method based on Hinfinity optimization to improve the aircraft dynamic behavior in spite of slow engine characteristics. At the same time, the proposed design method achieves other desirable performance goals such as insensitivity to sensor noise and wind gust rejection: all addressed in one unified framework. The second problem is based on a system level analysis of control structure hierarchy for aircraft flight control. The objective of the analysis problem is to translate outer-loop stability and performance specifications into a comprehensive inner-loop metric. The prime motivation is to make the flight control design process more systematic and the system-integration reliable and independent of design methodology. The analysis problem is posed within the robust control analysis framework. Structured singular value techniques and free controller parameterization ideas are used to impose a hierarchical structure for flight control architecture. The third problem involves development and demonstration of a new reconfiguration strategy in the flight control architecture that has the potential of improving flight safety while keeping cost and complexity low. This research proposes a fault tolerant feature based on active robust reconfiguration. The fault tolerant control problem is formulated in the Linear Parameter Varying (LPV) design framework. A prime advantage of this approach is that the synthesis results in a single nonlinear controller (as opposed to a bank

  3. Safety evaluation report related to the preliminary design of the Standard Reference System, RESAR-414

    International Nuclear Information System (INIS)

    1978-11-01

    The safety evaluation for the Westinghouse Standard Reactor includes information on general reactor characteristics; design criteria for systems and components; reactor coolant system; engineered safety systems; instrumentation and controls; electric power systems; auxiliary systems; steam and power conversion system; radioactive waste management; radiation protection; conduct of operations; accident analyses; and quality assurance

  4. 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

  5. Regulatory control of nuclear safety in Finland. Annual report 2008

    International Nuclear Information System (INIS)

    Kainulainen, E.

    2009-06-01

    facilities is examined using the employees' individual doses, the collective doses, and the results of emission and environmental radiation control. Summaries are also included for the regulation of the storage of spent nuclear fuel and the processing and storage of reactor waste. For the Olkiluoto 3 plant unit currently under construction, the report includes descriptions of the regulation of design, construction, manufacturing, installation and implementation preparations, as well as regulation of the operations of organisations participating in the construction project. The nuclear safety indicator system is used to examine the efficiency and effects of the regulatory activities targeted at nuclear power plants. Appendices to the report include detailed data and conclusions related to the indicators (Appendix 1) and any significant operational events (Appendix 3). The chapter concerning the regulation of the final disposal project for spent nuclear fuel describes the preparations for the final disposal project and the related regulatory activities. In addition, the oversight of the design and construction of the research facilities (Onkalo) under construction in Olkiluoto, as well as the assessment and oversight of the research, development and design work being carried out to further specify the safety case for final disposal are included in the report. The section concerning nuclear non-proliferation describes the nuclear non-proliferation control for Finnish nuclear facilities and final disposal of spent nuclear fuel, as well as measures required by the Additional Protocol of the Safeguards Agreement. Oversight of the nuclear test ban is also covered by the report. In addition to actual safety regulation, the report describes the enforcement of the regulatory oversight of nuclear facilities, regulatory indicators and the development of regulation, as well as safety research, emergency preparedness, communications and STUK's participation in international nuclear safety

  6. Practical Applications of Cosmic Ray Science: Spacecraft, Aircraft, Ground-Based Computation and Control Systems, Exploration, and Human Health and Safety

    Science.gov (United States)

    Koontz, Steve

    2015-01-01

    In this presentation a review of galactic cosmic ray (GCR) effects on microelectronic systems and human health and safety is given. The methods used to evaluate and mitigate unwanted cosmic ray effects in ground-based, atmospheric flight, and space flight environments are also reviewed. However not all GCR effects are undesirable. We will also briefly review how observation and analysis of GCR interactions with planetary atmospheres and surfaces and reveal important compositional and geophysical data on earth and elsewhere. About 1000 GCR particles enter every square meter of Earth’s upper atmosphere every second, roughly the same number striking every square meter of the International Space Station (ISS) and every other low- Earth orbit spacecraft. GCR particles are high energy ionized atomic nuclei (90% protons, 9% alpha particles, 1% heavier nuclei) traveling very close to the speed of light. The GCR particle flux is even higher in interplanetary space because the geomagnetic field provides some limited magnetic shielding. Collisions of GCR particles with atomic nuclei in planetary atmospheres and/or regolith as well as spacecraft materials produce nuclear reactions and energetic/highly penetrating secondary particle showers. Three twentieth century technology developments have driven an ongoing evolution of basic cosmic ray science into a set of practical engineering tools needed to design, test, and verify the safety and reliability of modern complex technological systems and assess effects on human health and safety effects. The key technology developments are: 1) high altitude commercial and military aircraft; 2) manned and unmanned spacecraft; and 3) increasingly complex and sensitive solid state micro-electronics systems. Space and geophysical exploration needs drove the development of the instruments and analytical tools needed to recover compositional and structural data from GCR induced nuclear reactions and secondary particle showers. Finally, the

  7. Food safety performance indicators to benchmark food safety output of food safety management systems.

    Science.gov (United States)

    Jacxsens, L; Uyttendaele, M; Devlieghere, F; Rovira, J; Gomez, S Oses; Luning, P A

    2010-07-31

    There is a need to measure the food safety performance in the agri-food chain without performing actual microbiological analysis. A food safety performance diagnosis, based on seven indicators and corresponding assessment grids have been developed and validated in nine European food businesses. Validation was conducted on the basis of an extensive microbiological assessment scheme (MAS). The assumption behind the food safety performance diagnosis is that food businesses which evaluate the performance of their food safety management system in a more structured way and according to very strict and specific criteria will have a better insight in their actual microbiological food safety performance, because food safety problems will be more systematically detected. The diagnosis can be a useful tool to have a first indication about the microbiological performance of a food safety management system present in a food business. Moreover, the diagnosis can be used in quantitative studies to get insight in the effect of interventions on sector or governmental level. Copyright 2010 Elsevier B.V. All rights reserved.

  8. Regulatory control of nuclear safety in Finland. Annual report 1998

    Energy Technology Data Exchange (ETDEWEB)

    Tossavainen, K. [ed.

    1999-10-01

    Kola and Leningrad NPPs; development of the organisational practices of nuclear safety authorities in Eastern Europe; and development of the nuclear material accounting and control system in Ukraine, the Baltic countries and Russia as well as of the International Atomic Energy Agency (IAEA) safeguards. EU funding was used to take part in activities to assist the activities of nuclear safety authorities in Eastern Europe. The total expenditure of nuclear safety control in 1998 was FIM 32.0 million and the total income FIM 28.0 million. The total expenditure of fee-charging regulatory activities was FIM 27.8 million and the income was FIM 27.8 million. (orig.)

  9. Regulatory control of nuclear safety in Finland. Annual report 1998

    International Nuclear Information System (INIS)

    Tossavainen, K.

    1999-10-01

    Leningrad NPPs; development of the organisational practices of nuclear safety authorities in Eastern Europe; and development of the nuclear material accounting and control system in Ukraine, the Baltic countries and Russia as well as of the International Atomic Energy Agency (IAEA) safeguards. EU funding was used to take part in activities to assist the activities of nuclear safety authorities in Eastern Europe. The total expenditure of nuclear safety control in 1998 was FIM 32.0 million and the total income FIM 28.0 million. The total expenditure of fee-charging regulatory activities was FIM 27.8 million and the income was FIM 27.8 million. (orig.)

  10. Plant control impact on IFR power plant passive safety response

    International Nuclear Information System (INIS)

    Vilim, R.B.

    1993-01-01

    A method is described for optimizing the closed-loop plant control strategy with respect to safety margins sustained in the unprotected upset response of a liquid metal reactor. The optimization is performed subject to the normal requirements for reactor startup, load change and compensation for reactivity changes over the cycle. The method provides a formal approach to the process of exploiting the innate self-regulating property of a metal fueled reactor to make it less dependent on operator action and less vulnerable to automatic control system fault and/or operator error

  11. Safety Analysis of Stochastic Dynamical Systems

    DEFF Research Database (Denmark)

    Sloth, Christoffer; Wisniewski, Rafael

    2015-01-01

    This paper presents a method for verifying the safety of a stochastic system. In particular, we show how to compute the largest set of initial conditions such that a given stochastic system is safe with probability p. To compute the set of initial conditions we rely on the moment method that via...... that shows how the p-safe initial set is computed numerically....

  12. Safety analysis of accident localization system

    International Nuclear Information System (INIS)

    1999-01-01

    A complex safety analysis of accident localization system of Ignalina NPP was performed. Calculation results obtained, results of non-destruct ing testing and experimental data of reinforced concrete testing of buildings does not revealed deficiencies of buildings of accident localization system at unit 1 of Ignalina NPP. Calculations were performed using codes NEPTUNE, ALGOR, CONTAIN

  13. Strata control in tunnels and an evaluation of support units and systems currently used with a view to improving the effectiveness of support stability and safety of tunnels.

    CSIR Research Space (South Africa)

    Haile, AT

    1995-12-01

    Full Text Available This project report addresses the issue of strata control in tunnel excavations with the aim of improving the stability of the excavation through improved design methodologies and support systems....

  14. Cyber Security Risk Assessment for the KNICS Safety Systems

    International Nuclear Information System (INIS)

    Lee, C. K.; Park, G. Y.; Lee, Y. J.; Choi, J. G.; Kim, D. H.; Lee, D. Y.; Kwon, K. C.

    2008-01-01

    In the Korea Nuclear I and C Systems Development (KNICS) project the platforms for plant protection systems are developed, which function as a reactor shutdown, actuation of engineered safety features and a control of the related equipment. Those are fully digitalized through the use of safety-grade programmable logic controllers (PLCs) and communication networks. In 2006 the Regulatory Guide 1.152 (Rev. 02) was published by the U.S. NRC and it describes the application of a cyber security to the safety systems in the Nuclear Power Plant (NPP). Therefore it is required that the new requirements are incorporated into the developed platforms to apply to NPP, and a cyber security risk assessment is performed. The results of the assessment were input for establishing the cyber security policies and planning the work breakdown to incorporate them

  15. Safety parameter display system: an operator support system for enhancement of safety in Indian PHWRs

    International Nuclear Information System (INIS)

    Subramaniam, K.; Biswas, T.

    1994-01-01

    Ensuring operational safety in nuclear power plants is important as operator errors are observed to contribute significantly to the occurrence of accidents. Computerized operator support systems, which process and structure information, can help operators during both normal and transient conditions, and thereby enhance safety and aid effective response to emergency conditions. An important operator aid being developed and described in this paper, is the safety parameter display system (SPDS). The SPDS is an event-independent, symptom-based operator aid for safety monitoring. Knowledge-based systems can provide operators with an improved quality of information. An information processing model of a knowledge based operator support system (KBOSS) developed for emergency conditions using an expert system shell is also presented. The paper concludes with a discussion of the design issues involved in the use of a knowledge based systems for real time safety monitoring and fault diagnosis. (author). 8 refs., 4 figs., 1 tab

  16. Wisdom Appliance Control System

    Science.gov (United States)

    Hendrick; Jheng, Jyun-Teng; Tsai, Chen-Chai; Liou, Jia-Wei; Wang, Zhi-Hao; Jong, Gwo-Jia

    2017-07-01

    Intelligent appliances wisdom involves security, home care, convenient and energy saving, but the home automation system is still one of the core unit, and also using micro-processing electronics technology to centralized and control the home electrical products and systems, such as: lighting, television, fan, air conditioning, stereo, it composed of front-controller systems and back-controller panels, user using front-controller to control command, and then through the back-controller to powered the device.

  17. Software qualification for digital safety system in KNICS project

    International Nuclear Information System (INIS)

    Kwon, Kee-Choon; Lee, Dong-Young; Choi, Jong-Gyun

    2012-01-01

    In order to achieve technical self-reliance in the area of nuclear instrumentation and control, the Korea Nuclear Instrumentation and Control System (KNICS) project had been running for seven years from 2001. The safety-grade Programmable Logic Controller (PLC) and the digital safety system were developed by KNICS project. All the software of the PLC and digital safety system were developed and verified following the software development life cycle Verification and Validation (V and V) procedure. The main activities of the V and V process are preparation of software planning documentations, verification of the Software Requirement Specification (SRS), Software Design Specification (SDS) and codes, and a testing of the software components, the integrated software, and the integrated system. In addition, a software safety analysis and a software configuration management are included in the activities. For the software safety analysis at the SRS and SDS phases, the software Hazard Operability (HAZOP) was performed and then the software fault tree analysis was applied. The software fault tree analysis was applied to a part of software module with some critical defects identified by the software HAZOP in SDS phase. The software configuration management was performed using the in-house tool developed in the KNICS project. (author)

  18. Formation of maintenance economic safety enterprise system

    Directory of Open Access Journals (Sweden)

    N. A. Serebryakova

    2016-01-01

    Full Text Available The article examines the issues of economic security. The operation of enterprises is being implemented in a volatile market environment, which requires a comprehensive assessment of not only the individual factors affecting the operation of the enterprise, but also encourages the need to develop a comprehensive system for the enterprise to ensure economic security. The purpose of this study is to examine the theoretical and methodological approaches to assessing and ensuring the economic security of the enterprise, the development of a mechanism to ensure the economic security of the enterprise. Measures to ensure the safety of personnel suggest preventive work with the personnel, training personnel of the security services division, formation of personnel reserve of security personnel, the organization of work with new employees, reducing staff turnover. Preventive measures to minimize include activities not directly related to the activities of security units, but to minimize losses of commercial enterprise in the course of maintenance operations: control of inventories; control document; scheduled and unscheduled inspections during the reception of the goods; selection and organization of the movement control risk goods. Development of guidelines and regulations involves the planning of a clear legal regulation of all processes for the operation of commercial facility, potentially dangerous from the point of view of any commercial activity or threats to the security risks. The success of the activities is largely determined by the speed and accuracy of enterprise responses to emerging threats, where a key determinant of the effectiveness of business, is to create a system to ensure the economic security of the enterprise.

  19. Job Demands-Control-Support model and employee safety performance.

    Science.gov (United States)

    Turner, Nick; Stride, Chris B; Carter, Angela J; McCaughey, Deirdre; Carroll, Anthony E

    2012-03-01

    The aim of this study was to explore whether work characteristics (job demands, job control, social support) comprising Karasek and Theorell's (1990) Job Demands-Control-Support framework predict employee safety performance (safety compliance and safety participation; Neal and Griffin, 2006). We used cross-sectional data of self-reported work characteristics and employee safety performance from 280 healthcare staff (doctors, nurses, and administrative staff) from Emergency Departments of seven hospitals in the United Kingdom. We analyzed these data using a structural equation model that simultaneously regressed safety compliance and safety participation on the main effects of each of the aforementioned work characteristics, their two-way interactions, and the three-way interaction among them, while controlling for demographic, occupational, and organizational characteristics. Social support was positively related to safety compliance, and both job control and the two-way interaction between job control and social support were positively related to safety participation. How work design is related to employee safety performance remains an important area for research and provides insight into how organizations can improve workplace safety. The current findings emphasize the importance of the co-worker in promoting both safety compliance and safety participation. Crown Copyright © 2011. Published by Elsevier Ltd. All rights reserved.

  20. A Microbial Assessment Scheme to measure microbial performance of Food Safety Management Systems

    NARCIS (Netherlands)

    Jacxsens, L.; Kussaga, J.; Luning, P.A.; Spiegel, van der M.; Devlieghere, F.; Uyttendaele, M.

    2009-01-01

    A Food Safety Management System (FSMS) implemented in a food processing industry is based on Good Hygienic Practices (GHP), Hazard Analysis Critical Control Point (HACCP) principles and should address both food safety control and assurance activities in order to guarantee food safety. One of the

  1. A study of software safety analysis system for safety-critical software

    International Nuclear Information System (INIS)

    Chang, H. S.; Shin, H. K.; Chang, Y. W.; Jung, J. C.; Kim, J. H.; Han, H. H.; Son, H. S.

    2004-01-01

    The core factors and requirements for the safety-critical software traced and the methodology adopted in each stage of software life cycle are presented. In concept phase, Failure Modes and Effects Analysis (FMEA) for the system has been performed. The feasibility evaluation of selected safety parameter was performed and Preliminary Hazards Analysis list was prepared using HAZOP(Hazard and Operability) technique. And the check list for management control has been produced via walk-through technique. Based on the evaluation of the check list, activities to be performed in requirement phase have been determined. In the design phase, hazard analysis has been performed to check the safety capability of the system with regard to safety software algorithm using Fault Tree Analysis (FTA). In the test phase, the test items based on FMEA have been checked for fitness guided by an accident scenario. The pressurizer low pressure trip algorithm has been selected to apply FTA method to software safety analysis as a sample. By applying CASE tool, the requirements traceability of safety critical system has been enhanced during all of software life cycle phases

  2. Personal exposure control system

    International Nuclear Information System (INIS)

    Tanabe, Ken-ichi; Akashi, Michio

    1994-01-01

    Nuclear power stations are under strict radiation control. Exposure control for nuclear workers is the most important operation, and so carefully thought out measures are taken. This paper introduces Fuji Electric's personal exposure control system that meets strict exposure control and rationalizes control operations. The system has a merit that it can provide required information in an optimum form using the interconnection of a super minicomputer and exposure control facilities and realizes sophisticated exposure control operations. (author)

  3. Radiation safety management system in a radioactive facility

    International Nuclear Information System (INIS)

    Amador, Zayda H.

    2008-01-01

    Full text: This paper illustrates the Cuban experience in implementing and promoting an effective radiation safety system for the Centre of Isotopes, the biggest radioactive facility of our country. Current management practice demands that an organization inculcate culture of safety in preventing radiation hazard. The aforementioned objectives of radiation protection can only be met when it is implemented and evaluated continuously. Commitment from the workforce to treat safety as a priority and the ability to turn a requirement into a practical language is also important to implement radiation safety policy efficiently. Maintaining and improving safety culture is a continuous process. There is a need to establish a program to measure, review and audit health and safety performance against predetermined standards. All those areas of the radiation protection program are considered (e.g. licensing and training of the staff, occupational exposure, authorization of the practices, control of the radioactive material, radiological occurrences, monitoring equipment, radioactive waste management, public exposure due to airborne effluents, audits and safety costs). A set of indicators designed to monitor key aspects of operational safety performance are used. Their trends over a period of time are analyzed with the modern information technologies, because this can provide an early warning to plant management for searching causes behind the observed changes. In addition to analyze the changes and trends, these indicators are compared against identified targets and goals to evaluate performance strengths and weaknesses. A structured and proper radiation self-auditing system is seen as a basic requirement to meet the current and future needs in sustainability of radiation safety. The integrated safety management system establishment has been identified as a goal and way for the continuous improvement. (author)

  4. Reliability Improved Design for a Safety System Channel

    International Nuclear Information System (INIS)

    Oh, Eung Se; Kim, Yun Goo

    2016-01-01

    Nowadays, these systems are implemented with a same platform type, such as a qualified programmable logic controller (PLC). The platform intensively uses digital communication with fiber-optic links to reduce cabling costs and to achieve effective signal isolation. These communication interface and redundancies within a channel increase the complexness of an overall system design. This paper proposes a simpler channel architecture design to reduce the complexity and to enhance overall channel reliability. Simplified safety channel configuration is proposed and the failure probabilities are compared with baseline safety channel configuration using an estimated generic value. The simplified channel configuration achieves 40 percent failure reduction compare to baseline safety channel configuration. If this configuration can be implemented within a processor module, overall safety channel reliability is increase and costs of fabrication and maintenance will be greatly reduced

  5. Reliability Improved Design for a Safety System Channel

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Eung Se; Kim, Yun Goo [KHNP, Daejeon (Korea, Republic of)

    2016-05-15

    Nowadays, these systems are implemented with a same platform type, such as a qualified programmable logic controller (PLC). The platform intensively uses digital communication with fiber-optic links to reduce cabling costs and to achieve effective signal isolation. These communication interface and redundancies within a channel increase the complexness of an overall system design. This paper proposes a simpler channel architecture design to reduce the complexity and to enhance overall channel reliability. Simplified safety channel configuration is proposed and the failure probabilities are compared with baseline safety channel configuration using an estimated generic value. The simplified channel configuration achieves 40 percent failure reduction compare to baseline safety channel configuration. If this configuration can be implemented within a processor module, overall safety channel reliability is increase and costs of fabrication and maintenance will be greatly reduced.

  6. From Safe Systems to Patient Safety

    DEFF Research Database (Denmark)

    Aarts, J.; Nøhr, C.

    2010-01-01

    for the third conference with the theme: The ability to design, implement and evaluate safe, useable and effective systems within complex health care organizations. The theme for this conference was "Designing and Implementing Health IT: from safe systems to patient safety". The contributions have reflected...... and implementation of safe systems and thus contribute to the agenda of patient safety? The contributions demonstrate how the health informatics community has contributed to the performance of significant research and to translating research findings to develop health care delivery and improve patient safety......This volume presents the papers from the fourth International Conference on Information Technology in Health Care: Socio-technical Approaches held in Aalborg, Denmark in June 2010. In 2001 the first conference was held in Rotterdam, The Netherlands with the theme: Sociotechnical' approaches...

  7. Recent advances in systems safety and security

    CERN Document Server

    Stamatescu, Grigore

    2016-01-01

    This book represents a timely overview of advances in systems safety and security, based on selected, revised and extended contributions from the 2nd and 3rd editions of the International Workshop on Systems Safety and Security – IWSSS, held in 2014 and 2015, respectively, in Bucharest, Romania. It includes 14 chapters, co-authored by 34 researchers from 7 countries. The book provides an useful reference from both theoretical and applied perspectives in what concerns recent progress in this area of critical interest. Contributions, broadly grouped by core topic, address challenges related to information theoretic methods for assuring systems safety and security, cloud-based solutions, image processing approaches, distributed sensor networks and legal or risk analysis viewpoints. These are mostly accompanied by associated case studies providing additional practical value and underlying the broad relevance and impact of the field.

  8. Proton beam therapy control system

    Science.gov (United States)

    Baumann, Michael A [Riverside, CA; Beloussov, Alexandre V [Bernardino, CA; Bakir, Julide [Alta Loma, CA; Armon, Deganit [Redlands, CA; Olsen, Howard B [Colton, CA; Salem, Dana [Riverside, CA

    2008-07-08

    A tiered communications architecture for managing network traffic in a distributed system. Communication between client or control computers and a plurality of hardware devices is administered by agent and monitor devices whose activities are coordinated to reduce the number of open channels or sockets. The communications architecture also improves the transparency and scalability of the distributed system by reducing network mapping dependence. The architecture is desirably implemented in a proton beam therapy system to provide flexible security policies which improve patent safety and facilitate system maintenance and development.

  9. 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

  10. Control rod blow out protection system

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  11. Reliability analysis of diverse safety logic systems of fast breeder reactor

    International Nuclear Information System (INIS)

    Ravi Kumar, Bh.; Apte, P.R.; Srivani, L.; Ilango Sambasivan, S.; Swaminathan, P.

    2006-01-01

    Safety Logic for Fast Breeder Reactor (FBR) is designed to initiate safety action against Design Basis Events. Based on the outputs of various processing circuits, Safety logic system drives the control rods of the shutdown system. So, Safety Logic system is classified as safety critical system. Therefore, reliability analysis has to be performed. This paper discusses the Reliability analysis of Diverse Safety logic systems of FBRs. For this literature survey on safety critical systems, system reliability approach and standards to be followed like IEC-61508 are discussed in detail. For Programmable Logic device based systems, Hardware Description Languages (HDL) are used. So this paper also discusses the Verification and Validation for HDLs. Finally a case study for the Reliability analysis of Safety logic is discussed. (author)

  12. Safety systems I/C equipment reliability analyses of the Kozloduy NPP units 3 and 4

    Energy Technology Data Exchange (ETDEWEB)

    Halev, G; Christov, N [Risk Engineering Ltd., Sofia (Bulgaria)

    1996-12-31

    The purpose of the analysis is to assess the safety systems I/C equipment reliability. The assessment includes: quantification of the safety systems unavailability due to component failures; definition of the minimal cut sets leading to the analysed safety systems failure; quantification of the I/C equipment importance measures of the dominant contribution components. The safety systems I/C equipment reliability has been analysed using PSAPACK (a code for probabilistic safety assessment). Fault trees for the following safety systems of the Kozloduy-3 and Kozloduy-4 reactors have been constructed: neutron flow control equipment, reactor protection system, main coolant pumps, pressurizer safety valves `Sempell`, steam dump systems, spray system, low pressure injection system, emergency feeding water system, essential service water system. THree separate reports have been issued containing the performed analyses and results. 1 ref.

  13. Incorporating Traffic Control and Safety Hardware Performance Functions into Risk-based Highway Safety Analysis

    Directory of Open Access Journals (Sweden)

    Zongzhi Li

    2017-04-01

    Full Text Available Traffic control and safety hardware such as traffic signs, lighting, signals, pavement markings, guardrails, barriers, and crash cushions form an important and inseparable part of highway infrastructure affecting safety performance. Significant progress has been made in recent decades to develop safety performance functions and crash modification factors for site-specific crash predictions. However, the existing models and methods lack rigorous treatments of safety impacts of time-deteriorating conditions of traffic control and safety hardware. This study introduces a refined method for computing the Safety Index (SI as a means of crash predictions for a highway segment that incorporates traffic control and safety hardware performance functions into the analysis. The proposed method is applied in a computation experiment using five-year data on nearly two hundred rural and urban highway segments. The root-mean square error (RMSE, Chi-square, Spearman’s rank correlation, and Mann-Whitney U tests are employed for validation.

  14. The remote control system

    International Nuclear Information System (INIS)

    Jansweijer, P.P.M.

    1988-01-01

    The remote-control system is applied in order to control various signals in the car of the spectrometer at distance. The construction (hardware and software) as well as the operation of the system is described. (author). 20 figs

  15. Control and automation systems

    International Nuclear Information System (INIS)

    Schmidt, R.; Zillich, H.

    1986-01-01

    A survey is given of the development of control and automation systems for energy uses. General remarks about control and automation schemes are followed by a description of modern process control systems along with process control processes as such. After discussing the particular process control requirements of nuclear power plants the paper deals with the reliability and availability of process control systems and refers to computerized simulation processes. The subsequent paragraphs are dedicated to descriptions of the operating floor, ergonomic conditions, existing systems, flue gas desulfurization systems, the electromagnetic influences on digital circuits as well as of light wave uses. (HAG) [de

  16. Plant air systems safety study: Portsmouth Gaseous Diffusion Plant

    International Nuclear Information System (INIS)

    1982-05-01

    The Portsmouth Gaseous Diffusion Plant Air System facilities and operations are reviewed for potential safety problems not covered by standard industrial safety procedures. Information is presented under the following section headings: facility and process description (general); air plant equipment; air distribution system; safety systems; accident analysis; plant air system safety overview; and conclusion

  17. THE FORMATION OF THE CONTOUR OF THE DOCUMENTED AND REAL FLIGHT SAFETY IN THE SYSTEM OF THE INFORMATION PROVISION OF SAFETY OF FLIGHTS

    Directory of Open Access Journals (Sweden)

    B. I. Bachkalo

    2015-01-01

    Full Text Available The article discusses the principles and mechanisms of formation of the contour of the real safety of flights and contour of the documented safety, allowing us to obtain information to control fligh safety. The proposed approach can be used in the algorithms of active on-board flight safety management system for the implementation of information support to the crew in flight and automatic control of flight safety.

  18. ISABELLE control system

    International Nuclear Information System (INIS)

    Humphrey, J.W.; Frankel, R.S.; Niederer, J.A.

    1980-01-01

    Design principles for the Brookhaven ISABELLE control intersecting storage ring accelerator are described. Principal features include a locally networked console and control computer complex, a system wide process data highway, and intelligent local device controllers. Progress to date is summarized

  19. Driving Simulator study for intelligent cooperative intersection safety system (IRIS)

    NARCIS (Netherlands)

    Vreeswijk, J.; Schendzielorz, T.; Mathias, P.; Feenstra, P.

    2008-01-01

    About forty percent of all accidents occur at intersections. The Intelligent Cooperative Intersection Safety system (IRIS), as part of the European research project SAFESPOT, is a roadside application and aims at minimizing the number of accidents at controlled and uncontrolled intersections. IRIS

  20. Seismic analysis of control and safety rod drive mechanism

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  1. Control system for NPP powerfull turbines

    International Nuclear Information System (INIS)

    Osipenko, V.D.; Rozhanskij, V.E.; Rokhlenko, V.Yu.

    1985-01-01

    A control system for NPP 1000 MW turbines safety is described. The turbine safety system has a hydraulic drive to actuate in case of increasipg of rotational speed of a turbine rotor and an electrohydraulic drce to operate in case of pressure reduction in the lubrication system, axial displacement deviation, etc. The system is highly reliable due to application of a safety system without slide valves and long-term operation of hydraulic controls in guarding conditions; the system epsures multifunctional control with high accuracy and speed due to application of the intricate electronic part, high speed of response with a limited use of high pressure oil due to application of two-pressure pumps, pneumohydraulic accumulators and oil discharge valves. Steady-state serviceability of the system is maintained by devices for valve cooling dawn. A shockless change from electrohydraulic to hydraulic control channels is provided

  2. A Nuclear Safety System based on Industrial Computer

    International Nuclear Information System (INIS)

    Kim, Ji Hyeon; Oh, Do Young; Lee, Nam Hoon; Kim, Chang Ho; Kim, Jae Hack

    2011-01-01

    The Plant Protection System(PPS), a nuclear safety Instrumentation and Control (I and C) system for Nuclear Power Plants(NPPs), generates reactor trip on abnormal reactor condition. The Core Protection Calculator System (CPCS) is a safety system that generates and transmits the channel trip signal to the PPS on an abnormal condition. Currently, these systems are designed on the Programmable Logic Controller(PLC) based system and it is necessary to consider a new system platform to adapt simpler system configuration and improved software development process. The CPCS was the first implementation using a micro computer in a nuclear power plant safety protection system in 1980 which have been deployed in Ulchin units 3,4,5,6 and Younggwang units 3,4,5,6. The CPCS software was developed in the Concurrent Micro5 minicomputer using assembly language and embedded into the Concurrent 3205 computer. Following the micro computer based CPCS, PLC based Common-Q platform has been used for the ShinKori/ShinWolsong units 1,2 PPS and CPCS, and the POSAFE-Q PLC platform is used for the ShinUlchin units 1,2 PPS and CPCS. In developing the next generation safety system platform, several factors (e.g., hardware/software reliability, flexibility, licensibility and industrial support) can be considered. This paper suggests an Industrial Computer(IC) based protection system that can be developed with improved flexibility without losing system reliability. The IC based system has the advantage of a simple system configuration with optimized processor boards because of improved processor performance and unlimited interoperability between the target system and development system that use commercial CASE tools. This paper presents the background to selecting the IC based system with a case study design of the CPCS. Eventually, this kind of platform can be used for nuclear power plant safety systems like the PPS, CPCS, Qualified Indication and Alarm . Pami(QIAS-P), and Engineering Safety

  3. A Nuclear Safety System based on Industrial Computer

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Ji Hyeon; Oh, Do Young; Lee, Nam Hoon; Kim, Chang Ho; Kim, Jae Hack [Korea Electric Power Corporation Engineering and Construction, Daejeon (Korea, Republic of)

    2011-05-15

    The Plant Protection System(PPS), a nuclear safety Instrumentation and Control (I and C) system for Nuclear Power Plants(NPPs), generates reactor trip on abnormal reactor condition. The Core Protection Calculator System (CPCS) is a safety system that generates and transmits the channel trip signal to the PPS on an abnormal condition. Currently, these systems are designed on the Programmable Logic Controller(PLC) based system and it is necessary to consider a new system platform to adapt simpler system configuration and improved software development process. The CPCS was the first implementation using a micro computer in a nuclear power plant safety protection system in 1980 which have been deployed in Ulchin units 3,4,5,6 and Younggwang units 3,4,5,6. The CPCS software was developed in the Concurrent Micro5 minicomputer using assembly language and embedded into the Concurrent 3205 computer. Following the micro computer based CPCS, PLC based Common-Q platform has been used for the ShinKori/ShinWolsong units 1,2 PPS and CPCS, and the POSAFE-Q PLC platform is used for the ShinUlchin units 1,2 PPS and CPCS. In developing the next generation safety system platform, several factors (e.g., hardware/software reliability, flexibility, licensibility and industrial support) can be considered. This paper suggests an Industrial Computer(IC) based protection system that can be developed with improved flexibility without losing system reliability. The IC based system has the advantage of a simple system configuration with optimized processor boards because of improved processor performance and unlimited interoperability between the target system and development system that use commercial CASE tools. This paper presents the background to selecting the IC based system with a case study design of the CPCS. Eventually, this kind of platform can be used for nuclear power plant safety systems like the PPS, CPCS, Qualified Indication and Alarm . Pami(QIAS-P), and Engineering Safety

  4. Project Aquarius. Control of radioisotopes and safety

    Energy Technology Data Exchange (ETDEWEB)

    Post, Roy G [Department of Nuclear Engineering, University of Arizona (United States)

    1970-05-15

    The potential application of nuclear explosives to the development of water resources provides real hope for substantial increases in the availability of water from our natural water supplies. A wide range, exploratory project sponsored by the United States Atomic Energy Commission, the Bureau of Reclamation, the Arizona Atomic Energy Commission, and The University of Arizona was conducted by the Hydrology and Water Resources Office, the Department of Nuclear Engineering, and various state and federal governmental agencies in exploring the potential applications of nuclear explosives for developing water resources in the State of Arizona. The primary objective of the project was of a scouting nature, a reconnaissance effort to assess the potential for Arizona. This work, Project Aquarius, is at an early state and any significant conclusions are certainly premature. Since this is a survey, detailed analyses are not justified. Our purpose is to define limiting problems and estimate our ability to solve them. We do not seek to formulate a detailed solution until the project has been defined better. In all of the plowshare activities the primary responsibility of the Atomic Energy Commission for safety and control of not only radiological but all hazards has been well defined and documented. Thus, the work here does not reflect any opinion or voice of the Atomic Energy Commission but is based on my own views and conclusions. I have referred to the work of the various laboratories, offices, and contractors of the Atomic Energy Commission.

  5. Project Aquarius. Control of radioisotopes and safety

    International Nuclear Information System (INIS)

    Post, Roy G.

    1970-01-01

    The potential application of nuclear explosives to the development of water resources provides real hope for substantial increases in the availability of water from our natural water supplies. A wide range, exploratory project sponsored by the United States Atomic Energy Commission, the Bureau of Reclamation, the Arizona Atomic Energy Commission, and The University of Arizona was conducted by the Hydrology and Water Resources Office, the Department of Nuclear Engineering, and various state and federal governmental agencies in exploring the potential applications of nuclear explosives for developing water resources in the State of Arizona. The primary objective of the project was of a scouting nature, a reconnaissance effort to assess the potential for Arizona. This work, Project Aquarius, is at an early state and any significant conclusions are certainly premature. Since this is a survey, detailed analyses are not justified. Our purpose is to define limiting problems and estimate our ability to solve them. We do not seek to formulate a detailed solution until the project has been defined better. In all of the plowshare activities the primary responsibility of the Atomic Energy Commission for safety and control of not only radiological but all hazards has been well defined and documented. Thus, the work here does not reflect any opinion or voice of the Atomic Energy Commission but is based on my own views and conclusions. I have referred to the work of the various laboratories, offices, and contractors of the Atomic Energy Commission

  6. Development of the Digital Reactor Safety System

    International Nuclear Information System (INIS)

    Lee, Dong Young; Lee, C. K.; Hwang, I. K.

    2008-04-01

    Objectives of Project - Development of Digital Safety Grade PLC and Licensing - Development of Safety System(RPS) and Licensing - Development of Safety System(ESF-CCS) and Licensing Content and Result of Project - POSAFE-Q PLC : Development of PLC platform for Shin-UCN unit 1 and 2 ·Development Scope : Processor module, Power module, 3 kinds of Communication module, Bus extension module(Master and Slave), 16 kinds of Input and Output module ·PLC application software development tool(pSET) - IDiPS RPS and IDiPS ESF-CCS : Development of PPS for Sin-UCN 1 and 2 ·Development Scope - 4-channels RPS with the KNICS inherent architecture - A part of 1-channels ESF-CCS with the KNICS inherent architecture - Licensing ·optical Report Submitted and Expected to finish the licensing process until Aug. 2008

  7. 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.)

  8. Use of digital computing devices in systems important to safety

    International Nuclear Information System (INIS)

    1986-01-01

    The incorporation of digital computing devices in systems important to safety now is progressing fast in several countries, including Canada, France, Federal Republic of Germany, Japan, USA. There are now reactors with microprocessors in some trip systems. The major functions of those systems are: reactor trip initiation, display, monitoring, testing, re-calibration of detectors. The benefits of moving to a fully computerized shut-down system should be improved reliability, greater flexibility, better man-machine interface, improved testing, higher reactor output and lower overall cost. With the introduction of computer devices in systems important to safety, plant availability and safety are improved because disturbances are treated before they lead to safety action, in this way helping the operator to avoid errors. The Meeting presentations were divided into sessions devoted to the following topics: Needs for the use of digital devices (DCD) in safety important systems (SIS) (5 papers); Problems raised by the integration SIS in the NPP control (7 papers); Description and presentation of DCD of SIS (6 papers); Results of experiences in engineering, manufacture, qualification operation of DCD hardware and software (5 papers). A separate abstract was prepared for each of these papers

  9. Railing for safety: job demands, job control, and safety citizenship role definition.

    Science.gov (United States)

    Turner, Nick; Chmiel, Nik; Walls, Melanie

    2005-10-01

    This study investigated job demands and job control as predictors of safety citizenship role definition, that is, employees' role orientation toward improving workplace safety. Data from a survey of 334 trackside workers were framed in the context of R. A. Karasek's (1979) job demands-control model. High job demands were negatively related to safety citizenship role definition, whereas high job control was positively related to this construct. Safety citizenship role definition of employees with high job control was buffered from the influence of high job demands, unlike that of employees with low job control, for whom high job demands were related to lower levels of the construct. Employees facing both high job demands and low job control were less likely than other employees to view improving safety as part of their role orientation. Copyright (c) 2005 APA, all rights reserved.

  10. Risk-based rules for crane safety systems

    Energy Technology Data Exchange (ETDEWEB)

    Ruud, Stian [Section for Control Systems, DNV Maritime, 1322 Hovik (Norway)], E-mail: Stian.Ruud@dnv.com; Mikkelsen, Age [Section for Lifting Appliances, DNV Maritime, 1322 Hovik (Norway)], E-mail: Age.Mikkelsen@dnv.com

    2008-09-15

    The International Maritime Organisation (IMO) has recommended a method called formal safety assessment (FSA) for future development of rules and regulations. The FSA method has been applied in a pilot research project for development of risk-based rules and functional requirements for systems and components for offshore crane systems. This paper reports some developments in the project. A method for estimating target reliability for the risk-control options (safety functions) by means of the cost/benefit decision criterion has been developed in the project and is presented in this paper. Finally, a structure for risk-based rules is proposed and presented.

  11. Risk-based rules for crane safety systems

    International Nuclear Information System (INIS)

    Ruud, Stian; Mikkelsen, Age

    2008-01-01

    The International Maritime Organisation (IMO) has recommended a method called formal safety assessment (FSA) for future development of rules and regulations. The FSA method has been applied in a pilot research project for development of risk-based rules and functional requirements for systems and components for offshore crane systems. This paper reports some developments in the project. A method for estimating target reliability for the risk-control options (safety functions) by means of the cost/benefit decision criterion has been developed in the project and is presented in this paper. Finally, a structure for risk-based rules is proposed and presented

  12. A study on enforcement effects of radiation safety control regulations for diagnostic X-ray equipment

    International Nuclear Information System (INIS)

    Sung, Mo IL; Park, Myeong Hwan; Kwon, Duk Moon; Lee, Joon IL

    1999-01-01

    The purposes of this study are to analyze the realities after enforcements of safety control regulations for diagnostic X-ray equipment and to suggest means for an improvement of low radiation safety control. A questionnaire survey for medical radiologic technologists was carried out to determine enforcement effects of the safety control regulations. The results of analysis from the survey are as follows. That is, most of he respondents realized the importance of the radiation safety control system, but about a half of them revealed that regulations were not well observed in accordance with their purposes. Only 43.9 percent of the respondents took an active part in quality control of radiation. And responsibility, sex, age, and knowledge for safety control were important indicators for observations of the regulations. Training for the safety control regulations are needed to ensure safety control and proper usage of diagnostic X-ray equipment. And management of organizations using diagnostic X-ray equipment have to understand and stress the importance of radiation safety control system. (author)

  13. Advanced reactor systems: safety and regulatory aspects

    International Nuclear Information System (INIS)

    Gopalakrishnan, A.

    1994-01-01

    Safety features which are desirable in futuristic reactor systems have been the subject of several studies over the past decade by different expert groups. When one discusses this subject, therefore, in a somewhat non-specific and qualitative manner, it is best to make use of the already available collective wisdom and literature on the matter. (author). 3 refs

  14. 76 FR 14592 - Safety Management System; Withdrawal

    Science.gov (United States)

    2011-03-17

    ...), Federal Aviation Administration, 800 Independence Avenue, SW., Washington, DC 20591; telephone (202) 494...). The FAA also chartered the Safety Management System Aviation Rulemaking Committee (ARC) (Order No..., including the ANPRM. On March 31, 2010, the ARC submitted its report to the FAA. As a result of the...

  15. Maintenance of radiation safety information system

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Ho Sun [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of); Park, Moon Il; Chung, Chong Kyu; Lim, Bock Soo; Kim, Hyung Uk; Chang, Kwang Il; Nam, Kwan Hyun; Cho, Hye Ryan [AD center incubation LAB, Taejon (Korea, Republic of)

    2001-12-15

    The objectives of radiation safety information system maintenance are to maintain the requirement of users, change of job process and upgrade of the system performance stably and effectively while system maintenance. We conduct the code of conduct recommended by IAEA, management of radioisotope inventory database systematically using analysis for the state of inventory database integrated in this system. This system and database will be support the regulatory guidance, rule making and information to the MOST, KINS, other regulatory related organization and general public optimizationally.

  16. Beam based systems and controls

    CERN Document Server

    Jacquet, D

    2012-01-01

    This presentation will give a review from the operations team of the performance and issues of the beam based systems, namely RF, ADT, beam instrumentation, controls and injection systems. For each of these systems, statistics on performance and availability will be presented with the main issues encountered in 2012. The possible improvements for operational efficiency and safety will be discussed, with an attempt to answer the question "Are we ready for the new challenges brought by the 25ns beam and increased energy after LSI? ".

  17. 77 FR 11120 - Patient Safety Organizations: Voluntary Relinquishment From UAB Health System Patient Safety...

    Science.gov (United States)

    2012-02-24

    ... Organizations: Voluntary Relinquishment From UAB Health System Patient Safety Organization AGENCY: Agency for... notification of voluntary relinquishment from the UAB Health System Patient Safety Organization of its status as a Patient Safety Organization (PSO). The Patient Safety and Quality Improvement Act of 2005...

  18. Integrated control systems

    International Nuclear Information System (INIS)

    Smith, D.J.

    1991-01-01

    This paper reports that instrument manufacturers must develop standard network interfaces to pull together interrelated systems such as automatic start-up, optimization programs, and online diagnostic systems. In the past individual control system manufacturers have developed their own data highways with proprietary hardware and software designs. In the future, electric utilities will require that future systems, irrespective of manufacturer, should be able to communicate with each other. Until now the manufactures of control systems have not agreed on the standard high-speed data highway system. Currently, the Electric Power Research Institute (EPRI), in conjunction with several electric utilities and equipment manufactures, is working on developing a standard protocol for communicating between various manufacturers' control systems. According to N. Michael of Sargent and Lundy, future control room designs will require that more of the control and display functions be accessible from the control room through CRTs. There will be less emphasis on traditional hard-wired control panels

  19. Passive components of NPP safety-related systems

    International Nuclear Information System (INIS)

    Ionaytis Romuald, R.; Bubnova Tatyana, A.

    2005-01-01

    This paper presents a new passive components with having drives: fast-response cutoff valves; modular actuators with opposite cocking pneumatic drives and actuation spring drives; voting electromagnetic valve units for control of pneumatic drives; passive initiators of actuation; visual diagnostics . All these devices have been developed and tested at mock-ups. This paper presents also the following direct-action passive safety components: modular pressure-relief safety valves; pilot safety valves with passive action; check valves with remote position indicator and after-tightening; modular inserts for limiting emergency coolant flow; vortex rectifier; critical weld fasteners; gas-liquid valves; fast-removable seal assembly; seal spring loaders; grooves for increasing hydraulic resistance. Replacement of active safety system components for passive ones improves the general reliability NPP by 1.5 or 2 orders of magnitudes. (authors)

  20. Safety design integrated in the Building Delivery System

    DEFF Research Database (Denmark)

    Jørgensen, Kirsten

    2012-01-01

    phases of the building delivery system by using the principle of the lean construction modelling. The method for the research was to go through the lean construction building delivery system step by step and create a normative description of what to do, when to do and how to do to fully integration...... of safety in each process. The group of participants who created the description had a high experience in a combination of research, safety and health in general and especial in construction and knowledge of the lean construction processes both from the clients perspective as well as from the designers...... and the consultants. The result is a concept and guideline including control schemes for how to integrate safety design in the lean construction building delivery system including what to do and when. The concept has been tested in an educational context and found useful by the designers. The practical value...