Total Quality Management and the System Safety Secretary
Elliott, Suzan E.
1993-01-01
The system safety secretary is a valuable member of the system safety team. As downsizing occurs to meet economic constraints, the Total Quality Management (TQM) approach is frequently adopted as a formula for success and, in some cases, for survival.
Data Analysis of Occupational Health and Safety Management and Total Quality Management Systems
Directory of Open Access Journals (Sweden)
Ahmet Yakut
2013-01-01
Full Text Available In our study, Total Quality Management, Occupational Health and Safety on the effects of the construction industry, building sites of Istanbul evaluated with the results of the survey of 25 firms. For Occupational Health and Safety program, walked healthy, active employees in her role increased and will increase the importance of education. Due to non-implementation of the OHS system in our country enough, work-related accidents and deaths and injuries resulting from these accidents is very high. Firms as a result of the analysis, an effective health and safety management system needs to be able to fulfill their responsibilities. This system is designated as OHSAS 18001 Occupational Health and Safety Management System and the construction industry can be regarded as the imperatives.
A total safety management model
International Nuclear Information System (INIS)
Obadia, I.J.; Vidal, M.C.R.; Melo, P.F.F.F.
2002-01-01
In nuclear organizations, quality and safety are inextricably linked. Therefore, the search for excellence means reaching excellence in nuclear safety. The International Atomic Energy Agency, IAEA, developed, after the Chernobyl accident, the organizational approach for improving nuclear safety based on the safety culture, which requires a framework necessary to provide modifications in personnel attitudes and behaviors in situations related to safety. This work presents a Total Safety Management Model, based on the Model of Excellence of the Brazilian Quality Award and on the safety culture approach, which represents an alternative to this framework. The Model is currently under validation at the Nuclear Engineering Institute, in Rio de Janeiro, Brazil, and the results of its initial safety culture self assessment are also presented and discussed. (author)
Directory of Open Access Journals (Sweden)
E Mohammadfam
2015-11-01
Full Text Available Introduction: All organizations, whether public or private, necessitate performance evaluation systems in regard with growth, stability, and development in the competitive fields. One of the existing models for performance evaluation of occupational health and safety management is Total Quality Safety Management model (TQSM. Therefore, the present study aimed to evaluate performance of safety management and occupational health utilizing TQSM model. Methods: In this descriptive-analytic study, the population consisted of 16 individuals, including managers, supervisors, and members of technical protection and work health committee. Then the participants were asked to respond to TQSM questionnaire before and after the implementation of Occupational Health & Safety Advisory Services 18001 (OHSAS18001. Ultimately, the level of each program as well as the TQSM status were determined before and after the implementation of OHSAS18001. Results: The study results showed that the scores obtained by the company before OHSAS 18001’s implementation, was 43.7 out of 312. After implementing OHSAS 18001 in the company and receiving the related certificate, the total score of safety program that company could obtain was 127.12 out of 312 demonstrating a rise of 83.42 scores (26.8%. The paired t-test revealed that mean difference of TQSM scores before and after OHSAS 18001 implementation was proved to be significant (p> 0.05. Conclusion: The study findings demonstrated that TQSM can be regarded as an appropriate model in order to monitor the performance of safety management system and occupational health, since it possesses the ability to quantitatively evaluate the system performance.
Role of systems safety in maintaining affordable safety in the 1980's
International Nuclear Information System (INIS)
Hollister, H.; Trauth, C.A. Jr.
1979-01-01
Historically, the Department of Energy and its predecessors have used and supported the development of systems safety programs, practices, and principles, finding them by and large adequate, effective, and managerially efficient. Today, attempts are bing made to resolve increasingly complex environmental, safety, and health problems by turning to increasingly complex and detailed regulation as the primary governmental answer. It is increasingly doubtful that such an approach will provide management of these issues and problems that is either effective or efficient. Challenge is issued to those in systems safety to develop and apply systems safety principles and practices more broadly to total operational systems and not just to hardware and to environmental and health protection and not just to safety, so that the total universe of environmental, safety, and health can be managed effectively and efficiently with encouragement of innovation and creativity, using a relatively brief and concise, but adequate, regulatory base
International Nuclear Information System (INIS)
Takizawa, Masayuki; Sekimura, Naoto; Miyano, Hiroshi; Aoyama, Katsunobu
2012-01-01
Japanese safety regulatory body, that is, Nuclear and Industrial Safety Agency (NISA) started a 5-year national safety research project as 'the first stage' from 2006 FY to 2010 FY whose objective is 'Improve the technical information basis in order to utilize knowledge as well as information related to ageing management and maintenance of NPPs. Fukushima disaster happened in March 2011, and the priority of research needs for ageing management dramatically changed in Japan. The second-stage national project started in October 2011 with the concept of 'system safety' of NNPs where not only ageing management on degradation phenomena of important components but also safety management on total plant systems are paid attention to. The second-stage project is so called 'Japanese Ageing Management Program for System Safety (JAMPSS)'. (author)
Issues associated with a total systems approach to designing dependable systems
International Nuclear Information System (INIS)
Chisholm, G.H.
1995-01-01
A total system approach, developed by the nuclear-reactor-safety community, is extrapolated to the design of complex, critical systems. The essential properties of these systems are described, and a generic paradigm for subsequent designs is proposed
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.)
Energy Technology Data Exchange (ETDEWEB)
NONE
1978-03-01
Studies have been made on preparing criteria for the safety assuring technologies for hydrogen energy total systems. The outline of the technological guideline for hydrogen manufacturing processes in the high temperature and pressure water decomposition method is the same as that in the normal pressure water decomposition method. However, its high temperature and pressure environment can cause new safety problems. Considerations should be given on, for example, material problems in structural materials and insulation materials including electrodes and membranes, introduction of gas-liquid separation and pressure balancing devices, problems in electrolyte circulation, and safety problems that may occur because of generation of hydrogen and oxygen under high temperature and pressure conditions. This paper summarizes these matters by surveying literature data. In order to provide basic information to prepare criteria for safety assuring technologies for the gaseous hydrogen liquefaction process, surveys and studies were made based on different items of technological information and experimental study results. Safety assuring technologies were discussed on metal hydrides (promising means for storing hydrogen). Powder is used to enhance hydrogen absorbing performance, whereas the metal hydrides are pulverized as a result of repetition of absorption and discharge of hydrogen. This paper describes also metal dust explosion disaster and its risk of occurrence. (NEDO)
Safety Precautions for Total Release Foggers
Total release foggers, also known as bug bombs, are pesticide products containing aerosol propellants that release their contents at once to fumigate an area. They can pose a hazard if used incorrectly. Find safety information and videos on this page.
Food safety and total quality management
Barendsz, A.W.
1998-01-01
Food safety is a growing global concern not only because of its continuing importance for public health but also because of its impact on international trade. The application of total quality management (TQM) provides the best possible care by continuously improving products and services to meet or
Total-System Performance Assessment for the Yucca Mountain Site
International Nuclear Information System (INIS)
Wilson, M.L.
2001-01-01
Yucca Mountain, Nevada, is under consideration as a potential site for a repository for high-level radioactive waste. Total-system performance-assessment simulations are performed to evaluate the safety of the site. Features, events, and processes have been systematically evaluated to determine which ones are significant to the safety assessment. Computer models of the disposal system have been developed within a probabilistic framework, including both engineered and natural components. Selected results are presented for three different total-system simulations, and the behavior of the disposal system is discussed. The results show that risk is dominated by igneous activity at early times, because the robust waste-package design prevents significant nominal (non-disruptive) releases for tens of thousands of years or longer. The uncertainty in the nominal performance is dominated by uncertainties related to waste-package corrosion at early times and by uncertainties in the natural system, most significantly infiltration, at late times
Does the concept of safety culture help or hinder systems thinking in safety?
Reiman, Teemu; Rollenhagen, Carl
2014-07-01
The concept of safety culture has become established in safety management applications in all major safety-critical domains. The idea that safety culture somehow represents a "systemic view" on safety is seldom explicitly spoken out, but nevertheless seem to linger behind many safety culture discourses. However, in this paper we argue that the "new" contribution to safety management from safety culture never really became integrated with classical engineering principles and concepts. This integration would have been necessary for the development of a more genuine systems-oriented view on safety; e.g. a conception of safety in which human, technological, organisational and cultural factors are understood as mutually interacting elements. Without of this integration, researchers and the users of the various tools and methods associated with safety culture have sometimes fostered a belief that "safety culture" in fact represents such a systemic view about safety. This belief is, however, not backed up by theoretical or empirical evidence. It is true that safety culture, at least in some sense, represents a holistic term-a totality of factors that include human, organisational and technological aspects. However, the departure for such safety culture models is still human and organisational factors rather than technology (or safety) itself. The aim of this paper is to critically review the various uses of the concept of safety culture as representing a systemic view on safety. The article will take a look at the concepts of culture and safety culture based on previous studies, and outlines in more detail the theoretical challenges in safety culture as a systems concept. The paper also presents recommendations on how to make safety culture more systemic. Copyright © 2013 Elsevier Ltd. All rights reserved.
Passive safety system of a super fast reactor
Energy Technology Data Exchange (ETDEWEB)
Sutanto, E-mail: sutanto@fuji.waseda.jp [Cooperative Major in Nuclear Energy, Waseda University, Tokyo (Japan); Polytechnic Institute of Nuclear Technology—National Nuclear Energy Agency, Yogyakarta (Indonesia); Oka, Yoshiaki [The University of Tokyo, Tokyo (Japan)
2015-08-15
Highlights: • Passive safety system of a Super FR is proposed. • Total loss of feedwater flow and large LOCA are analyzed. • The criteria of MCST and core pressure are satisfied. - Abstract: Passive safety systems of a Super Fast Reactor are studied. The passive safety systems consist of isolation condenser (IC), automatic depressurization system (ADS), core make-up tank (CMT), gravity driven cooling system (GDCS), and passive containment cooling system (PCCS). Two accidents of total loss of feedwater flow and 100% cold-leg break large LOCA are analyzed by using the passive systems and the criteria of maximum cladding surface temperature (MCST) and maximum core pressure are satisfied. The isolation condenser can be used for mitigation of the accident of total loss of feedwater flow at both supercritical and subcritical pressures. The ADS is used for depressurization leading to a loss of coolant during line switching to operation of the isolation condenser at subcritical pressure. Use of CMT during line switching recovers the lost coolant. In case of large LOCA, GDCS can be used for core reflooding. Coolant vaporization in the core released to containment through the break is condensed by passive containment cooling system. The condensate flows to the GDCS pool by gravity force. The maximum cladding surface temperature (MCST) of the accident satisfies the criterion.
The reliability of nuclear power plant safety systems
International Nuclear Information System (INIS)
Susnik, J.
1978-01-01
A criterion was established concerning the protection that nuclear power plant (NPP) safety systems should afford. An estimate of the necessary or adequate reliability of the total complex of safety systems was derived. The acceptable unreliability of auxiliary safety systems is given, provided the reliability built into the specific NPP safety systems (ECCS, Containment) is to be fully utilized. A criterion for the acceptable unreliability of safety (sub)systems which occur in minimum cut sets having three or more components of the analysed fault tree was proposed. A set of input MTBF or MTTF values which fulfil all the set criteria and attain the appropriate overall reliability was derived. The sensitivity of results to input reliability data values was estimated. Numerical reliability evaluations were evaluated by the programs POTI, KOMBI and particularly URSULA, the last being based on Vesely's kinetic fault tree theory. (author)
International Nuclear Information System (INIS)
Takano, Ken-ichi; Hasegawa, Naoko; Hirose, Ayako; Hayase, Ken-ichi
2004-01-01
For past five years, CRIEPI has been continuing efforts to develop and make applications of a 'safety assessment system' which enable to measure the safety level of organization. This report describe about frame of the system, assessment results and its reliability, and relation between labor accident rate in the site and total safety index (TSI), which can be obtained by the principal factors analysis. The safety assessment in this report is based on questionnaire survey of employee. The format and concrete questionnaires were developed using existing literatures including organizational assessment tools. The tailored questionnaire format involved 124 questionnaire items. The assessment results could be considered as a well indicator of the safety level of organization, safety management, and safety awareness of employee. (author)
Safety status system for operating room devices.
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.
System safety education focused on flight safety
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.
Software Safety Risk in Legacy Safety-Critical Computer Systems
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.
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.
30 CFR 7.103 - Safety system control test.
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...
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
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
Krumsvik, Ole Andreas; Babic, Ankica
2017-01-01
This paper presents a safety reporting smartphone application which is expected to reduce the occurrence of postoperative adverse events after total hip arthroplasty (THA). A user-centered design approach was utilized to facilitate optimal user experience. Two main implemented functionalities capture patient pain levels and well-being, the two dimensions of patient status that are intuitive and commonly checked. For these and other functionalities, mobile technology could enable timely safety reporting and collection of patient data out of a hospital setting. The HCI expert, and healthcare professionals from the Haukeland University Hospital in Bergen have assessed the design with respect to the interaction flow, information content, and self-reporting functionalities. They have found it to be practical, intuitive, sufficient and simple for users. Patient self-reporting could help recognizing safety issues and adverse events.
Total safety management: An approach to improving safety culture
International Nuclear Information System (INIS)
Blush, S.M.
1993-01-01
A little over 4 yr ago, Admiral James D. Watkins became Secretary of Energy. President Bush, who had appointed him, informed Watkins that his principal task would be to clean up the nuclear weapons complex and put the US Department of Energy (DOE) back in the business of producing tritium for the nation's nuclear deterrent. Watkins recognized that in order to achieve these objectives, he would have to substantially improve the DOE's safety culture. Safety culture is a relatively new term. The International Atomic Energy Agency (IAEA) used it in a 1986 report on the root causes of the Chernobyl nuclear accident. In 1990, the IAEA's International Nuclear Safety Advisory Group issued a document focusing directly on safety culture. It provides guidelines to the international nuclear community for measuring the effectiveness of safety culture in nuclear organizations. Safety culture has two principal aspects: an organizational framework conducive to safety and the necessary organizational and individual attitudes that promote safety. These obviously go hand in hand. An organization must create the right framework to foster the right attitudes, but individuals must have the right attitudes to create the organizational framework that will support a good safety culture. The difficulty in developing such a synergistic relationship suggests that achieving and sustaining a strong safety culture is not easy, particularly in an organization whose safety culture is in serious disrepair
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
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.
Traceability of Software Safety Requirements in Legacy Safety Critical Systems
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?
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
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.
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
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)
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
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
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)
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
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
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.
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.
Passive safety systems reliability and integration of these systems in nuclear power plant PSA
International Nuclear Information System (INIS)
La Lumia, V.; Mercier, S.; Marques, M.; Pignatel, J.F.
2004-01-01
Innovative nuclear reactor concepts could lead to use passive safety features in combination with active safety systems. A passive system does not need active component, external energy, signal or human interaction to operate. These are attractive advantages for safety nuclear plant improvements and economic competitiveness. But specific reliability problems, linked to physical phenomena, can conduct to stop the physical process. In this context, the European Commission (EC) starts the RMPS (Reliability Methods for Passive Safety functions) program. In this RMPS program, a quantitative reliability evaluation of the RP2 system (Residual Passive heat Removal system on the Primary circuit) has been realised, and the results introduced in a simplified PSA (Probabilistic Safety Assessment). The scope is to get out experience of definition of characteristic parameters for reliability evaluation and PSA including passive systems. The simplified PSA, using event tree method, is carried out for the total loss of power supplies initiating event leading to a severe core damage. Are taken into account: failures of components but also failures of the physical process involved (e.g. natural convection) by a specific method. The physical process failure probabilities are assessed through uncertainty analyses based on supposed probability density functions for the characteristic parameters of the RP2 system. The probabilities are calculated by MONTE CARLO simulation coupled to the CATHARE thermalhydraulic code. The yearly frequency of the severe core damage is evaluated for each accident sequence. This analysis has identified the influence of the passive system RP2 and propose a re-dimensioning of the RP2 system in order to satisfy the safety probabilistic objectives for reactor core severe damage. (authors)
NASA System Safety Handbook. Volume 1; System Safety Framework and Concepts for Implementation
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
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
How could intelligent safety transport systems enhance safety ?
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
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
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
Comprehensive Lifecycle for Assuring System Safety
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.
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
A Methodological Framework for Software Safety in Safety Critical Computer Systems
P. V. Srinivas Acharyulu; P. Seetharamaiah
2012-01-01
Software safety must deal with the principles of safety management, safety engineering and software engineering for developing safety-critical computer systems, with the target of making the system safe, risk-free and fail-safe in addition to provide a clarified differentaition for assessing and evaluating the risk, with the principles of software risk management. Problem statement: Prevailing software quality models, standards were not subsisting in adequately addressing the software safety ...
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
Evaluating the effectiveness of active vehicle safety systems.
Jeong, Eunbi; Oh, Cheol
2017-03-01
Advanced vehicle safety systems have been widely introduced in transportation systems and are expected to enhance traffic safety. However, these technologies mainly focus on assisting individual vehicles that are equipped with them, and less effort has been made to identify the effect of vehicular technologies on the traffic stream. This study proposed a methodology to assess the effectiveness of active vehicle safety systems (AVSSs), which represent a promising technology to prevent traffic crashes and mitigate injury severity. The proposed AVSS consists of longitudinal and lateral vehicle control systems, which corresponds to the Level 2 vehicle automation presented by the National Highway Safety Administration (NHTSA). The effectiveness evaluation for the proposed technology was conducted in terms of crash potential reduction and congestion mitigation. A microscopic traffic simulator, VISSIM, was used to simulate freeway traffic stream and collect vehicle-maneuvering data. In addition, an external application program interface, VISSIM's COM-interface, was used to implement the AVSS. A surrogate safety assessment model (SSAM) was used to derive indirect safety measures to evaluate the effectiveness of the AVSS. A 16.7-km freeway stretch between the Nakdong and Seonsan interchanges on Korean freeway 45 was selected for the simulation experiments to evaluate the effectiveness of AVSS. A total of five simulation runs for each evaluation scenario were conducted. For the non-incident conditions, the rear-end and lane-change conflicts were reduced by 78.8% and 17.3%, respectively, under the level of service (LOS) D traffic conditions. In addition, the average delay was reduced by 55.5%. However, the system's effectiveness was weakened in the LOS A-C categories. Under incident traffic conditions, the number of rear-end conflicts was reduced by approximately 9.7%. Vehicle delays were reduced by approximately 43.9% with 100% of market penetration rate (MPR). These results
International Nuclear Information System (INIS)
Wetzel, N.; Scharfe, A.
1998-01-01
The present report examines the possibilities and limits of a probabilistic safety analysis to evaluate passive safety systems and inherent safety characteristics. The inherent safety characteristics are based on physical principles, that together with the safety system lead to no damage. A probabilistic evaluation of the inherent safety characteristic is not made. An inventory of passive safety systems of accomplished nuclear power plant types in the Federal Republic of Germany was drawn up. The evaluation of the passive safety system in the analysis of the accomplished nuclear power plant types was examined. The analysis showed that the passive manner of working was always assumed to be successful. A probabilistic evaluation was not performed. The unavailability of the passive safety system was determined by the failure of active components which are necessary in order to activate the passive safety system. To evaluate the passive safety features in new concepts of nuclear power plants the AP600 from Westinghouse, the SBWR from General Electric and the SWR 600 from Siemens, were selected. Under these three reactor concepts, the SWR 600 is specially attractive because the safety features need no energy sources and instrumentation in this concept. First approaches for the assessment of the reliability of passively operating systems are summarized. Generally it can be established that the core melt frequency for the passive concepts AP600 and SBWR is advantageous in comparison to the probabilistic objectives from the European Pressurized Water Reactor (EPR). Under the passive concepts is the SWR 600 particularly interesting. In this concept the passive systems need no energy sources and instrumentation, and has active operational systems and active safety equipment. Siemens argues that with this concept the frequency of a core melt will be two orders of magnitude lower than for the conventional reactors. (orig.) [de
The aviation safety reporting system
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.
NASA Aviation Safety Reporting System (ASRS)
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.
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
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
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
International Nuclear Information System (INIS)
Van Luik, Abraham E.
2004-01-01
The session started with Abe Van Luik (IGSC Chair, US-DOE-YM, USA) who presented the feedback of the international peer review of the US-DOE Yucca Mountain TSPA (Total System Performance Assessment) supporting the successful designation of the site by the Congress and the President of the U.S. In particular, he listed key implications of the IRT (International Review team) recommendations on the forthcoming US-DOE documentation of its case for safety to be submitted to the regulator, the U.S. Nuclear Regulatory Commission, mainly: - The documentation submitted to the licensing authority should address technical aspects and compliance with regulatory criteria. - That documentation should reflect sound science and good engineering practice; it should present detailed and rigorous modelling. - In addition, it should present both quantitative and qualitative arguments, make a statement on why there can be confidence in the face of uncertainty, acknowledge remaining issues and provide the strategy to resolve them. - Demonstrating understanding is as important as demonstrating compliance. - There is a need to provide a clear explanation of the case made to the regulator for more general audiences to complement the large amount of technical documents that will be produced. The US-DOE response to these recommendations for the License Application, which is under preparation, is that the recommendations will be implemented to the maximum extent possible. In subsequent discussion, with respect to the License Application, it was acknowledged that detailed guidance from the U.S. regulator was very useful, and guidance of this type would be generally useful. At the current time, the words 'safety case' are not mentioned in U.S. regulations, but if one reads both the regulation and guidance documents it becomes evident that all aspects of a safety case need to be provided in the License Application and its accompanying documents
Jefferson Lab IEC 61508/61511 Safety PLC Based Safety System
International Nuclear Information System (INIS)
Mahoney, Kelly; Robertson, Henry
2009-01-01
This paper describes the design of the new 12 GeV Upgrade Personnel Safety System (PSS) at the Thomas Jefferson National Accelerator Facility (TJNAF). The new PSS design is based on the implementation of systems designed to meet international standards IEC61508 and IEC 61511 for programmable safety systems. In order to meet the IEC standards, TJNAF engineers evaluated several SIL 3 Safety PLCs before deciding on an optimal architecture. In addition to hardware considerations, software quality standards and practices must also be considered. Finally, we will discuss R and D that may lead to both high safety reliability and high machine availability that may be applicable to future accelerators such as the ILC.
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)
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.
Integrating system safety into the basic systems engineering process
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.
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
The contribution of safety issues to public perceptions of energy systems
International Nuclear Information System (INIS)
Otway, H.J.; Thomas, Kerry
1978-01-01
Public opposition is an important consideration for those responsible for energy planning. An attitude model was applied to identify the underlying determinants of public perceptions of five energy systems: nuclear, coal, oil, solar and hydro. Empirical results are reported in which these energy systems were found to be perceived in terms of four basic dimensions: psychological aspects; economics benefits; socio-political implications; environmental and physical safety issues. For the total sample, safety issues made an appreciable contribution to attitudes toward all of the systems except nuclear energy, where it was not significant. A differential analysis of two sub-samples, those respondents PRO and CON nuclear energy, showed that benefits and safety issues were important determinants of PRO attitudes while CON attitudes were primarily due to psychological aspects and concerns about personal and political power. The role of technical information in the formation of public attitudes toward technological policies is discussed [fr
A concept of JAERI passive safety light water reactor system (JPSR)
Energy Technology Data Exchange (ETDEWEB)
Murao, Y.; Araya, F.; Iwamura, T. [Japan Atomic Energy Research Institute, Tokai-mura (Japan)
1995-09-01
The Japan Atomic Energy Research Institute (JAERI) proposed a passive safety reactor system concept, JPSR, which was developed for reducing manpower in operation and maintenance and influence of human errors on reactor safety. In the concept the system was extremely simplified. The inherent matching nature of core generation and heat removal rate within a small volume change of the primary coolant is introduced by eliminating chemical shim and adopting in-vessel control rod drive mechanism units, a low power density core and once-through steam generators. In order to simplify the system, a large pressurizer, canned pumps, passive engineered-safety-features-system (residual heat removal system and coolant injection system) are adopted and the total system can be significantly simplified. The residual heat removal system is completely passively actuated in non-LOCAs and is also used for depressurization of the primary coolant system to actuate accumulators in small break LOCAs and reactor shutdown cooling system in normal operation. All of systems for nuclear steam supply system are built in the containment except for the air coolers as a the final heat sink of the passive residual heat removal system. Accordingly the reliability of the safety system and the normal operation system is improved, since most of residual heat removal system is always working and a heat sink for normal operation system is {open_quotes}safety class{close_quotes}. In the passive coolant injection system, depressurization of the primary cooling system by residual heat removal system initiates injection from accumulators designed for the MS-600 in medium pressure and initiates injection from the gravity driven coolant injection pool at low pressure. Analysis with RETRAN-02/MOD3 code demonstrated the capability of passive load-following, self-power-controllability, cooling and depressurization.
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)
System safety engineering analysis handbook
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.
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.
2013-05-20
... NUCLEAR REGULATORY COMMISSION [NRC-2013-0098] Embedded Digital Devices in Safety-Related Systems, Systems Important to Safety, and Items Relied on for Safety AGENCY: Nuclear Regulatory Commission. ACTION... (NRC) is issuing for public comment Draft Regulatory Issue Summary (RIS) 2013-XX, ``Embedded Digital...
The Evolution of System Safety at NASA
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.
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
Safety system function trend indicator: Theory and test application
International Nuclear Information System (INIS)
Azarm, M.A.; Carbonaro, J.F.; Boccio, J.L.; Vesely, W.E.
1989-01-01
The purpose of this paper is to summarize research conducted on the development and validation of quantitative indicators of safety performance. This work, performed under the Risk-Based Performance Indicator (RBPI) Project, FIN A-3295, for the Office of Research (RES), is considered part of NRC's Performance Indicator Program which is being coordinated through the Office for the Analysis and Evaluation of Operational Data (AEOD). The program originally focused on risk-based indicators at high levels of safety indices (e.g., core-damage frequency, functional unavailabilities, and sequence monitoring). The program was then redirected towards a more amenable goal, safety system unavailability indicators, mainly due to the lack of PRA models and plant data. In that regard, BNL published a technical report that introduced the concept of cycle-based indicators and also described various alternatives of monitoring safety system unavailabilities. Further simplification of these indicators was requested by NRC to facilitate their applications to all plants in a timely manner. This resulted in the development of Safety System Function Trend (SSFT) indicators which minimize the need for detailed system model as well as component history. The theoretical bases for these indicators were developed through various simulation studies to determine the ease of detecting a trend and/or unacceptable performance. These indicators, along with several other indicators, were then generated and compared using plant data as a part of a test application. The SSFT indicators, specifically, were constructed for a total of eight plants, consisting of two systems per plant. Emphasis was placed on examining relative changes, as well as the indicator's actual level. Both the trend and actual indicator level were found to be important in identifying plants with potential problems
77 FR 70409 - System Safety Program
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...
Simulation study of coal mine safety investment based on system dynamics
Institute of Scientific and Technical Information of China (English)
Tong Lei; Dou Yuanyuan
2014-01-01
To generate dynamic planning for coal mine safety investment, this study applies system dynamics to decision-making, classifying safety investments by accident type. It validates the relationship between safety investments and accident cost, by structurally analyzing the causality between safety investments and their influence factors. Our simulation model, based on Vensim software, conducts simulation anal-ysis on a series of actual data from a coalmine in Shanxi Province. Our results indicate a lag phase in safety investments, and that increasing pre-phase safety investment reduces accident costs. We found that a 24%increase in initial safety investment could help reach the target accident costs level 14 months earlier. Our simulation test included nine kinds of variation trends of accident costs brought by different investment ratios on accident prevention. We found an optimized ratio of accident prevention invest-ments allowing a mine to reach accident cost goals 4 months earlier, without changing its total investment.
Modelling safety of multistate systems with ageing components
Energy Technology Data Exchange (ETDEWEB)
Kołowrocki, Krzysztof; Soszyńska-Budny, Joanna [Gdynia Maritime University, Department of Mathematics ul. Morska 81-87, Gdynia 81-225 Poland (Poland)
2016-06-08
An innovative approach to safety analysis of multistate ageing systems is presented. Basic notions of the ageing multistate systems safety analysis are introduced. The system components and the system multistate safety functions are defined. The mean values and variances of the multistate systems lifetimes in the safety state subsets and the mean values of their lifetimes in the particular safety states are defined. The multi-state system risk function and the moment of exceeding by the system the critical safety state are introduced. Applications of the proposed multistate system safety models to the evaluation and prediction of the safty characteristics of the consecutive “m out of n: F” is presented as well.
Modelling safety of multistate systems with ageing components
International Nuclear Information System (INIS)
Kołowrocki, Krzysztof; Soszyńska-Budny, Joanna
2016-01-01
An innovative approach to safety analysis of multistate ageing systems is presented. Basic notions of the ageing multistate systems safety analysis are introduced. The system components and the system multistate safety functions are defined. The mean values and variances of the multistate systems lifetimes in the safety state subsets and the mean values of their lifetimes in the particular safety states are defined. The multi-state system risk function and the moment of exceeding by the system the critical safety state are introduced. Applications of the proposed multistate system safety models to the evaluation and prediction of the safty characteristics of the consecutive “m out of n: F” is presented as well.
International Nuclear Information System (INIS)
Ohba, Hisateru; Ogasawara, Katsuhiko; Aburano, Tamio
2005-01-01
In this study, a questionnaire survey was carried out to determine the actual situation of radiation safety management systems in Japanese medical institutions with nuclear medicine facilities. The questionnaire consisted of questions concerning the Radiation Protection Supervisor license, safety management organizations, and problems related to education and training in safety management. Analysis was conducted according to region, type of establishment, and number of beds. The overall response rate was 60%, and no significant difference in response rate was found among regions. Medical institutions that performed nuclear medicine practices without a radiologist participating accounted for 10% of the total. Medical institutions where nurses gave patients intravenous injections of radiopharmaceuticals as part of the nuclear medicine practices accounted for 28% of the total. Of these medical institutions, 59% provided education and training in safety management for nurses. The rate of acquisition of Radiation Protection Supervisor licenses was approximately 70% for radiological technologists and approximately 20% for physicians (regional difference, p=0.02). The rate of medical institutions with safety management organizations was 71% of the total. Among the medical institutions (n=208) without safety management organizations, approximately 56% had 300 beds or fewer. In addition, it became clear that 35% of quasi-public organizations and 44% of private organizations did not provide education and training in safety management (p<0.001, according to establishment). (author)
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
Contribution of safety issues to public perceptions of energy systems
International Nuclear Information System (INIS)
Otway, H.J.; Thomas, K.
1978-01-01
Public opposition is an important consideration for those responsible for energy planning; however, the formulation of socially viable policies requires an understanding of the reasons for this opposition. An attitude model was applied to identify the underlying determinants of public perceptions of five energy systems: nuclear, coal, oil, solar and hydro. Empirical results (heterogeneous sample of the general public, N = 224) are reported in which these energy systems were found to be perceived in terms of four basic dimensions: psychological aspects; economic benefits; socio-political implications; environmental and physical safety issues. For the total sample, safety issues made an appreciable contribution to attitudes toward all of the systems except nuclear energy, where it was not significant. A differential analysis of two sub-samples, those respondents PRO and CON nuclear energy, showed that benefits and safety issues were important determinants of PRO attitudes while CON attitudes were primarily due to psychological aspects and concerns about personal and political power. The role of technical information in the formation of public attitudes toward technological policies is discussed
System safety education focused on industrial engineering
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.
The WIPP transportation system: Dedicated to safety
International Nuclear Information System (INIS)
Ward, T.; McFadden, M.
1993-01-01
When developing a transportation system to transport transuranic (TRU) waste from ten widely-dispersed generator sites, the Department of Energy (DOE) recognized and addressed many challenges. Shipments of waste to the Waste Isolation Pilot Plant (WIPP) were to cover a twenty-five year period and utilize routes covering over twelve thousand miles in twenty-three states. Enhancing public safety by maximizing the payload, thus reducing the number of shipments, was the primary objective. To preclude the requirement for overweight permits, the DOE started with a total shipment weight limit of 80,000 pounds and developed an integrated transportation system consisting of a Type ''B'' package to transport the material, a lightweight tractor and trailer, stringent driver requirements, and a shipment tracking system referred to as ''TRANSCOM''
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
International Nuclear Information System (INIS)
Moreno Chamorro, P.; Gallego Diaz, C.
2011-01-01
The main objective of this work is to show the current status of the implementation of integrated analysis of safety (ISA) methodology and its SCAIS associated tool (system of simulation codes for ISA) to the sequence analysis of total loss of feedwater in a PWR reactor model Westinghouse of three loops with large, dry containment.
Systems Safety and Engineering Division
Federal Laboratory Consortium — Volpe's Systems Safety and Engineering Division conducts engineering, research, and analysis to improve transportation safety, capacity, and resiliency. We provide...
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.
International Nuclear Information System (INIS)
Matzie, Regis A.; Carpentino, Frederick L.; Robertson, James E.
1996-01-01
Safely systems in the System 80+ TM Standard Plant are designed with more redundancy, diversity and simplicity than earlier nuclear power plant designs. These gains were accomplished by an evolutionary process that preserved the desirable and proven features in currently operating nuclear plants, while improving reliability and defense-in-depth. The System 80+ safety systems are the primary contributors to a core damage frequency that is more than 100 times lower than 1980's vintage U. S. designs, including the predecessor System 80 R standard nuclear steam supply system (NSSS) design. The System 80+ design includes significant improvements to the safety injection system, emergency feedwater system, shutdown cooling system, containment spray system, reactor coolant gas vent system, and to their vital support systems. These improvements enhance performance for traditional design basis events and significantly reduce the probability of a severe accident. The System 80+ design also incorporates safety systems to mitigate a severe accident. The added systems include the rapid depressurization system, the in-containment refueling water storage tank, the cavity flooding system. These systems fully address the U. S. Nuclear Regulatory Commission's (US NRC) severe accident policy. The System 80+ safety systems are integrated with the System 80+ Nuclear Island (NI) design. The NI general arrangement provides quadrant separation of the safety systems for protection from fire and flooding, and large equipment pull spaces and lay down areas for maintenance. This paper will describe the System 80+ safety systems advanced design features, the improved accident prevention and mitigation capabilities, and startup, operating and maintenance benefits
Dependability analysis of proposed I and C architecture for safety systems of a large PWR
International Nuclear Information System (INIS)
Kabra, Ashutosh; Karmakar, G.; Tiwari, A.P.; Manoj Kumar; Marathe, P.P.
2014-01-01
Instrumentation and Control (I and C) systems in a reactor provide protection against unsafe operation during steady-state and transient power operations. Indian reactors traditionally adopted 2-out-of-3 (2oo3) architecture for safety systems. But, contemporary reactor safety systems are employing 2-out-of-4 (2oo4) architecture in spite of the increased cost due to the additional channel. This motivated us to carry out a comparative study of 2oo3 and 2oo4 architecture, especially for their dependability attributes - safety and availability. Quantitative estimation of safety and availability has been used to adjudge the worthiness of adopting 2oo4 architecture in I and C safety systems of a large PWR. Our analysis using Markov model shows that 2oo4 architecture, even with lower diagnostic coverage and longer proof test interval, can provide better safety and availability in comparison of 2oo3 architecture. This reduces total life cycle cost of system during development phase and complexity and frequency of surveillance test during operational phase. The paper also describes the proposed architecture for Reactor Protection System (RPS), a representative safety system, and determines its dependability using Markov analysis and Failure Mode Effect Analysis (FMEA). The proposed I and C safety system architecture also has been qualitatively analyzed for their effectiveness against common cause failures (CCFs). (author)
Product Engineering Class in the Software Safety Risk Taxonomy for Building Safety-Critical Systems
Hill, Janice; Victor, Daniel
2008-01-01
When software safety requirements are imposed on legacy safety-critical systems, retrospective safety cases need to be formulated as part of recertifying the systems for further use and risks must be documented and managed to give confidence for reusing the systems. The SEJ Software Development Risk Taxonomy [4] focuses on general software development issues. It does not, however, cover all the safety risks. The Software Safety Risk Taxonomy [8] was developed which provides a construct for eliciting and categorizing software safety risks in a straightforward manner. In this paper, we present extended work on the taxonomy for safety that incorporates the additional issues inherent in the development and maintenance of safety-critical systems with software. An instrument called a Software Safety Risk Taxonomy Based Questionnaire (TBQ) is generated containing questions addressing each safety attribute in the Software Safety Risk Taxonomy. Software safety risks are surfaced using the new TBQ and then analyzed. In this paper we give the definitions for the specialized Product Engineering Class within the Software Safety Risk Taxonomy. At the end of the paper, we present the tool known as the 'Legacy Systems Risk Database Tool' that is used to collect and analyze the data required to show traceability to a particular safety standard
Safety of retransfusing shed blood after local infiltration analgesia in total knee arthroplasty
Thomassen, B.J.; Pool, L.; Van Der Flier, R.; Stienstra, R.; in 't Veld, B.A.
2012-01-01
We investigated the safety of LIA (local infiltration analgesia) combined with retransfusion of drained blood. Total knee arthroplasty patients received two peri-articular injections during surgery followed by continuous infusion, both with ropivacaine (567 mg). Ropivacaine plasma concentrations
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.
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
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
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
Safety systems and safety analysis of the Qinshan phase III CANDU nuclear power plant
International Nuclear Information System (INIS)
Cai Jianping; Shen Sen; Barkman, N.
1999-01-01
The author introduces the Canadian nuclear reactor safety philosophy and the Qinshan Phase III CANDU NPP safety systems and safety analysis, which are designed and performed according to this philosophy. The concept of 'defence-in-depth' is a key element of the Canadian nuclear reactor safety philosophy. The design concepts of redundancy, diversity, separation, equipment qualification, quality assurance, and use of appropriate design codes and standards are adopted in the design. Four special safety systems as well as a set of reliable safety support systems are incorporated in the design of Qinshan phase III CANDU for accident mitigation. The assessment results for safety systems performance show that the fundamental safety criteria for public dose, and integrity of fuel, channels and the reactor building, are satisfied
Pius, self-protective thermohydraulics transient without safety system intervention
International Nuclear Information System (INIS)
Fredell, J.; Bredolt, V.
1989-01-01
In this paper, the self-protective thermohydraulic feedback of the PIUS reactor system is illustrated by an in-depth discussion of one typical transient. The selected transient is an undetected total loss of feedwater in the complete absence of conventional safety system intervention. The reactor shuts itself down to residual power in two steps. First, the power decreases due to the strongly negative moderator temperature reactivity coefficient, and then a complete shutdown occurs by ingress of cold, highly borated water from the reactor pool. The transient is terminated without any harm to the fuel or paint systems
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.
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
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...
5th Total System Performance Assessment Workshop
International Nuclear Information System (INIS)
Hwang, Yong Soo; Lee, Youn Myoung; Kang, Chul Hyung; Lee, Sung Ho
2009-07-01
Research items on safety assessment of high-level waste repository have been proposed by external invited experts outside KAERI and discussed extensively during the annual 5th performance assessment workshop prepared by safety assessment group in KAERI. This could be useful to set up R and D plans necessary for the next phase of mid- and long-term reaserch area regarding the safety assessment of high-level waste repository. Through the research and the presentation, HLW-related research and development area including such specific research items as current status of HLW safety assessment research, current requirement for the licensing of the repository system, priority on research area, data base building for the safety assessment, source-term modeling as well as safety case, among many others, have been discussed and summarized
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
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
Directory of Open Access Journals (Sweden)
M. S. Poursoleiman
2015-09-01
Full Text Available Introduction: Work-related accidents may cause damage to people, environment and lead to waste of time and money. Health, Safety and Environment Management System has been developed in order to reduce accidents. This study aimed to investigate the effect of implementation of this system on reduction of the accidents and its consequences and also on the safety performance indices in Kermanshah Petrochemical Company. Material and Method: In this study, records of accidents were collected by OSHA incident report form 301 over 4 years. Following, the mean annual accidents and its consequences and safety performance indices were calculated and reported. Then, using statistical analysis, the impacts of two years implementation of this system on the accidents and its consequences and safety performance indices were evaluated. Result: The results showed that the implementation of HSE system was significantly correlated with Frequency Severity Indicator, Accident Severity Rate, lost days, minor accidents and total incidents (P-value 0.05. Conclusion: The implementation of Health, Safety and the Environment Management System caused a reduction in accidents and its consequences and most of the safety performance indices in the entire process cycle of Kermanshah Petrochemical Company. Overall, safety condition has been improved considerably.
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.
Industrial Personal Computer based Display for Nuclear Safety System
International Nuclear Information System (INIS)
Kim, Ji Hyeon; Kim, Aram; Jo, Jung Hee; Kim, Ki Beom; Cheon, Sung Hyun; Cho, Joo Hyun; Sohn, Se Do; Baek, Seung Min
2014-01-01
The safety display of nuclear system has been classified as important to safety (SIL:Safety Integrity Level 3). These days the regulatory agencies are imposing more strict safety requirements for digital safety display system. To satisfy these requirements, it is necessary to develop a safety-critical (SIL 4) grade safety display system. This paper proposes industrial personal computer based safety display system with safety grade operating system and safety grade display methods. The description consists of three parts, the background, the safety requirements and the proposed safety display system design. The hardware platform is designed using commercially available off-the-shelf processor board with back plane bus. The operating system is customized for nuclear safety display application. The display unit is designed adopting two improvement features, i.e., one is to provide two separate processors for main computer and display device using serial communication, and the other is to use Digital Visual Interface between main computer and display device. In this case the main computer uses minimized graphic functions for safety display. The display design is at the conceptual phase, and there are several open areas to be concreted for a solid system. The main purpose of this paper is to describe and suggest a methodology to develop a safety-critical display system and the descriptions are focused on the safety requirement point of view
Industrial Personal Computer based Display for Nuclear Safety System
Energy Technology Data Exchange (ETDEWEB)
Kim, Ji Hyeon; Kim, Aram; Jo, Jung Hee; Kim, Ki Beom; Cheon, Sung Hyun; Cho, Joo Hyun; Sohn, Se Do; Baek, Seung Min [KEPCO, Youngin (Korea, Republic of)
2014-08-15
The safety display of nuclear system has been classified as important to safety (SIL:Safety Integrity Level 3). These days the regulatory agencies are imposing more strict safety requirements for digital safety display system. To satisfy these requirements, it is necessary to develop a safety-critical (SIL 4) grade safety display system. This paper proposes industrial personal computer based safety display system with safety grade operating system and safety grade display methods. The description consists of three parts, the background, the safety requirements and the proposed safety display system design. The hardware platform is designed using commercially available off-the-shelf processor board with back plane bus. The operating system is customized for nuclear safety display application. The display unit is designed adopting two improvement features, i.e., one is to provide two separate processors for main computer and display device using serial communication, and the other is to use Digital Visual Interface between main computer and display device. In this case the main computer uses minimized graphic functions for safety display. The display design is at the conceptual phase, and there are several open areas to be concreted for a solid system. The main purpose of this paper is to describe and suggest a methodology to develop a safety-critical display system and the descriptions are focused on the safety requirement point of view.
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)
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.)
Safer Systems: A NextGen Aviation Safety Strategic Goal
Darr, Stephen T.; Ricks, Wendell R.; Lemos, Katherine A.
2008-01-01
The Joint Planning and Development Office (JPDO), is charged by Congress with developing the concepts and plans for the Next Generation Air Transportation System (NextGen). The National Aviation Safety Strategic Plan (NASSP), developed by the Safety Working Group of the JPDO, focuses on establishing the goals, objectives, and strategies needed to realize the safety objectives of the NextGen Integrated Plan. The three goal areas of the NASSP are Safer Practices, Safer Systems, and Safer Worldwide. Safer Practices emphasizes an integrated, systematic approach to safety risk management through implementation of formalized Safety Management Systems (SMS) that incorporate safety data analysis processes, and the enhancement of methods for ensuring safety is an inherent characteristic of NextGen. Safer Systems emphasizes implementation of safety-enhancing technologies, which will improve safety for human-centered interfaces and enhance the safety of airborne and ground-based systems. Safer Worldwide encourages coordinating the adoption of the safer practices and safer systems technologies, policies and procedures worldwide, such that the maximum level of safety is achieved across air transportation system boundaries. This paper introduces the NASSP and its development, and focuses on the Safer Systems elements of the NASSP, which incorporates three objectives for NextGen systems: 1) provide risk reducing system interfaces, 2) provide safety enhancements for airborne systems, and 3) provide safety enhancements for ground-based systems. The goal of this paper is to expose avionics and air traffic management system developers to NASSP objectives and Safer Systems strategies.
Using Total Lightning Observations to Enhance Lightning Safety
Stano, Geoffrey T.
2012-01-01
. SPoRT has been collaborating with the Huntsville National Weather Service (NWS) Office since 2003 and has since included several other offices to better implement LMA observations into real-time applications. Much of that work has focused on the LMA s ability to detect intra-cloud lightning in addition to cloud-to-ground lightning strikes. Combined, these observations are called total lightning. With total lightning observations, NWS offices can enhance their situational awareness and improve severe weather warnings. Just as importantly, the observed intra-cloud flashes often precede the first cloud-to-ground strike by a few minutes. SPoRT and its partner NWS offices are working to develop visualizations and applications to better utilize these data. However, there is a drawback. The LMAs have a short range of no more than 200 km. This is being addressed with the next generation geostationary satellite, GOES-R, which will boast the Geostationary Lightning Mapper (GLM). SPoRT, in conjunction with NOAA s GOES-R Proving Ground, is working to prepare the end user community for the GLM era using the LMA observations as a demonstration tool. Working collaboratively with our NWS partners, SPoRT is working to determine how best to integrate these future observations to improve both severe storm warnings and lightning safety.
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
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
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.
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...
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
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.
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.
The Road to Total Earthquake Safety
Frohlich, Cliff
Cinna Lomnitz is possibly the most distinguished earthquake seismologist in all of Central and South America. Among many other credentials, Lomnitz has personally experienced the shaking and devastation that accompanied no fewer than five major earthquakes—Chile, 1939; Kern County, California, 1952; Chile, 1960; Caracas,Venezuela, 1967; and Mexico City, 1985. Thus he clearly has much to teach someone like myself, who has never even actually felt a real earthquake.What is this slim book? The Road to Total Earthquake Safety summarizes Lomnitz's May 1999 presentation at the Seventh Mallet-Milne Lecture, sponsored by the Society for Earthquake and Civil Engineering Dynamics. His arguments are motivated by the damage that occurred in three earthquakes—Mexico City, 1985; Loma Prieta, California, 1989; and Kobe, Japan, 1995. All three quakes occurred in regions where earthquakes are common. Yet in all three some of the worst damage occurred in structures located a significant distance from the epicenter and engineered specifically to resist earthquakes. Some of the damage also indicated that the structures failed because they had experienced considerable rotational or twisting motion. Clearly, Lomnitz argues, there must be fundamental flaws in the usually accepted models explaining how earthquakes generate strong motions, and how we should design resistant structures.
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.
Practical application of the KMS: 1) total system performance assessment - 16349
International Nuclear Information System (INIS)
Makino, Hitoshi; Hioki, Kazumasa; Umeki, Hiroyuki; Yang, Hongzhi; Takase, Hiroyasu; McKinley, Ian
2009-01-01
Comprehensive total system performance assessment (PA) is a key component of the safety case. Within this PA there are a number of tasks that reuse specific models and datasets, together with associated knowledge base for the disposal system considered. These are tasks where recent developments in the Knowledge Management System by Japan Atomic Energy Agency (JAEA KMS) can lead to optimisation of procedures. This paper will outline the reformulation of PA as a Knowledge Management (KM) task, discuss application of KM technologies to PA tasks, and illustrate how these can be handled electronically in a 'Performance assessment All-In-one Report System (PAIRS)' utilising hyper-links and embedded tools to minimise duplication of material, ease Quality Assurance (QA) and facilitate the regular updating required in the Japanese programme. (authors)
Gil, Gustavo; Savino, Giovanni; Piantini, Simone; Baldanzini, Niccolò; Happee, Riender; Pierini, Marco
2017-11-17
Motorcycle riders are involved in significantly more crashes per kilometer driven than passenger car drivers. Nonetheless, the development and implementation of motorcycle safety systems lags far behind that of passenger cars. This research addresses the identification of the most effective motorcycle safety solutions in the context of different countries. A knowledge-based system of motorcycle safety (KBMS) was developed to assess the potential for various safety solutions to mitigate or avoid motorcycle crashes. First, a set of 26 common crash scenarios was identified from the analysis of multiple crash databases. Second, the relative effectiveness of 10 safety solutions was assessed for the 26 crash scenarios by a panel of experts. Third, relevant information about crashes was used to weigh the importance of each crash scenario in the region studied. The KBMS method was applied with an Italian database, with a total of more than 1 million motorcycle crashes in the period 2000-2012. When applied to the Italian context, the KBMS suggested that automatic systems designed to compensate for riders' or drivers' errors of commission or omission are the potentially most effective safety solution. The KBMS method showed an effective way to compare the potential of various safety solutions, through a scored list with the expected effectiveness of each safety solution for the region to which the crash data belong. A comparison of our results with a previous study that attempted a systematic prioritization of safety systems for motorcycles (PISa project) showed an encouraging agreement. Current results revealed that automatic systems have the greatest potential to improve motorcycle safety. Accumulating and encoding expertise in crash analysis from a range of disciplines into a scalable and reusable analytical tool, as proposed with the use of KBMS, has the potential to guide research and development of effective safety systems. As the expert assessment of the crash
System theory and safety models in Swedish, UK, Dutch and Australian road safety strategies.
Hughes, B P; Anund, A; Falkmer, T
2015-01-01
Road safety strategies represent interventions on a complex social technical system level. An understanding of a theoretical basis and description is required for strategies to be structured and developed. Road safety strategies are described as systems, but have not been related to the theory, principles and basis by which systems have been developed and analysed. Recently, road safety strategies, which have been employed for many years in different countries, have moved to a 'vision zero', or 'safe system' style. The aim of this study was to analyse the successful Swedish, United Kingdom and Dutch road safety strategies against the older, and newer, Australian road safety strategies, with respect to their foundations in system theory and safety models. Analysis of the strategies against these foundations could indicate potential improvements. The content of four modern cases of road safety strategy was compared against each other, reviewed against scientific systems theory and reviewed against types of safety model. The strategies contained substantial similarities, but were different in terms of fundamental constructs and principles, with limited theoretical basis. The results indicate that the modern strategies do not include essential aspects of systems theory that describe relationships and interdependencies between key components. The description of these strategies as systems is therefore not well founded and deserves further development. Copyright © 2014 Elsevier Ltd. All rights reserved.
Study on 'Safety qualification of process computers used in safety systems of nuclear power plants'
International Nuclear Information System (INIS)
Bertsche, K.; Hoermann, E.
1991-01-01
The study aims at developing safety standards for hardware and software of computer systems which are increasingly used also for important safety systems in nuclear power plants. The survey of the present state-of-the-art of safety requirements and specifications for safety-relevant systems and, additionally, for process computer systems has been compiled from national and foreign rules. In the Federal Republic of Germany the KTA safety guides and the BMI/BMU safety criteria have to be observed. For the design of future computer-aided systems in nuclear power plants it will be necessary to apply the guidelines in [DIN-880] and [DKE-714] together with [DIN-192]. With the aid of a risk graph the various functions of a system, or of a subsystem, can be evaluated with regard to their significance for safety engineering. (orig./HP) [de
Design an optimum safety policy for personnel safety management - A system dynamic approach
International Nuclear Information System (INIS)
Balaji, P.
2014-01-01
Personnel safety management (PSM) ensures that employee's work conditions are healthy and safe by various proactive and reactive approaches. Nowadays it is a complex phenomenon because of increasing dynamic nature of organisations which results in an increase of accidents. An important part of accident prevention is to understand the existing system properly and make safety strategies for that system. System dynamics modelling appears to be an appropriate methodology to explore and make strategy for PSM. Many system dynamics models of industrial systems have been built entirely for specific host firms. This thesis illustrates an alternative approach. The generic system dynamics model of Personnel safety management was developed and tested in a host firm. The model was undergone various structural, behavioural and policy tests. The utility and effectiveness of model was further explored through modelling a safety scenario. In order to create effective safety policy under resource constraint, DOE (Design of experiment) was used. DOE uses classic designs, namely, fractional factorials and central composite designs. It used to make second order regression equation which serve as an objective function. That function was optimized under budget constraint and optimum value used for safety policy which shown greatest improvement in overall PSM. The outcome of this research indicates that personnel safety management model has the capability for acting as instruction tool to improve understanding of safety management and also as an aid to policy making
Design an optimum safety policy for personnel safety management - A system dynamic approach
Energy Technology Data Exchange (ETDEWEB)
Balaji, P. [The Glocal University, Mirzapur Pole, Delhi- Yamuntori Highway, Saharanpur 2470001 (India)
2014-10-06
Personnel safety management (PSM) ensures that employee's work conditions are healthy and safe by various proactive and reactive approaches. Nowadays it is a complex phenomenon because of increasing dynamic nature of organisations which results in an increase of accidents. An important part of accident prevention is to understand the existing system properly and make safety strategies for that system. System dynamics modelling appears to be an appropriate methodology to explore and make strategy for PSM. Many system dynamics models of industrial systems have been built entirely for specific host firms. This thesis illustrates an alternative approach. The generic system dynamics model of Personnel safety management was developed and tested in a host firm. The model was undergone various structural, behavioural and policy tests. The utility and effectiveness of model was further explored through modelling a safety scenario. In order to create effective safety policy under resource constraint, DOE (Design of experiment) was used. DOE uses classic designs, namely, fractional factorials and central composite designs. It used to make second order regression equation which serve as an objective function. That function was optimized under budget constraint and optimum value used for safety policy which shown greatest improvement in overall PSM. The outcome of this research indicates that personnel safety management model has the capability for acting as instruction tool to improve understanding of safety management and also as an aid to policy making.
Design an optimum safety policy for personnel safety management - A system dynamic approach
Balaji, P.
2014-10-01
Personnel safety management (PSM) ensures that employee's work conditions are healthy and safe by various proactive and reactive approaches. Nowadays it is a complex phenomenon because of increasing dynamic nature of organisations which results in an increase of accidents. An important part of accident prevention is to understand the existing system properly and make safety strategies for that system. System dynamics modelling appears to be an appropriate methodology to explore and make strategy for PSM. Many system dynamics models of industrial systems have been built entirely for specific host firms. This thesis illustrates an alternative approach. The generic system dynamics model of Personnel safety management was developed and tested in a host firm. The model was undergone various structural, behavioural and policy tests. The utility and effectiveness of model was further explored through modelling a safety scenario. In order to create effective safety policy under resource constraint, DOE (Design of experiment) was used. DOE uses classic designs, namely, fractional factorials and central composite designs. It used to make second order regression equation which serve as an objective function. That function was optimized under budget constraint and optimum value used for safety policy which shown greatest improvement in overall PSM. The outcome of this research indicates that personnel safety management model has the capability for acting as instruction tool to improve understanding of safety management and also as an aid to policy making.
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.
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
System safety education focused on system management
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.
Safety Management System in Croatia Control Ltd.
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...
Energy Technology Data Exchange (ETDEWEB)
NONE
1983-03-01
In this research on a hydrogen energy total system, studies are conducted on the plan of a hydrogen energy proving pilot base and on hydrogen as fuel for automobiles. It is estimated that the construction of a hydrogen energy proving pilot base will cost 7.125-billion yen in total. The sum includes 6.410-billion yen for the construction of a system on an island named Island A, 500-million yen for structures on an island named Island B, and 215-million yen for the construction of a marine transportation system between the two islands. Large shares will go to a hydroelectric power plant and a hydrogen liquefaction system, the two occupying approximately half of the total sum. In the study of hydrogen as fuel for automobiles, it is concluded that hydrogen is advantageously employed as fuel for automobiles. When comparison is made in terms of heat value, it is found that even a hydrogen engine which is a mere modification of a currently used engine is comparable to the currently used engine in terms of performance. As for abnormal combustion, a hydrogen/air 2-system injection method is contrived, and this solves the problem almost completely. Cryogenic hydrogen is advantageous in both NOx emission and heat efficiency though within certain limitations. From the viewpoint of safety, the recommended automobile fuel structural formula is GH{sub 2}(MH). (NEDO)
Primary break with total loss of high pressure safety injection
International Nuclear Information System (INIS)
Cordelle, F.; Champ, M.; Pochard, R.
1988-10-01
The probabilitic safety assessment of a 900 MW plant has displayed the potential importance, with regard to the risk, of intermediate primary breaks with failure of the high pressure safety injection system. The probability of such sequence is about 10 -6 /plant X year. Therefore, it is necessary to establish: - if this sequence can lead to core melt down, - if clad ruptures can occur. This event must be taken into account to determine the repair time of contaminated systems. For these studies, a three inch equivalent diameter break is considerd, as this is the most sensitive in its category with regard to these phenomena. In addition to the above objectives, the purpose of these studies is to evaluate the sensitivity of the results to the following parameters: - the time limit at which the operator starts cooling down the plant via the steam generators. Two calculations have been made with the RELAP code (1 and 2) and two with the CATHARE code (3 and 4) - the pump trip time. Four calculations have been made with the CATHARE code (5, 6, 7 and 8). In the case of failure of only one high pressure safety injection file, 6 calculations have been made with the CATHARE code, concerning the influence of pump trip time (9, 10, 11, 12, 13 and 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
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.
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
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)
Safety climate and culture: Integrating psychological and systems perspectives.
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).
2005-01-01
This document concerns the award of a contract for the supply, installation, commissioning and maintenance of the LHC access safety system. The Finance Committee is invited to agree to the negotiation of a contract with the consortium CEGELEC CENTRE EST (FR) â?" CEGELEC (NL) â?" SEMER (FR), the lowest bidder, for the supply, installation, commissioning and maintenance for five years of the LHC access safety system for a total amount of 891 423 euros (1 367 633 Swiss francs), subject to revision for inflation for the maintenance of the system from 1 January 2007, with an option for the supply of safety PLCs for a total amount of 894 951 euros (1 373 046 Swiss francs), not subject to revision, bringing the total amount to 1 786 374 euros (2 740 679 CHF).The rate of exchange used is that stipulated in the tender.
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
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
Quantitative safety assessment of air traffic control systems through system control capacity
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
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)
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)
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
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
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
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....
International Nuclear Information System (INIS)
Almen, A.; Lundh, C.
2015-01-01
Quality assurance has been identified as an important part of radiation protection and safety for a considerable time period. A rational expansion and improvement of quality assurance is to integrate radiation protection and safety in a management system. The aim of this study was to explore factors influencing the implementing strategy when introducing a management system including radiation protection and safety in hospitals and to outline benefits of such a system. The main experience from developing a management system is that it is possible to create a vast number of common policies and routines for the whole hospital, resulting in a cost-efficient system. One of the key benefits is the involvement of management at all levels, including the hospital director. Furthermore, a transparent system will involve staff throughout the organisation as well. A management system supports a common view on what should be done, who should do it and how the activities are reviewed. An integrated management system for radiation protection and safety includes key elements supporting a safety culture. (authors)
Regulatory Oversight of Safety Culture in Finland: A Systemic Approach to Safety
International Nuclear Information System (INIS)
Oedewald, P.; Väisäsvaara, J.
2016-01-01
In Finland the Radiation and Nuclear Safety Authority STUK specifies detailed regulatory requirements for good safety culture. Both the requirements and the practical safety culture oversight activities reflect a systemic approach to safety: the interconnections between the technical, human and organizational factors receive special attention. The conference paper aims to show how the oversight of safety culture can be integrated into everyday oversight activities. The paper also emphasises that the scope of the safety culture oversight is not specific safety culture activities of the licencees, but rather the overall functioning of the licence holder or the new build project organization from safety point of view. The regulatory approach towards human and organizational factors and safety culture has evolved throughout the years of nuclear energy production in Finland. Especially the recent new build projects have highlighted the need to systematically pay attention to the non-technical aspects of safety as it has become obvious how the HOF issues can affect the design processes and quality of construction work. Current regulatory guides include a set of safety culture related requirements. The requirements are binding to the licence holders and they set both generic and specific demands on the licencee to understand, monitor and to develop safety culture of their own organization but also that of their supplier network. The requirements set for the licence holders has facilitated the need to develop the regulator’s safety culture oversight practices towards a proactive and systemic approach.
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)
Safety aspects of preoperative high-dose glucocorticoid in primary total knee replacement
DEFF Research Database (Denmark)
Jørgensen, C C; Pitter, F T; Kehlet, H
2017-01-01
Background: Preoperative single high-dose glucocorticoid may have early outcome benefits in total hip arthroplasty (THA) and knee arthroplasty (TKA), but long-term safety aspects have not been evaluated. Methods: From October 2013, the departments reporting to the prospective Lundbeck Foundation....... Conclusions: In this detailed prospective cohort study, preoperative high-dose glucocorticoid administration was not associated with LOS >4 days, readmissions or infectious complications in TKA patients without contraindications....
The ATLAS Detector Safety System
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...
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
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)
Analyzing Software Requirements Errors in Safety-Critical, Embedded Systems
Lutz, Robyn R.
1993-01-01
This paper analyzes the root causes of safety-related software errors in safety-critical, embedded systems. The results show that software errors identified as potentially hazardous to the system tend to be produced by different error mechanisms than non- safety-related software errors. Safety-related software errors are shown to arise most commonly from (1) discrepancies between the documented requirements specifications and the requirements needed for correct functioning of the system and (2) misunderstandings of the software's interface with the rest of the system. The paper uses these results to identify methods by which requirements errors can be prevented. The goal is to reduce safety-related software errors and to enhance the safety of complex, embedded systems.
Using system dynamics simulation for assessment of hydropower system safety
King, L. M.; Simonovic, S. P.; Hartford, D. N. D.
2017-08-01
Hydropower infrastructure systems are complex, high consequence structures which must be operated safely to avoid catastrophic impacts to human life, the environment, and the economy. Dam safety practitioners must have an in-depth understanding of how these systems function under various operating conditions in order to ensure the appropriate measures are taken to reduce system vulnerability. Simulation of system operating conditions allows modelers to investigate system performance from the beginning of an undesirable event to full system recovery. System dynamics simulation facilitates the modeling of dynamic interactions among complex arrangements of system components, providing outputs of system performance that can be used to quantify safety. This paper presents the framework for a modeling approach that can be used to simulate a range of potential operating conditions for a hydropower infrastructure system. Details of the generic hydropower infrastructure system simulation model are provided. A case study is used to evaluate system outcomes in response to a particular earthquake scenario, with two system safety performance measures shown. Results indicate that the simulation model is able to estimate potential measures of system safety which relate to flow conveyance and flow retention. A comparison of operational and upgrade strategies is shown to demonstrate the utility of the model for comparing various operational response strategies, capital upgrade alternatives, and maintenance regimes. Results show that seismic upgrades to the spillway gates provide the largest improvement in system performance for the system and scenario of interest.
Withrow, Colleen A.; Reveley, Mary S.
2015-01-01
The Aviation Safety Program (AvSP) System-Wide Safety and Assurance Technologies (SSAT) Project asked the AvSP Systems and Portfolio Analysis Team to identify SSAT-related trends. SSAT had four technical challenges: advance safety assurance to enable deployment of NextGen systems; automated discovery of precursors to aviation safety incidents; increasing safety of human-automation interaction by incorporating human performance, and prognostic algorithm design for safety assurance. This report reviews incident data from the NASA Aviation Safety Reporting System (ASRS) for system-component-failure- or-malfunction- (SCFM-) related and human-factor-related incidents for commercial or cargo air carriers (Part 121), commuter airlines (Part 135), and general aviation (Part 91). The data was analyzed by Federal Aviation Regulations (FAR) part, phase of flight, SCFM category, human factor category, and a variety of anomalies and results. There were 38 894 SCFM-related incidents and 83 478 human-factorrelated incidents analyzed between January 1993 and April 2011.
Soft systems methodology as a systemic approach to nuclear safety management
International Nuclear Information System (INIS)
Vieira Neto, Antonio S.; Guilhen, Sabine N.; Rubin, Gerson A.; Caldeira Filho, Jose S.; Camargo, Iara M.C.
2017-01-01
Safety approach currently adopted by nuclear installations is built almost exclusively upon analytical methodologies based, mainly, on the belief that the properties of a system, such as its safety, are given by its constituent parts. This approach, however, does not properly address the complex dynamic interactions between technical, human and organizational factors occurring within and outside the organization. After the accident at Fukushima Daiichi nuclear power plant in March 2011, experts of the International Atomic Energy Agency (IAEA) recommended a systemic approach as a complementary perspective to nuclear safety. The aim of this paper is to present an overview of the systems thinking approach and its potential use for structuring socio technical problems involved in the safety of nuclear installations, highlighting the methodologies related to the soft systems thinking, in particular the Soft Systems Methodology (SSM). The implementation of a systemic approach may thus result in a more holistic picture of the system by the complex dynamic interactions between technical, human and organizational factors. (author)
Soft systems methodology as a systemic approach to nuclear safety management
Energy Technology Data Exchange (ETDEWEB)
Vieira Neto, Antonio S.; Guilhen, Sabine N.; Rubin, Gerson A.; Caldeira Filho, Jose S.; Camargo, Iara M.C., E-mail: asvneto@ipen.br, E-mail: snguilhen@ipen.br, E-mail: garubin@ipen.br, E-mail: jscaldeira@ipen.br, E-mail: icamargo@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNE-SP), Sao Paulo, SP (Brazil)
2017-07-01
Safety approach currently adopted by nuclear installations is built almost exclusively upon analytical methodologies based, mainly, on the belief that the properties of a system, such as its safety, are given by its constituent parts. This approach, however, does not properly address the complex dynamic interactions between technical, human and organizational factors occurring within and outside the organization. After the accident at Fukushima Daiichi nuclear power plant in March 2011, experts of the International Atomic Energy Agency (IAEA) recommended a systemic approach as a complementary perspective to nuclear safety. The aim of this paper is to present an overview of the systems thinking approach and its potential use for structuring socio technical problems involved in the safety of nuclear installations, highlighting the methodologies related to the soft systems thinking, in particular the Soft Systems Methodology (SSM). The implementation of a systemic approach may thus result in a more holistic picture of the system by the complex dynamic interactions between technical, human and organizational factors. (author)
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
Safety analysis and evaluation methodology for fusion systems
International Nuclear Information System (INIS)
Fujii-e, Y.; Kozawa, Y.; Namba, C.
1987-03-01
Fusion systems which are under development as future energy systems have reached a stage that the break even is expected to be realized in the near future. It is desirable to demonstrate that fusion systems are well acceptable to the societal environment. There are three crucial viewpoints to measure the acceptability, that is, technological feasibility, economy and safety. These three points have close interrelation. The safety problem is more important since three large scale tokamaks, JET, TFTR and JT-60, start experiment, and tritium will be introduced into some of them as the fusion fuel. It is desirable to establish a methodology to resolve the safety-related issues in harmony with the technological evolution. The promising fusion system toward reactors is not yet settled. This study has the objective to develop and adequate methodology which promotes the safety design of general fusion systems and to present a basis for proposing the R and D themes and establishing the data base. A framework of the methodology, the understanding and modeling of fusion systems, the principle of ensuring safety, the safety analysis based on the function and the application of the methodology are discussed. As the result of this study, the methodology for the safety analysis and evaluation of fusion systems was developed. New idea and approach were presented in the course of the methodology development. (Kako, I.)
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?
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
Model-based safety architecture framework for complex systems
Schuitemaker, Katja; Rajabali Nejad, Mohammadreza; Braakhuis, J.G.; Podofillini, Luca; Sudret, Bruno; Stojadinovic, Bozidar; Zio, Enrico; Kröger, Wolfgang
2015-01-01
The shift to transparency and rising need of the general public for safety, together with the increasing complexity and interdisciplinarity of modern safety-critical Systems of Systems (SoS) have resulted in a Model-Based Safety Architecture Framework (MBSAF) for capturing and sharing architectural
Using wide area differential GPS to improve total system error for precision flight operations
Alter, Keith Warren
Total System Error (TSE) refers to an aircraft's total deviation from the desired flight path. TSE can be divided into Navigational System Error (NSE), the error attributable to the aircraft's navigation system, and Flight Technical Error (FTE), the error attributable to pilot or autopilot control. Improvement in either NSE or FTE reduces TSE and leads to the capability to fly more precise flight trajectories. The Federal Aviation Administration's Wide Area Augmentation System (WAAS) became operational for non-safety critical applications in 2000 and will become operational for safety critical applications in 2002. This navigation service will provide precise 3-D positioning (demonstrated to better than 5 meters horizontal and vertical accuracy) for civil aircraft in the United States. Perhaps more importantly, this navigation system, which provides continuous operation across large regions, enables new flight instrumentation concepts which allow pilots to fly aircraft significantly more precisely, both for straight and curved flight paths. This research investigates the capabilities of some of these new concepts, including the Highway-In-The Sky (HITS) display, which not only improves FTE but also reduces pilot workload when compared to conventional flight instrumentation. Augmentation to the HITS display, including perspective terrain and terrain alerting, improves pilot situational awareness. Flight test results from demonstrations in Juneau, AK, and Lake Tahoe, CA, provide evidence of the overall feasibility of integrated, low-cost flight navigation systems based on these concepts. These systems, requiring no more computational power than current-generation low-end desktop computers, have immediate applicability to general aviation flight from Cessnas to business jets and can support safer and ultimately more economical flight operations. Commercial airlines may also, over time, benefit from these new technologies.
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
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.)
Energy Technology Data Exchange (ETDEWEB)
NONE
1999-03-01
Under the hydrogen-utilizing international clean energy system technology project WE-NET (World Energy NET Work) in fiscal 1998, researches and studies were conducted to clearly define safety designs and to improve on accident-and-safety analyses. In relation with system safety design, investigations continued into Japanese and foreign manuals and regulations about the handling of hydrogen and its peripherals, and safe design guidelines (draft) were compiled. Anomalies and accidents supposed to be typical of each of the systems concerned were investigated. As for accident-and-safety analyses, incorporation of a turbulence model was studied in relation to models representing the leak, evaporation, and diffusion of liquid hydrogen, and improvement was achieved when the scope of evaluation was enlarged concerning the hydrogen detonation model. The integration of the two models was discussed for the due evaluation of a series of processes of liquid hydrogen leak, evaporation, diffusion, and detonation. Calculation was performed for two assumed accidents, and the results were found to justify the integration of the two models. (NEDO)
Identifying the most significant indicators of the total road safety performance index.
Tešić, Milan; Hermans, Elke; Lipovac, Krsto; Pešić, Dalibor
2018-04-01
The review of the national and international literature dealing with the assessment of the road safety level has shown great efforts of the authors who tried to define the methodology for calculating the composite road safety index on a territory (region, state, etc.). The procedure for obtaining a road safety composite index of an area has been largely harmonized. The question that has not been fully resolved yet concerns the selection of indicators. There is a wide range of road safety indicators used to show a road safety situation on a territory. Road safety performance index (RSPI) obtained on the basis of a larger number of safety performance indicators (SPIs) enable decision makers to more precisely define the earlier goal- oriented actions. However, recording a broader comprehensive set of SPIs helps identify the strengths and weaknesses of a country's road safety system. Providing high quality national and international databases that would include comparable SPIs seems to be difficult since a larger number of countries dispose of a small number of identical indicators available for use. Therefore, there is a need for calculating a road safety performance index with a limited number of indicators (RSPI ln n ) which will provide a comparison of a sufficient quality, of as many countries as possible. The application of the Data Envelopment Analysis (DEA) method and correlative analysis has helped to check if the RSPI ln n is likely to be of sufficient quality. A strong correlation between the RSPI ln n and the RSPI has been identified using the proposed methodology. Based on this, the most contributing indicators and methodologies for gradual monitoring of SPIs, have been defined for each country analyzed. The indicator monitoring phases in the analyzed countries have been defined in the following way: Phase 1- the indicators relating to alcohol, speed and protective systems; Phase 2- the indicators relating to roads and Phase 3- the indicators relating to
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)
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.)
System Safety in an IT Service Organization
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.
Aviation Safety Reporting System: Process and Procedures
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.
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.
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
Nitrogen-system safety study: Portsmouth Gaseous Diffusion Plant
International Nuclear Information System (INIS)
1982-07-01
The Department of Energy has primary responsibility for the safety of operations at DOE-owned nuclear facilities. The guidelines for the analysis of credible accidents are outlined in DOE Order 5481.1. DOE has requested that existing plant facilities and operations be reviewed for potential safety problems not covered by standard industrial safety procedures. This review is being conducted by investigating individual facilities and documenting the results in Safety Study Reports which will be compiled to form the Existing Plant Final Safety Analysis Report which is scheduled for completion in September, 1984. This Safety Study documents the review of the Plant Nitrogen System facilities and operations and consists of Section 4.0, Facility and Process Description, and Section 5.0, Accident Analysis, of the Final Safety Analysis Report format. The existing nitrogen system consists of a Superior Air Products Company Type D Nitrogen Plant, nitrogen storage facilities, vaporization facilities and a distribution system. The system is designed to generate and distribute nitrogen gas used in the cascade for seal feed, buffer systems, and for servicing equipment when exceptionally low dew points are required. Gaseous nitrogen is also distributed to various process auxiliary buildings. The average usage is approximately 130,000 standard cubic feet per day
Integrated therapy safety management system.
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.
Integrated therapy safety management system
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
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...
Benefits of a systematic approach to maintenance for safety and safety related systems
International Nuclear Information System (INIS)
Dam, R.F.; Ayazzudin, S.; Nickerson, J.H.
2003-01-01
For safety and safety-related systems, nuclear plants have to balance the requirements of demonstrating the reliability of each system, while maintaining the system and plant availability. With the goal of demonstrating statistical reliability, these systems have extensive testing programs, which often results in system unavailability and this can impact the plant capacity. The inputs to the process are often safety and regulatory related, resulting in programs that provide a high level of scrutiny. In such cases, the value of the application of a Systematic Assessment of Maintenance (SAM) process, such as Reliability Centered Maintenance (RCM), is questioned. The special case of Standby-Safety systems was discussed in a previous paper, where it was demonstrated how SAM techniques provide useful insight into current system performance, the impact of testing on component and system reliability, and how PSA considerations can be integrated into a comprehensive Maintenance, Surveillance, and Inspection (MSI) strategy. Although the system reliability requirements are an important part of the strategy evaluation, SAM techniques provide a systematic assessment within a broader context. Testing is only one part of an overall strategy focused on ensuring that component function is maintained through a combination of monitoring technologies (including testing), predictive techniques, and intrusive maintenance strategies. Each strategy is targeted to known component degradation mechanisms. This thinking can be extended to safety and safety related systems in general. Over the past 6 years, AECL has been working with CANDU utilities in the development and implementation of a comprehensive and integrated Plant Life Management (PLiM) program. As part of developing a comprehensive plant asset management approach, SAM techniques are used to develop a technical basis that not only works towards ensuring reliable operation of plant systems, but also facilitates the optimization and
Declarative Rule-based Safety for Robotic Perception Systems
DEFF Research Database (Denmark)
Mogensen, Johann Thor Ingibergsson; Kraft, Dirk; Schultz, Ulrik Pagh
2017-01-01
Mobile robots are used across many domains from personal care to agriculture. Working in dynamic open-ended environments puts high constraints on the robot perception system, which is critical for the safety of the system as a whole. To achieve the required safety levels the perception system needs...... to be certified, but no specific standards exist for computer vision systems, and the concept of safe vision systems remains largely unexplored. In this paper we present a novel domain-specific language that allows the programmer to express image quality detection rules for enforcing safety constraints...
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
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.
Operation safety of complex industrial systems. Main concepts
International Nuclear Information System (INIS)
Zwingelstein, G.
2009-01-01
Operation safety consists in knowing, evaluating, foreseeing, measuring and mastering the technological system and human failures in order to avoid their impacts on health and people's safety, on productivity, and on the environment, and to preserve the Earth's resources. This article recalls the main concepts of operation safety: 1 - evolutions in the domain; 2 - failures, missions and functions of a system and of its components: functional failure, missions and functions, industrial processes, notions of probability; 3 - basic concepts and operation safety: reliability, unreliability, failure density, failure rate, relations between them, availability, maintainability, safety. (J.S.)
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
2010-07-01
... Production, Storage, and Offloading System Safety Zone. 147.847 Section 147.847 Navigation and Navigable... ZONES § 147.847 Safety Zone; BW PIONEER Floating Production, Storage, and Offloading System Safety Zone. (a) Description. The BW PIONEER, a Floating Production, Storage and Offloading (FPSO) system, is in...
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
Evaluating Safety Culture Under the Socio-Technical Complex Systems Perspective
International Nuclear Information System (INIS)
Lemos, F. L. de
2016-01-01
Since the term “safety culture” was coined, it has gained more and more attention as an effort to achieve higher levels of system safety. A good deal of effort has been done in order to better define, evaluate and implement safety culture programs in organizations throughout all industries, and especially in the Nuclear Industry. Unfortunately, despite all those efforts, we continue to witness accidents that are, in great part, attributed to flaws in the safety culture of the organization. Fukushima nuclear accident is one example of a serious accident in which flaws in the safety culture has been pointed to as one of the main contributors. In general, the definitions of safety culture emphasise the social aspect of the system. While the definitions also include the relations with the technical aspects, it does so in a general sense. For example, the International Nuclear Safety Advisory Group (INSAG) defines safety culture as: “The assembly of characteristics and attitudes in organizations and individuals which establishes that, as an overriding priority, nuclear plant safety issues receives the attention warranted by their significance.” By the way safety culture is defined we can infer that it represents a property of a social system, or a property of the social aspect of the system. In this sense, the social system is a component of the whole system. Where, “system” is understood to be comprised of a social (humans) and technical (equipment) aspects, as a Nuclear Power Plant, for example. Therefore, treating safety culture as an identity on its own right, finding and fixing flaws in the safety culture may not be enough to improve safety of the system. We also needed to evaluate all the interactions between the components that comprise all the aspects of the system. In some cases a flaw in the safety culture can easily be detected, such as an employee not wearing appropriate individual protection equipment, e.g., dosimeter, or when basic safety
Intelligent monitoring-based safety system of massage robot
Institute of Scientific and Technical Information of China (English)
胡宁; 李长胜; 王利峰; 胡磊; 徐晓军; 邹雲鹏; 胡玥; 沈晨
2016-01-01
As an important attribute of robots, safety is involved in each link of the full life cycle of robots, including the design, manufacturing, operation and maintenance. The present study on robot safety is a systematic project. Traditionally, robot safety is defined as follows: robots should not collide with humans, or robots should not harm humans when they collide. Based on this definition of robot safety, researchers have proposed ex ante and ex post safety standards and safety strategies and used the risk index and risk level as the evaluation indexes for safety methods. A massage robot realizes its massage therapy function through applying a rhythmic force on the massage object. Therefore, the traditional definition of safety, safety strategies, and safety realization methods cannot satisfy the function and safety requirements of massage robots. Based on the descriptions of the environment of massage robots and the tasks of massage robots, the present study analyzes the safety requirements of massage robots; analyzes the potential safety dangers of massage robots using the fault tree tool; proposes an error monitoring-based intelligent safety system for massage robots through monitoring and evaluating potential safety danger states, as well as decision making based on potential safety danger states; and verifies the feasibility of the intelligent safety system through an experiment.
Development and implementation of setpoint tolerances for special safety systems
International Nuclear Information System (INIS)
Oliva, A.F.; Balog, G.; Parkinson, D.G.; Archinoff, G.H.
1991-01-01
The establishment of tolerances and impairment limits for special safety system setpoints is part of the process whereby the plant operator demonstrates to the regulatory authority that the plant operates safely and within the defined plant licensing envelope. The licensing envelope represents the set of limits and plant operating state and for which acceptably safe plant operation has been demonstrated by the safety analysis. By definition, operation beyond this envelope contributes to overall safety system unavailability. Definition of the licensing envelope is provided in a wide range of documents including the plant operating licence, the safety report, and the plant operating policies and principles documents. As part of the safety analysis, limits are derived for each special safety system initiating parameter such that the relevant safety design objectives are achieved for all design basis events. If initiation on a given parameter occurs at a level beyond its limit, there is a potential reduction in safety system effectiveness relative to the performance credited in the plant safety analysis. These safety system parameter limits, when corrected for random and systematic instrument errors and other errors inherent in the process of periodic testing or calibration, are then used to derive parameter impairment levels and setpoint tolerances. This paper describes the methodology that has evolved at Ontario Hydro for developing and implementing tolerances for special safety system parameters (i.e., the shutdown systems, emergency coolant injection system and containment system). Tolerances for special safety system initiation setpoints are addressed specifically, although many of the considerations discussed here will apply to performance limits for other safety system components. The first part of the paper deals with the approach that has been adopted for defining and establishing setpoint limits and tolerances. The remainder of the paper addresses operational
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
Identifying behaviour patterns of construction safety using system archetypes.
Guo, Brian H W; Yiu, Tak Wing; González, Vicente A
2015-07-01
Construction safety management involves complex issues (e.g., different trades, multi-organizational project structure, constantly changing work environment, and transient workforce). Systems thinking is widely considered as an effective approach to understanding and managing the complexity. This paper aims to better understand dynamic complexity of construction safety management by exploring archetypes of construction safety. To achieve this, this paper adopted the ground theory method (GTM) and 22 interviews were conducted with participants in various positions (government safety inspector, client, health and safety manager, safety consultant, safety auditor, and safety researcher). Eight archetypes were emerged from the collected data: (1) safety regulations, (2) incentive programs, (3) procurement and safety, (4) safety management in small businesses (5) production and safety, (6) workers' conflicting goals, (7) blame on workers, and (8) reactive and proactive learning. These archetypes capture the interactions between a wide range of factors within various hierarchical levels and subsystems. As a free-standing tool, they advance the understanding of dynamic complexity of construction safety management and provide systemic insights into dealing with the complexity. They also can facilitate system dynamics modelling of construction safety process. Copyright © 2015 Elsevier Ltd. All rights reserved.
International Nuclear Information System (INIS)
Bertrand, R.
1990-01-01
This memorandum shall provide the background for the work of the European Community Commission which is to analyze safety principles relating to redundancy. The redundancy-related basic safety principles applied in French nuclear power plants are the following: . the single-failure criterion, . provisions additional to application of the single-failure criterion. These are mainly provisions made at the design stage to minimize risks associated with common cause failures or the risks of human error which can lead to such failures: - protection against hazards of internal and external origin, - the geographical or physical separation of equipment, - the independence of electrical power supplies and distribution systems, - the additional resources and associated operating procedures making it possible to accommodate total loss of the safety systems. The scope also includes the operating rules which ensure availability of redundant safety-related equipment. The provisions relating to the single-failure criterion are detailed in Basic Safety Rule 1.3.A appended. The application of these principles proposed by the operating organization and accepted by the safety authorities for the design and operation of the standby core cooling system (System RIS) is explained
Classification of Aeronautics System Health and Safety Documents
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...
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
Priority ranking of safety-related systems for structural assessment at Savannah River Site
International Nuclear Information System (INIS)
Kao, G.C.; Daugherty, W.L.; Barnes, D.M.
1993-01-01
In order to extend the service life of safety related structures and systems in a logical manner, a Structural Enhancement Program was initiated to evaluate the structural integrity of eight systems, namely: cooling water system, emergency cooling system, moderator recovery system, supplementary safety system, water removal system, service raw water system, service clarified water system, and river water system. Since the level of importance of each system to reactor operations varies from one system to another, the scope of structural integrity evaluation for each system should be prioritized accordingly. This paper presents the assessment of system priority for structural evaluation based on a ranking methodology and specifies the level of structural evaluation consistent with the established priority. The effort was undertaken by a five-member panel representing four major disciplines, including: structures, reactor engineering/operations, risk management, and materials. The above systems were divided into a total of thirty-five subsystems. These subsystems were then ranked with six attributes, namely: safety classification, degradation mechanisms, difficulty of replacement, failure mode, radiation dose to workers, and consequence of failure. Each attribute was assigned a set of consequences or events with corresponding weighting scores. The results of the ranking process yielded two groups of subsystems, categorized as Priority I and II subsystems. The level of structural assessment was then formulated accordingly. The prioritized approach will allow more efficient allocation of resources, so that the Structural Enhancement Program can be implemented in a cost-effective and efficient manner
Challenges in the management of gas voids in safety related systems
International Nuclear Information System (INIS)
Ezekoye, L.I.; Turkowski, W.M.; Ferraraccio, F.P.; Swartz, M.M.
2009-01-01
Gas intrusion into Safety Related Systems, such as the Emergency Core Cooling System (ECCS), Decay Heat Removal (DHR) and Containment Spray (CS) in nuclear power plants is undesirable and can lead to pump binding (depending on the void fraction and flow rate) and damaging water hammer events. Gas ingestion in pumps can result in total or momentary loss of hydraulic performance resulting in possible pump shaft seizure rendering the pumps unable to perform their safety functions or reduce the pump discharge pressure and flow capacity to the point that the system cannot perform its design function. Extreme cases of gas water hammer can result in physical damage to system piping, components and supports, and possible relief valve lifting events with consequential loss of inventory. NRC Generic Letter GL 2008 01, 'Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal, and Containment Spray Systems,' requires US utilities to demonstrate that suitable design, operational and testing measures are in place to maintain licensing commitments. The Generic Letter (GL 2008 01) outlines a number of actions that are detailed in nature, such as establishing pump void tolerance limits; establishing limits on pump suction void fractions, assuring adequate system venting capability, identification of all possible sources of gas intrusion, preventing vortex formation in tanks, and determining acceptable limits of gas in system discharge piping.. Regarding one of these issues, GL 2008 01 indicates that the amount of gas that can be ingested without significant impact on pump design, gas dispersion and flow rate. Each US nuclear power plant licensee is required to evaluate their ECCS, DHR and CS system design, operation and test procedures to assure that gas intrusion is minimized and monitored in order to maintain system operability and compliance with the requirements of 10 CFR 50 Appendix B. Typically, gas pockets get into the safety related systems through a number
Challenges in the management of gas voids in safety related systems
Energy Technology Data Exchange (ETDEWEB)
Ezekoye, L.I.; Turkowski, W.M.; Ferraraccio, F.P.; Swartz, M.M. [Westinghouse Electric Company LLC, Pittsburgh (United States)
2009-04-15
Gas intrusion into Safety Related Systems, such as the Emergency Core Cooling System (ECCS), Decay Heat Removal (DHR) and Containment Spray (CS) in nuclear power plants is undesirable and can lead to pump binding (depending on the void fraction and flow rate) and damaging water hammer events. Gas ingestion in pumps can result in total or momentary loss of hydraulic performance resulting in possible pump shaft seizure rendering the pumps unable to perform their safety functions or reduce the pump discharge pressure and flow capacity to the point that the system cannot perform its design function. Extreme cases of gas water hammer can result in physical damage to system piping, components and supports, and possible relief valve lifting events with consequential loss of inventory. NRC Generic Letter GL 2008 01, 'Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal, and Containment Spray Systems,' requires US utilities to demonstrate that suitable design, operational and testing measures are in place to maintain licensing commitments. The Generic Letter (GL 2008 01) outlines a number of actions that are detailed in nature, such as establishing pump void tolerance limits; establishing limits on pump suction void fractions, assuring adequate system venting capability, identification of all possible sources of gas intrusion, preventing vortex formation in tanks, and determining acceptable limits of gas in system discharge piping.. Regarding one of these issues, GL 2008 01 indicates that the amount of gas that can be ingested without significant impact on pump design, gas dispersion and flow rate. Each US nuclear power plant licensee is required to evaluate their ECCS, DHR and CS system design, operation and test procedures to assure that gas intrusion is minimized and monitored in order to maintain system operability and compliance with the requirements of 10 CFR 50 Appendix B. Typically, gas pockets get into the safety related systems through
Tranexamic Acid in Total Joint Arthroplasty: Efficacy and Safety
Directory of Open Access Journals (Sweden)
Mohammad Reza Rasouli
2015-01-01
relatively rare, the majority of studies that evaluated the efficacy of TA are underpowered for evaluation of its safety. In a recent study by Poeran et al. using a large national database including 872,416 patients who had total hip or knee arthroplasty, the authors suggested that TA is effective in reducing the need for blood transfusions while not increasing the risk of VTE and renal complications (8. However, it is still advised that patients with cardiac stents and previous thromboembolic events including ischemic stroke not be administered TA. TA can also cause gastrointestinal disturbance and its dose needs to be adjusted in patients with renal impairment (5. In conclusion, administration of TA in TJA patient is a cost-effective blood conservation strategy and there is strong evidence to support the efficacy and safety of the drug in reducing blood loss and transfusion in TJA patients. Given the adverse effects of allogeneic blood transfusion on the outcome of TJA, administration of TA should be considered in patients with no contraindication for its use.
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
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.
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
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)
Microbiological performance of a food safety management system in a food service operation.
Lahou, E; Jacxsens, L; Daelman, J; Van Landeghem, F; Uyttendaele, M
2012-04-01
The microbiological performance of a food safety management system in a food service operation was measured using a microbiological assessment scheme as a vertical sampling plan throughout the production process, from raw materials to final product. The assessment scheme can give insight into the microbiological contamination and the variability of a production process and pinpoint bottlenecks in the food safety management system. Three production processes were evaluated: a high-risk sandwich production process (involving raw meat preparation), a medium-risk hot meal production process (starting from undercooked raw materials), and a low-risk hot meal production process (reheating in a bag). Microbial quality parameters, hygiene indicators, and relevant pathogens (Listeria monocytogenes, Salmonella, Bacillus cereus, and Escherichia coli O157) were in accordance with legal criteria and/or microbiological guidelines, suggesting that the food safety management system was effective. High levels of total aerobic bacteria (>3.9 log CFU/50 cm(2)) were noted occasionally on gloves of food handlers and on food contact surfaces, especially in high contamination areas (e.g., during handling of raw material, preparation room). Core control activities such as hand hygiene of personnel and cleaning and disinfection (especially in highly contaminated areas) were considered points of attention. The present sampling plan was used to produce an overall microbiological profile (snapshot) to validate the food safety management system in place.
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)
Experimental research progress on passive safety systems of Chinese advanced PWR
International Nuclear Information System (INIS)
Xiao Zejun; Zhuo Wenbin; Zheng Hua; Chen Bingde; Zong Guifang; Jia Dounan
2003-01-01
TMI and Chernobyl accidents, having pronounced impact on nuclear industries, triggered the governments as well as interested institutions to devote much attention to the safety of nuclear power plant and public's requirements on nuclear power plant safety were also going to be stricter and stricter. It is obvious that safety level of an ordinary light water reactor is no longer satisfactory to these requirements. Recently, the safety authorities have recommended the implementation of passive system to improve the safety of nuclear reactors. Passive safety system is one of the main differences between Chinese advanced PWR and other conventional PWR. The working principle of passive safety system is to utilize the gravity, natural convection (natural circulation) and stored energy to implement the system's safety function. Reactors with passive safety systems are not only safer, but also more economical. The passive safety system of Chinese advanced PWR is composed of three independent systems, i.e. passive containment cooling system, passive residual heat removal system and passive core makeup tank injection system. This paper is a summary of experimental research progress on passive containment cooling system, passive residual heat removal system and passive core makeup tank injection system
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
Research on advanced system safety assessment procedures (4)
International Nuclear Information System (INIS)
Suzuki, Kazuhiko; Shimada, Yukiyasu
2001-03-01
The past research reports in the area of safety engineering proposed the Computer-aided HAZOP system to be applied to Nuclear Reprocessing Facilities. Automated HAZOP system has great advantage compared with human analysts in terms of accuracy of the results, and time required to conduct HAZOP studies. This report surveys the literature on risk assessment and safety design based on the concept of independent protection layers (IPLs). Furthermore, to improve HAZOP System, tool is proposed to construct the basic model and the internal state model. Such HAZOP system is applied to analyze two kinds of processes, where the ability of the proposed system is verified. In addition, risk assessment support system is proposed to integrate safety design environment and assessment result to be used by other plants as well as to enable the underline plant to use other plants' information. This technique can be implemented using web-based safety information systems. (author)
ABWR (K-6/7) construction experience (computer-based safety system)
International Nuclear Information System (INIS)
Yokomura, T.
1998-01-01
TEPCO applied a digital safety system to Kashiwazaki-Kariwa Nuclear Power Station Unit Nos. 6 and 7, the world's first ABWR plant. Although this was the first time to apply a digital safety logic system in Japan, we were able to complete construction of K-6/7 very successfully and without any delay. TEPCO took a approach of developing a substantial amount of experience in digital non- safety systems before undertaking the design of the safety protection system. This paper describes the history, techniques and experience behind achieving a highly reliable digital safety system. (author)
SACS2: Dynamic and Formal Safety Analysis Method for Complex Safety Critical System
International Nuclear Information System (INIS)
Koh, Kwang Yong; Seong, Poong Hyun
2009-01-01
Fault tree analysis (FTA) is one of the most widely used safety analysis technique in the development of safety critical systems. However, over the years, several drawbacks of the conventional FTA have become apparent. One major drawback is that conventional FTA uses only static gates and hence can not capture dynamic behaviors of the complex system precisely. Although several attempts such as dynamic fault tree (DFT), PANDORA, formal fault tree (FFT) and so on, have been made to overcome this problem, they can not still do absolute or actual time modeling because they adapt relative time concept and can capture only sequential behaviors of the system. Second drawback of conventional FTA is its lack of rigorous semantics. Because it is informal in nature, safety analysis results heavily depend on an analyst's ability and are error-prone. Finally reasoning process which is to check whether basic events really cause top events is done manually and hence very labor-intensive and timeconsuming for the complex systems. In this paper, we propose a new safety analysis method for complex safety critical system in qualitative manner. We introduce several temporal gates based on timed computational tree logic (TCTL) which can represent quantitative notion of time. Then, we translate the information of the fault trees into UPPAAL query language and the reasoning process is automatically done by UPPAAL which is the model checker for time critical system
PROPOSAL OF VOIVODESHIP ROAD SAFETY IMPROVEMENT PROGRAMME
Tomasz SZCZURASZEK; Jan KEMPA
2016-01-01
The article presents a proposal of the ‘GAMBIT KUJAWSKO-POMORSKI’ Road Safety Improvement Programme. The main idea of the Programme is to establish and initiate systems that will be responsible for the most important areas of activity within road safety, including road safety control, supervision, and management systems in the whole Voivodeship. In total, the creation and start of nine such systems has been proposed, namely: the Road Safety Management, the Integrated Road Rescue Service, the ...
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.
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
Energy Technology Data Exchange (ETDEWEB)
Lee, Youho, E-mail: euo@kaist.ac.kr; Kim, Bokyung, E-mail: bkkim2@kaist.ac.kr; NO, Hee Cheon, E-mail: hcno@kaist.ac.kr
2016-10-15
Highlights: • Zircaloy embrittlement criteria can increase to 1370 °C for CP-ECR lower than 13%. • The draft ECCS criteria of U.S. NRC allow less than 5% in power margin. • The Japanese fracture-based criteria allow around 5% in power margin. • Increasing SIT inventory is effective in assuring safety margin for power uprates. - Abstract: This study investigates the engineering compatibility between emergency core cooling system criteria and safety water injection systems, in the pursuit of safety margin increase of light water reactors. This study proposes an acceptable temperature increase to 1370 °C as long as equivalent cladding reacted calculated by the Cathcart–Pawel equation is below 13%, after an extensive literature review. The influence of different ECCS criteria on the safety margin during large break loss of coolant accident is investigated for OPR-1000 by the system code MARS-KS, implemented with the KINS-REM method. The fracture-based emergency core cooling system (ECCS) criteria proposed in this study are shown to enable power margins up to 10%. In the meantime, the draft U.S. NRC’s embrittlement criteria (burnup-sensitive) and Japanese fracture-based criteria are shown to allow less than 5%, and around 5% of power margins, respectively. Increasing safety injection tank (SIT) water inventory is the key, yet convenient, way of assuring safety margin for power increase. More than 20% increase in the SIT water inventory is required to allow 15% power margins, for the U.S. NRC’s burnup-dependent embrittlement criteria. Controlling SIT water inventory would be a useful option that could allow the industrial desire to pursue power margins even under the recent atmosphere of imposing stricter ECCS criteria for the considerable burnup effects.
Integrated environment, safety, and health management system description
International Nuclear Information System (INIS)
Zoghbi, J. G.
2000-01-01
The Integrated Environment, Safety, and Health Management System Description that is presented in this document describes the approach and management systems used to address integrated safety management within the Richland Environmental Restoration Project
Safety philosophy in upgrading the EBR-II plant protection system
International Nuclear Information System (INIS)
Sackett, J.I.
1976-01-01
The EBR-II plant protection system (PPS) has been substantially modified, upgrading its performance to more fully comply with modern safety philosophy and criteria. The upgrading effort required that the total reactor system be evaluated for possible faults and that a PPS be designed to accommodate them. The result was deletion of a number of existing trip functions and upgrading of others. Particular attention was given to loss of primary pumping power and reactivity insertion events. The design and performance criteria for the PPS has been more firmly established, understanding of the PPS function has been improved and the reactor has been subjected to fewer spurious trips, improving operational reliability
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
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
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
Design of the reactor coolant system and associated systems in nuclear power plants. Safety guide
International Nuclear Information System (INIS)
2008-01-01
This Safety Guide was prepared under the IAEA programme for establishing safety standards for nuclear power plants. The basic requirements for the design of safety systems for nuclear power plants are established in the Safety Requirements publication, Safety Standards Series No. NS-R-1 on Safety of Nuclear Power Plants: Design, which it supplements. This Safety Guide describes how the requirements for the design of the reactor coolant system (RCS) and associated systems in nuclear power plants should be met. 1.2. This publication is a revision and combination of two previous Safety Guides, Safety Series No. 50-SG-D6 on Ultimate Heat Sink and Directly Associated Heat Transport Systems for Nuclear Power Plants (1981), and Safety Series No. 50-SG-D13 on Reactor Coolant and Associated Systems in Nuclear Power Plants (1986), which are superseded by this new Safety Guide. 1.3. The revision takes account of developments in the design of the RCS and associated systems in nuclear power plants since the earlier Safety Guides were published in 1981 and 1986, respectively. The other objectives of the revision are to ensure consistency with Ref., issued in 2000, and to update the technical content. In addition, an appendix on pressurized heavy water reactors (PHWRs) has been included
Preliminary investigation on reliability assessment of passive safety system
International Nuclear Information System (INIS)
Huang Changfan; Kuang Bo
2012-01-01
The reliability evaluation of passive safety system plays an important part in probabilistic safety assessment (PSA) of nuclear power plant applying passive safety design, which depends quantitatively on reliabilities of passive safety system. According to the object of reliability assessment of passive safety system, relevant parameters are identified. Then passive system behavior during accident scenarios are studied. A practical example of this method is given for the case of reliability assessment of AP1000 passive heat removal system in loss of normal feedwater accident. Key and design parameters of PRHRS are identified and functional failure criteria are established. Parameter combinations acquired by Latin hyper~ cube sampling (LHS) in possible parametric ranges are input and calculations of uncertainty propagation through RELAP5/MOD3 code are carried out. Based on the calculations, sensitivity assessment on PRHRS functional criteria and reliability evaluation of the system are presented, which might provide further PSA with PRHR system reliability. (authors)
Directory of Open Access Journals (Sweden)
Mohd Kamar I.F.
2014-01-01
Full Text Available Occupational Health and Safety Management Systems is part of the overall management system that facilitates the management of the OS&H risks associated with the business of the organization. This includes the organizational structure, planning activities, responsibilities, practices, procedures, processes and resources for developing, implementing, achieving, reviewing and maintaining the organization’s OS&H policy. The purpose of this research is to determine the level of awareness of contractors on OSH management systems. A total of 34 numbers of class A contractors in Kelantan registered with Pusat Khidmat Kontraktor (PKK were randomly selected. Data was collected using self-administered questionnaire. The findings indicate that most of the Class A Contractor in Kelantan aware that the occupational safety and health management system are important and should be practiced to achieve zero accident and death on site
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.
A new concept of safety parameter display system
International Nuclear Information System (INIS)
Martinez, A.S.; Oliveira, L.F.S. de; Schirru, R.; Thome Filho, Z.D.; Silva, R.A. da.
1986-07-01
A general description of Angra-1 Parameter Display System (SSPA), a real time and on-line computerized monitoring system for the parameters related to the power plant safety is presented. This system has the main purpose of diminish the load on the Angra-1 power plant operators at an emergency event by supplying them with the additional tools serving as the basis for a prompt identification of the accident. The SSPA is a kind of safety parameter display system whose concept was introduced after Three Mile Island accident in USA. The SSPA comprises two nuclear applications independently considered. They are included into the Parameters Monitoring Integrated System (SIMP) and the safety critical function system (SFCS). (Author) [pt
International Nuclear Information System (INIS)
Zhang Qihong; Lu Jigen; Zhang Ping; Wang Wanping; Dai Xia
2012-01-01
As the rapid development of nuclear technology, the safety supervision of nuclear and radiation becomes very important. The safety radiation frame system should be constructed, the safety super- vision ability for nuclear and radiation should be improved. How to implement effectively above mission should be a new subject of Provincial environmental protection department. Through investigating the innovation of nuclear and radiation supervision system, innovation of mechanism, innovation of capacity, innovation of informatization and so on, the provincial nuclear and radiation safety supervision model is proposed, and the safety framework of nuclear and radiation in Jiangsu is elementally established in the paper. (authors)
Concept of system safety on operating nuclear power plant
International Nuclear Information System (INIS)
Miyano, Hiroshi; Yamaguchi, Akira; Demachi, Kazuyuki; Takata, Takashi; Arai, Shigeki; Sugiyama, Naoki
2015-01-01
The total system design on Nuclear Plant ensures 'Nuclear safety' with making practically achievable efforts to prevent and mitigate nuclear and radiological accidents. The performance based system design with 'Defence in depth (D-I-D)' has been laid out as the key means in 'preventing accidents', 'controlling escalation to serious consequences', and 'preventing harmful consequences to the public'. D-I-D is extended to the management of severe accidents, and is an approach intended to provide protection against the development of a wide variety of events by means of redundant, diverse and independent protective barriers. It is crucial to maintain plant integrity with mass quantity of radioactive material present in reactor core, against potential consequences (risk) on people and the environment caused by external hazards, particularly, earthquake and tsunami. The fundamental approach on D-I-D is to address uncertainties by means of successive measures, so that if one measure fails, other, or subsequent measure will be available to ensure safety. Risk analysis should be conducted to validate and enhance reliability of the defence barriers against consequences on people and the environment. (author)
Development of the Advanced Nuclear Safety Information Management (ANSIM) System
Energy Technology Data Exchange (ETDEWEB)
Sohn, Jae Min; Ko, Young Cheol; Song, Tai Gil [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)
2012-05-15
Korea has become a technically independent nuclear country and has grown into an exporter of nuclear technologies. Thus, nuclear facilities are increasing in significance at KAERI (Korea Atomic Energy Research Institute), and it is time to address the nuclear safety. The importance of nuclear safety cannot be overemphasized. Therefore, a management system is needed urgently to manage the safety of nuclear facilities and to enhance the efficiency of nuclear information. We have established ISP (Information Strategy Planning) for the Integrated Information System of nuclear facility and safety management. The purpose of this paper is to develop a management system for nuclear safety. Therefore, we developed the Advanced Nuclear Safety Information Management system (hereinafter referred to as the 'ANSIM system'). The ANSIM system has been designed and implemented to computerize nuclear safety information for standardization, integration, and sharing in real-time. Figure 1 shows the main home page of the ANSIM system. In this paper, we describe the design requirements, contents, configurations, and utilizations of the ANSIM system
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
Total energy system in the future
International Nuclear Information System (INIS)
Hijikata, K.
1994-01-01
The possibility of improving the thermal efficiency of energy systems from an exergy point of view is discussed. In total energy systems, we should employ multi-pass recycling consisting of thermal and chemical energies. The recycling system is supported by electrical energy, which is provided by a renewable energy source or by excess commercial electric power. This total energy system should be considered not only in one country, but all around the globe. (author). 6 figs., 4 tabs., 8 refs
Development of web-based safety review advisory system
International Nuclear Information System (INIS)
Kim, M. W.; Lee, H. C.; Park, S. O.; Lee, K. H.; Hur, K. Y.; Lee, S. J.; Choi, S. S.; Kang, C. M.
2002-01-01
For the development of an expert system supporting the safety review of nuclear power plants, the application was implemented after gathering necessary theoretical background and practical requirements. The general and the detail functional specifications were established, and they are investigated by KINS (Korea Institute of Nuclear Safety). The Safety Review Advisory System(SRAS), this application on web-server environment was developed according to the above specifications. Reviews can do their safety reviewing regardless of their speciality or reviewing experiences because SRAS is operated by the safety review plans which are converted to standardized format. When the safety reviewing is carried out by using SRAS, the results of safety reviewing are accumulated in the database and may be utilized later usefully, and we can grasp safety reviewing progress. Users of SRAS are categorized into four groups, administrator, project manager, project reviewer and general reviewer. Each user group is delegated appropriate access capability. The function and some screen shots of SRAS are described
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.)
International Nuclear Information System (INIS)
DAVIS, S.J.
2000-01-01
This document identifies critical characteristics of components to be dedicated for use in Safety Significant (SS) Systems, Structures, or Components (SSCs). This document identifies the requirements for the components of the common, radiation area, monitor alarm in the WESF pool cell. These are procured as Commercial Grade Items (CGI), with the qualification testing and formal dedication to be performed at the Waste Encapsulation Storage Facility (WESF) for use in safety significant systems. System modifications are to be performed in accordance with the approved design. Components for this change are commercially available and interchangeable with the existing alarm configuration This document focuses on the operational requirements for alarm, declaration of the safety classification, identification of critical characteristics, and interpretation of requirements for procurement. Critical characteristics are identified herein and must be verified, followed by formal dedication, prior to the components being used in safety related applications
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
LOCA analysis of SCWR-M with passive safety system
Energy Technology Data Exchange (ETDEWEB)
Liu, X.J., E-mail: xiaojingliu@sjtu.edu.cn [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, 800 Dong Chuan Road, Shanghai 200240 (China); Fu, S.W. [Navy University of Engineering, Wuhan, Hubei (China); Xu, Z.H. [Shanghai Nuclear Engineering Research and Design Institute, Shanghai (China); Yang, Y.H. [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, 800 Dong Chuan Road, Shanghai 200240 (China); Cheng, X. [Institute of Fusion and Nuclear Technology, Karlsruhe Institute of Technology (KIT), Kaiserstr. 12, 76131 Karlsruhe (Germany)
2013-06-15
Highlights: • Application of the ATHLET-SC code to the trans-critical analysis for SCWR. • Development of a passive safety system for SCWR-M. • Analysis of hot/cold leg LOCA behaviour with different break size. • Introduction of some mitigation measures for SCWR-M -- Abstract: A new SCWR conceptual design (mixed spectrum supercritical water cooled reactor: SCWR-M) is proposed by Shanghai Jiao Tong University (SJTU). R and D activities covering core design, safety system design and code development of SCWR-M are launched at SJTU. Safety system design and analysis is one of the key tasks during the development of SCWR-M. Considering the current advanced reactor design, a new passive safety system for SCWR-M including isolation cooling system (ICS), accumulator injection system (ACC), gravity driven cooling system (GDCS) and automatic depressurization system (ADS) is proposed. Based on the modified and preliminarily assessed system code ATHLET-SC, loss of coolant accident (LOCA) analysis for hot and cold leg is performed in this paper. Three different break sizes are analyzed to clarify the hot and cold LOCA characteristics of the SCWR-M. The influence of the break location and break size on the safety performance of SCWR-M is also concluded. Several measures to induce the core coolant flow and to mitigate core heating up are also discussed. The results achieved so far demonstrate the feasibility of the proposed passive safety system to keep the SCWR-M core at safety condition during loss of coolant accident.
Passive safety systems for integral reactors
International Nuclear Information System (INIS)
Kuul, V.S.; Samoilov, O.B.
1996-01-01
In this paper, a wide range of passive safety systems intended for use on integral reactors is considered. The operation of these systems relies on natural processes and does not require external power supplies. Using these systems, there is the possibility of preventing serious consequences for all classes of accidents including reactivity, loss-of-coolant and loss of heat sink as well as severe accidents. Enhancement of safety system reliability has been achieved through the use of self-actuating devices, capable of providing passive initiation of protective and isolation systems, which respond immediately to variations in the physical parameters of the fluid in the reactor or in a guard vessel. For beyond design base accidents accompanied by complete loss of heat removal capability, autonomous self-actuated ERHR trains have been proposed. These trains are completely independent of the secondary loops and need no action to isolate them from the steam turbine plant. Passive safety principles have been consistently implemented in AST-500, ATETS-200 and VPBER 600 which are new generation NPPs developed by OKBM. Their main characteristic is enhanced stability over a wide range of internal and external emergency initiators. (author). 10 figs
Passive safety systems for integral reactors
Energy Technology Data Exchange (ETDEWEB)
Kuul, V S; Samoilov, O B [OKB Mechanical Engineering (Russian Federation)
1996-12-01
In this paper, a wide range of passive safety systems intended for use on integral reactors is considered. The operation of these systems relies on natural processes and does not require external power supplies. Using these systems, there is the possibility of preventing serious consequences for all classes of accidents including reactivity, loss-of-coolant and loss of heat sink as well as severe accidents. Enhancement of safety system reliability has been achieved through the use of self-actuating devices, capable of providing passive initiation of protective and isolation systems, which respond immediately to variations in the physical parameters of the fluid in the reactor or in a guard vessel. For beyond design base accidents accompanied by complete loss of heat removal capability, autonomous self-actuated ERHR trains have been proposed. These trains are completely independent of the secondary loops and need no action to isolate them from the steam turbine plant. Passive safety principles have been consistently implemented in AST-500, ATETS-200 and VPBER 600 which are new generation NPPs developed by OKBM. Their main characteristic is enhanced stability over a wide range of internal and external emergency initiators. (author). 10 figs.
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
Development of Network Protocol for the Integrated Safety System
Energy Technology Data Exchange (ETDEWEB)
Park, S. W.; Baek, J. I.; Lee, S. H.; Park, C. S.; Park, K. H.; Shin, J. M. [Hannam Univ., Daejeon (Korea, Republic of)
2007-06-15
Communication devices in the safety system of nuclear power plants are distinguished from those developed for commercial purposes in terms of a strict requirement of safety. The concept of safety covers the determinability, the reliability, and the separation/isolation to prevent the undesirable interactions among devices. The safety also requires that these properties be never proof less. Most of the current commercialized communication products rarely have the safety properties. Moreover, they can be neither verified nor validated to satisfy the safety property of implementation process. This research proposes the novel architecture and protocol of a data communication network for the safety system in nuclear power plants.
Development of Network Protocol for the Integrated Safety System
International Nuclear Information System (INIS)
Park, S. W.; Baek, J. I.; Lee, S. H.; Park, C. S.; Park, K. H.; Shin, J. M.
2007-06-01
Communication devices in the safety system of nuclear power plants are distinguished from those developed for commercial purposes in terms of a strict requirement of safety. The concept of safety covers the determinability, the reliability, and the separation/isolation to prevent the undesirable interactions among devices. The safety also requires that these properties be never proof less. Most of the current commercialized communication products rarely have the safety properties. Moreover, they can be neither verified nor validated to satisfy the safety property of implementation process. This research proposes the novel architecture and protocol of a data communication network for the safety system in nuclear power plants
Radiological safety system based on real-time tritium-in-air monitoring indoors and in effluents
International Nuclear Information System (INIS)
Bidica, N.; Sofalca, N; Balteanu, O.; Srefan, I.
2006-01-01
Exposure to tritium is an important health hazard in any tritium processing facility so that implementing a real-time tritium monitoring system is necessary for its operation in safety conditions. The tritium processing facility operators need to be informed at any time about the in-air tritium concentration indoors or in the stack effluents, in order to detect immediately any leaks in tritium containments, or any releases inside the buildings or to the environment. This information is very important for adopting if necessary protection measures and correcting actions as quickly as possible. In this paper we describe an improved real-time tritium monitoring system designed for the Heavy Water Detritiation Pilot Plant of National Institute for Cryogenics and Isotopes Separation, Rm. Valcea, Romania. The design of the Radiological Safety System implemented for the ICIT Water Detritiation Pilot Plant is intended to provide the maximum safety level based on the ALARA concept. The main functions of tritium monitoring system are: - monitoring the working areas and gaseous effluents by determination of the tritium-in-air activity concentration; - local and remote data display; - assessing of environment dose equivalent rates and dose equivalents in the working environment (for personnel exposure control and work planning); - assessing the total tritium activity released to the environment through ventilation exhaust stack; - safety functions, i.e., local and remote, locking/unlocking personnel access, process shut-down in emergency conditions and start of the air cleaning systems. With all these features our tritium monitoring system is really a safety system adequate for personnel and environmental protection. (authors)
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)
Survey of systems safety analysis methods and their application to nuclear waste management systems
International Nuclear Information System (INIS)
Pelto, P.J.; Winegardner, W.K.; Gallucci, R.H.V.
1981-11-01
This report reviews system safety analysis methods and examines their application to nuclear waste management systems. The safety analysis methods examined include expert opinion, maximum credible accident approach, design basis accidents approach, hazard indices, preliminary hazards analysis, failure modes and effects analysis, fault trees, event trees, cause-consequence diagrams, G0 methodology, Markov modeling, and a general category of consequence analysis models. Previous and ongoing studies on the safety of waste management systems are discussed along with their limitations and potential improvements. The major safety methods and waste management safety related studies are surveyed. This survey provides information on what safety methods are available, what waste management safety areas have been analyzed, and what are potential areas for future study
Survey of systems safety analysis methods and their application to nuclear waste management systems
Energy Technology Data Exchange (ETDEWEB)
Pelto, P.J.; Winegardner, W.K.; Gallucci, R.H.V.
1981-11-01
This report reviews system safety analysis methods and examines their application to nuclear waste management systems. The safety analysis methods examined include expert opinion, maximum credible accident approach, design basis accidents approach, hazard indices, preliminary hazards analysis, failure modes and effects analysis, fault trees, event trees, cause-consequence diagrams, G0 methodology, Markov modeling, and a general category of consequence analysis models. Previous and ongoing studies on the safety of waste management systems are discussed along with their limitations and potential improvements. The major safety methods and waste management safety related studies are surveyed. This survey provides information on what safety methods are available, what waste management safety areas have been analyzed, and what are potential areas for future study.
Overview of Risk Mitigation for Safety-Critical Computer-Based Systems
Torres-Pomales, Wilfredo
2015-01-01
This report presents a high-level overview of a general strategy to mitigate the risks from threats to safety-critical computer-based systems. In this context, a safety threat is a process or phenomenon that can cause operational safety hazards in the form of computational system failures. This report is intended to provide insight into the safety-risk mitigation problem and the characteristics of potential solutions. The limitations of the general risk mitigation strategy are discussed and some options to overcome these limitations are provided. This work is part of an ongoing effort to enable well-founded assurance of safety-related properties of complex safety-critical computer-based aircraft systems by developing an effective capability to model and reason about the safety implications of system requirements and design.
Simplified safety and containment systems for the iris reactor
International Nuclear Information System (INIS)
Conway, L.E.; Lombardi, C.; Ricotti, M.; Oriani, L.
2001-01-01
The IRIS (International Reactor Innovative and Secure) is a 100 - 300 MW modular type pressurized water reactor supported by the U.S. DOE NERI Program. IRIS features a long-life core to provide proliferation resistance and to reduce the volume of spent fuel, as well as reduce maintenance requirements. IRIS utilizes an integral reactor vessel that contains all major primary system components. This integral reactor vessel makes it possible to reduce containment size; making the IRIS more cost competitive. IRIS is being designed to enhance reactor safety, and therefore a key aspect of the IRIS program is the development of the safety and containment systems. These systems are being designed to maximize containment integrity, prevent core uncover following postulated accidents, minimize the probability and consequences of severe accidents, and provide a significant simplification over current safety system designs. The design of the IRIS containment and safety systems has been identified and preliminary analyses have been completed. The IRIS safety concept employs some unique features that minimize the consequences of postulated design basis events. This paper will provide a description of the containment design and safety systems, and will summarize the analysis results. (author)
Autonomous system for launch vehicle range safety
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. .
System code improvements for modelling passive safety systems and their validation
Energy Technology Data Exchange (ETDEWEB)
Buchholz, Sebastian; Cron, Daniel von der; Schaffrath, Andreas [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Garching (Germany)
2016-11-15
GRS has been developing the system code ATHLET over many years. Because ATHLET, among other codes, is widely used in nuclear licensing and supervisory procedures, it has to represent the current state of science and technology. New reactor concepts such as Generation III+ and IV reactors and SMR are using passive safety systems intensively. The simulation of passive safety systems with the GRS system code ATHLET is still a big challenge, because of non-defined operation points and self-setting operation conditions. Additionally, the driving forces of passive safety systems are smaller and uncertainties of parameters have a larger impact than for active systems. This paper addresses the code validation and qualification work of ATHLET on the example of slightly inclined horizontal heat exchangers, which are e. g. used as emergency condensers (e. g. in the KERENA and the CAREM) or as heat exchanger in the passive auxiliary feed water systems (PAFS) of the APR+.
Aviation Safety Hotline Information System -
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...
Estimating total 239240Pu in blow-sand mounds of two safety-shot sites
International Nuclear Information System (INIS)
Gilbert, R.O.; Essington, E.H.
1977-01-01
A study for estimating the total amount (inventory) of 239 240 Pu in blow-sand mounds at two safety-shot sites (Area 13-Project 57 on the Nellis Air Force Base and Clean Slate 3 on the Tonopah Test Range in Nevada) is described. The total amount in blow-sand mounds at these two sites is estimated to be 5.8 +- 1.3 (total +- standard error) and 10.6 +- 2.5 curies, respectively. The total 239 240 Pu in mounds plus desert pavement areas, both to a depth of 5 cm below desert pavement level, is estimated to be 39 +- 5.7 curies at the Project 57 site and 36 +- 4.8 curies at Clean Slate 3. These estimates are compared with the somewhat higher estimates of 46 +- 9 and 37 +- 5.4 curies reported that pertain to only the top 5 cm of mounds and desert pavement. The possibility is discussed that these differences are due to sampling variability arising from the skewed nature of plutonium concentrations, particularly near ground zero
21 CFR 862.1710 - Total triiodothyronine test system.
2010-04-01
... 21 Food and Drugs 8 2010-04-01 2010-04-01 false Total triiodothyronine test system. 862.1710... Systems § 862.1710 Total triiodothyronine test system. (a) Identification. A total triiodothyronine test system is a device intended to measure the hormone triiodothyronine in serum and plasma. Measurements...
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
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
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)
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
Development of web-based safety review advisory system
International Nuclear Information System (INIS)
Kim, M. W.; Hur, K. Y.; Lee, S. J.; Choi, S. J.
2002-01-01
For the development of an expert system supporting the safety review of nuclear power plants, the application was implemented after gathering necessary theoretical background and practical requirements. The general and the detail functional specifications were established, and they are investigated by KINS. Safety Review Advisory System (SRAS), this application on web-server environment was developed according to the above specifications. Reviews can do their safety reviewing regardless of their speciality or reviewing experiences because SRAS is operated by the safety review plans which are converted to standardized format. When the safety reviewing is carried out by using SRAS, the results of safety reviewing are accumulated in the database and may be utilized later usefully, and we can grasp safety reviewing progress. Users of SRAS are categorized into four groups, administrator, project manager, project reviewer and general reviewer. Each user group is delegated appropriate access capability. The function and some screen shots of SRAS are described
Towards predictive cardiovascular safety : a systems pharmacology approach
Snelder, Nelleke
2014-01-01
Cardiovascular safety issues related to changes in blood pressure, arise frequently in drug development. In the thesis “Towards predictive cardiovascular safety – a systems pharmacology approach”, a system-specific model is described to quantify drug effects on the interrelationship between mean
Safety program considerations for space nuclear reactor systems
International Nuclear Information System (INIS)
Cropp, L.O.
1984-08-01
This report discusses the necessity for in-depth safety program planning for space nuclear reactor systems. The objectives of the safety program and a proposed task structure is presented for meeting those objectives. A proposed working relationship between the design and independent safety groups is suggested. Examples of safety-related design philosophies are given
Qualification of safety-critical software for digital reactor safety system in nuclear power plants
International Nuclear Information System (INIS)
Kwon, Kee-Choon; Park, Gee-Yong; Kim, Jang-Yeol; Lee, Jang-Soo
2013-01-01
This paper describes the software qualification activities for the safety-critical software of the digital reactor safety system in nuclear power plants. The main activities of the software qualification processes are the preparation of software planning documentations, verification and validation (V and V) of the software requirements specifications (SRS), software design specifications (SDS) and codes, and the testing of the integrated software and integrated system. Moreover, the software safety analysis and software configuration management are involved in the software qualification processes. The V and V procedure for SRS and SDS contains a technical evaluation, licensing suitability evaluation, inspection and traceability analysis, formal verification, software safety analysis, and an evaluation of the software configuration management. The V and V processes for the code are a traceability analysis, source code inspection, test case and test procedure generation. Testing is the major V and V activity of the software integration and system integration phases. The software safety analysis employs a hazard operability method and software fault tree analysis. The software configuration management in each software life cycle is performed by the use of a nuclear software configuration management tool. Through these activities, we can achieve the functionality, performance, reliability, and safety that are the major V and V objectives of the safety-critical software in nuclear power plants. (author)
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....
Modular reliability modeling of the TJNAF personnel safety system
International Nuclear Information System (INIS)
Cinnamon, J.; Mahoney, K.
1997-01-01
A reliability model for the Thomas Jefferson National Accelerator Facility (formerly CEBAF) personnel safety system has been developed. The model, which was implemented using an Excel spreadsheet, allows simulation of all or parts of the system. Modularity os the model's implementation allows rapid open-quotes what if open-quotes case studies to simulate change in safety system parameters such as redundancy, diversity, and failure rates. Particular emphasis is given to the prediction of failure modes which would result in the failure of both of the redundant safety interlock systems. In addition to the calculation of the predicted reliability of the safety system, the model also calculates availability of the same system. Such calculations allow the user to make tradeoff studies between reliability and availability, and to target resources to improving those parts of the system which would most benefit from redesign or upgrade. The model includes calculated, manufacturer's data, and Jefferson Lab field data. This paper describes the model, methods used, and comparison of calculated to actual data for the Jefferson Lab personnel safety system. Examples are given to illustrate the model's utility and ease of use
Innovation in the Safety of nuclear systems: fundamental aspects
International Nuclear Information System (INIS)
Herranz, L. E.
2009-01-01
Safety commercial nuclear reactors has been an indispensable condition for future enlargement of power generation based on nuclear technology. Its fundamental principle, defence in depth, far from being outdated, is still adopted as a key foundation in the advanced nuclear system (generations III and IV). Nevertheless, the cumulative experience gained in the operation and maintenance of nuclear reactors, the development of methodologies like the probabilistic safety analysis, the use of passive safety systems and, even, the inherent characteristics of some new design (which exclude accident scenarios), allow estimating safety figures of merit even more outstanding that those achieved in the second generation of nuclear reactors. This safety innovation of upcoming nuclear reactors has entailed a huge investigation program (generation III) that will be focused on optimizing and demonstrating the postulated safety of future nuclear systems (Generation IV). (Author)
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
Energy Technology Data Exchange (ETDEWEB)
Moreno Chamorro, P.; Gallego Diaz, C.
2011-07-01
The main objective of this work is to show the current status of the implementation of integrated analysis of safety (IAS) methodology and its SCAIS associated tool (system of simulation codes for IAS) to the sequence analysis of total loss of feedwater in a PWR reactor model Westinghouse of three loops with large, dry containment.
Energy Technology Data Exchange (ETDEWEB)
NONE
1997-03-01
Safety measures and assessment technology were studied for the WE-NET project. As the study result in fiscal 1996, the information on safety design, anomaly and accident was collected and arranged. The information on safety measures, ideology and criterion was also collected by visiting some domestic and overseas organizations experienced about handling of liquid hydrogen (LH). The initial survey was made for the safety design ideology, analytical technique and disaster preventive measures of LNG systems as the similar cold liquid system. Accidents and explosion accident of a hydrogen production plant (water electrolysis) in Germany were analyzed. Events on storage tanks and leakage around the tanks were studied as typical risk of LH considering temporary and LNG system design information. The model based on the LH spillage test result and 3-D dispersion of vapor cloud were prepared by modifying a simulation code. The model allowed evaluation of the effect of explosion and fire accidents of compressed hydrogen gas and flying fragments on structures and people, and visual display of distances from a tank and damage conditions. 19 refs., 29 figs., 18 tabs.
49 CFR 659.19 - System safety program plan: contents.
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...
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.
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
Heget, Jeffrey R; Bagian, James P; Lee, Caryl Z; Gosbee, John W
2002-12-01
In 1998 the Veterans Health Administration (VHA) created the National Center for Patient Safety (NCPS) to lead the effort to reduce adverse events and close calls systemwide. NCPS's aim is to foster a culture of safety in the Department of Veterans Affairs (VA) by developing and providing patient safety programs and delivering standardized tools, methods, and initiatives to the 163 VA facilities. To create a system-oriented approach to patient safety, NCPS looked for models in fields such as aviation, nuclear power, human factors, and safety engineering. Core concepts included a non-punitive approach to patient safety activities that emphasizes systems-based learning, the active seeking out of close calls, which are viewed as opportunities for learning and investigation, and the use of interdisciplinary teams to investigate close calls and adverse events through a root cause analysis (RCA) process. Participation by VA facilities and networks was voluntary. NCPS has always aimed to develop a program that would be applicable both within the VA and beyond. NCPS's full patient safety program was tested and implemented throughout the VA system from November 1999 to August 2000. Program components included an RCA system for use by caregivers at the front line, a system for the aggregate review of RCA results, information systems software, alerts and advisories, and cognitive acids. Following program implementation, NCPS saw a 900-fold increase in reporting of close calls of high-priority events, reflecting the level of commitment to the program by VHA leaders and staff.
Brack, Antoni; Gieskes, J.F.B.
1990-01-01
The purpose of this paper is to show that Product Liability (PL) and Product Safety (PS) are relevant categories in terms of the Total Quality Management literature. From a societal point of view PL and PS are two sides of a coin. The effects of the PL and PS regulation on companies are described
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
Safety applications of computer based systems for the process industry
International Nuclear Information System (INIS)
Bologna, Sandro; Picciolo, Giovanni; Taylor, Robert
1997-11-01
Computer based systems, generally referred to as Programmable Electronic Systems (PESs) are being increasingly used in the process industry, also to perform safety functions. The process industry as they intend in this document includes, but is not limited to, chemicals, oil and gas production, oil refining and power generation. Starting in the early 1970's the wide application possibilities and the related development problems of such systems were recognized. Since then, many guidelines and standards have been developed to direct and regulate the application of computers to perform safety functions (EWICS-TC7, IEC, ISA). Lessons learnt in the last twenty years can be summarised as follows: safety is a cultural issue; safety is a management issue; safety is an engineering issue. In particular, safety systems can only be properly addressed in the overall system context. No single method can be considered sufficient to achieve the safety features required in many safety applications. Good safety engineering approach has to address not only hardware and software problems in isolation but also their interfaces and man-machine interface problems. Finally, the economic and industrial aspects of the safety applications and development of PESs in process plants are evidenced throughout all the Report. Scope of the Report is to contribute to the development of an adequate awareness of these problems and to illustrate technical solutions applied or being developed
The Intelligent Safety System: could it introduce complex computing into CANDU shutdown systems
International Nuclear Information System (INIS)
Hall, J.A.; Hinds, H.W.; Pensom, C.F.; Barker, C.J.; Jobse, A.H.
1984-07-01
The Intelligent Safety System is a computerized shutdown system being developed at the Chalk River Nuclear Laboratories (CRNL) for future CANDU nuclear reactors. It differs from current CANDU shutdown systems in both the algorithm used and the size and complexity of computers required to implement the concept. This paper provides an overview of the project, with emphasis on the computing aspects. Early in the project several needs leading to an introduction of computing complexity were identified, and a computing system that met these needs was conceived. The current work at CRNL centers on building a laboratory demonstration of the Intelligent Safety System, and evaluating the reliability and testability of the concept. Some fundamental problems must still be addressed for the Intelligent Safety System to be acceptable to a CANDU owner and to the regulatory authorities. These are also discussed along with a description of how the Intelligent Safety System might solve these problems
Safety Characteristics in System Application Software for Human Rated Exploration
Mango, E. J.
2016-01-01
NASA and its industry and international partners are embarking on a bold and inspiring development effort to design and build an exploration class space system. The space system is made up of the Orion system, the Space Launch System (SLS) and the Ground Systems Development and Operations (GSDO) system. All are highly coupled together and dependent on each other for the combined safety of the space system. A key area of system safety focus needs to be in the ground and flight application software system (GFAS). In the development, certification and operations of GFAS, there are a series of safety characteristics that define the approach to ensure mission success. This paper will explore and examine the safety characteristics of the GFAS development.
Tuning permissiveness of active safety monitors for autonomous systems
Masson , Lola; Guiochet , Jérémie; Waeselynck , Hélène; Cabrera , Kalou; Cassel , Sofia; Törngren , Martin
2018-01-01
International audience; Robots and autonomous systems have become a part of our everyday life, therefore guaranteeing their safety is crucial.Among the possible ways to do so, monitoring is widely used, but few methods exist to systematically generate safety rules to implement such monitors. Particularly, building safety monitors that do not constrain excessively the system's ability to perform its tasks is necessary as those systems operate with few human interventions.We propose in this pap...
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)
Energy Technology Data Exchange (ETDEWEB)
NONE
1984-03-01
The automatic sewing system technique research association has been commissioned by the Agency of Industrial Science and Technology for (research and development of automatic sewing systems). This program covers R and D of the elementary techniques for total systems and sewing preparation/processing, sewing/assembling, cloth handling, and system management/control. This report describes the results of the R and D efforts for the total systems. The basic data are collected for a total of 5 representative types of apparels. The technical problems involved in the current and automatic sewing systems are clarified for each apparel type, to collect the data for studying the test plant plans. The automatic sewing system for each apparel type is illustrated by images, for expressing the system concepts. The concept for the basic designs of the total system is drawn for FMS. For achieving the objective of reducing time for producing unit quantity of diversified types of clothes in small quantities at least by 50% from the current level, the study is focused on how the works are (flown) in the system, for which configuration of each device is studied. The basic specifications are drawn, viewed from development of the total system. (NEDO)
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
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)
Safety systems and features of boiling and pressurized water reactors
International Nuclear Information System (INIS)
Khair, H. O. M.
2012-06-01
The safe operation of nuclear power plants (NPP) requires a deep understanding of the functioning of physical processes and systems involved. This study was carried out to present an overview of the features of safety systems of boiling and pressurized water reactors that are available commercially. Brief description of purposes and functions of the various safety systems that are employed in these reactors was discussed and a brief comparison between the safety systems of BWRs and PWRs was made in an effort to emphasize of safety in NPPs.(Author)
International Nuclear Information System (INIS)
Zhang Yuangfang
1992-02-01
The method and results for safety parameter selection and its verification in safety parameter display system of nuclear power plants are introduced. According to safety analysis, the overall safety is divided into six critical safety functions, and a certain amount of safety parameters which can represent the integrity degree of each function and the causes of change are strictly selected. The verification of safety parameter selection is carried out from the view of applying the plant emergency procedures and in the accident man oeuvres on a full scale nuclear power plant simulator
The regulatory system of nuclear safety in Russia
International Nuclear Information System (INIS)
Mizoguchi, Shuhei
2013-01-01
This article explains what type of mechanism the nuclear system has and how nuclear safety is regulated in Russia. There are two main organizations in this system : ROSATOM and ROSTEKHADZOR. ROSATOM, which was founded in 2007, incorporates all the nuclear industries in Russia, including civil nuclear companies as well as nuclear weapons complex facilities. ROSTEKHNADZOR is the federal body that secures and supervises the safety in using atomic energy. This article also reviews three laws on regulating nuclear safety. (author)
Directory of Open Access Journals (Sweden)
I. S. Shumilov
2017-01-01
Full Text Available The paper deals with design requirements for an aviation fuel system (AFS, AFS basic design requirements, reliability, and design precautions to avoid AFS failure. Compares the reliability and fail-safety of AFS and aircraft hydraulic system (AHS, considers the promising alternative ways to raise reliability of fuel systems, as well as elaborates recommendations to improve reliability of the pipeline system components and pipeline systems, in general, based on the selection of design solutions.It is extremely advisable to design the AFS and AHS in accordance with Aviation Regulations АП25 and Accident Prevention Guidelines, ICAO (International Civil Aviation Association, which will reduce risk of emergency situations, and in some cases even avoid heavy disasters.ATS and AHS designs should be based on the uniform principles to ensure the highest reliability and safety. However, currently, this principle is not enough kept, and AFS looses in reliability and fail-safety as compared with AHS. When there are the examined failures (single and their combinations the guidelines to ensure the AFS efficiency should be the same as those of norm-adopted in the Regulations АП25 for AHS. This will significantly increase reliability and fail-safety of the fuel systems and aircraft flights, in general, despite a slight increase in AFS mass.The proposed improvements through the use of components redundancy of the fuel system will greatly raise reliability of the fuel system of a passenger aircraft, which will, without serious consequences for the flight, withstand up to 2 failures, its reliability and fail-safety design will be similar to those of the AHS, however, above improvement measures will lead to a slightly increasing total mass of the fuel system.It is advisable to set a second pump on the engine in parallel with the first one. It will run in case the first one fails for some reasons. The second pump, like the first pump, can be driven from the
Vibration analysis of the Golfech 2 safety injection system
International Nuclear Information System (INIS)
Morilhat, P.
1993-01-01
The main function of the safety injection system in a PWR plant is to ensure cooling of fuel elements in the event of a loss of coolant accident. The multistage centrifugal pump mounted-on this system induces pressure fluctuations, resulting in dynamic loads on piping. In certain plant units, these loads have caused cracking in the nozzles connected to the safety injection system, whereas in others, no damage has been observed. In order to understand the differences in dynamic behavior observed from one site to another, tests were performed on a real safety injection system, that of Golfech-2. They enabled determination of the modal characteristics of the system and identification of the hydro-acoustic source of the low head safety injection pump. They also enabled assessment of the pressure fluctuation levels in the pump suction and discharge areas as well as the vibratory response of the system when operating under partial and nominal flow conditions. Finally, these test results were used to estimate fatigue damage in the safety injection system. The experimental results will later be used to validate the model of the system undertaken with the piping design code CIRCUS and define the boundary conditions to be taken into account. (author). 6 figs., 2 refs
International Nuclear Information System (INIS)
Kim, Ma Woong; Shin, Hyeong Ki; Lee, Sang Kyu; Kim, Hyun Koon; Yoo, Kun Joong; Ryu, Yong Ho; Son, Han Seong; Song, Deok Yong
2007-01-01
As increase of operating nuclear power plants, an accident monitoring system is essential to ensure the operational safety of nuclear power plant. Thus, KINS has developed the Computerized Advisory System for a Radiological Emergency (CARE) system to monitor the operating status of nuclear power plant continuously. However, during the accidents or/and incidents some parameters could not be provided from the process computer of nuclear power plant to the CARE system due to limitation of To enhance the CARE system more effective for CANDU reactors, there is a need to provide complement the feature of the CARE in such a way to providing the operating parameters using to using safety analysis tool such as CANDU Integrated Safety Analysis System (CISAS) for CANDU reactors. In this study, to enhance the safety monitoring measurement two computerized systems such as a CANDU Operational Safety Monitoring System (COSMOS) and prototype of CANDU Emergency Preparedness Advisory System (CEPAS) are developed. This study introduces the two integrated safety monitoring system using the R and D products of the national mid- and long-term R and D such as CISAS and ISSAC code
Research on the improvement of nuclear safety -Thermal hydraulic tests for reactor safety system-
Energy Technology Data Exchange (ETDEWEB)
Jung, Moon Kee; Park, Choon Kyung; Yang, Sun Kyoo; Chun, Se Yung; Song, Chul Hwa; Jun, Hyung Kil; Jung, Heung Joon; Won, Soon Yun; Cho, Yung Roh; Min, Kyung Hoh; Jung, Jang Hwan; Jang, Suk Kyoo; Kim, Bok Deuk; Kim, Wooi Kyung; Huh, Jin; Kim, Sook Kwan; Moon, Sang Kee; Lee, Sang Il [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)
1995-06-01
The present research aims at the development of the thermal hydraulic verification test technology for the safety system of the conventional and advanced nuclear power plant and the development of the advanced thermal hydraulic measuring techniques. In this research, test facilities simulating the primary coolant system and safety system are being constructed for the design verification tests of the existing and advanced nuclear power plant. 97 figs, 14 tabs, 65 refs. (Author).
System Interface for an Integrated Intelligent Safety System (ISS for Vehicle Applications
Directory of Open Access Journals (Sweden)
Mahammad A. Hannan
2010-01-01
Full Text Available This paper deals with the interface-relevant activity of a vehicle integrated intelligent safety system (ISS that includes an airbag deployment decision system (ADDS and a tire pressure monitoring system (TPMS. A program is developed in LabWindows/CVI, using C for prototype implementation. The prototype is primarily concerned with the interconnection between hardware objects such as a load cell, web camera, accelerometer, TPM tire module and receiver module, DAQ card, CPU card and a touch screen. Several safety subsystems, including image processing, weight sensing and crash detection systems, are integrated, and their outputs are combined to yield intelligent decisions regarding airbag deployment. The integrated safety system also monitors tire pressure and temperature. Testing and experimentation with this ISS suggests that the system is unique, robust, intelligent, and appropriate for in-vehicle applications.
A Reliability Assessment Method for the VHTR Safety Systems
International Nuclear Information System (INIS)
Lee, Hyung Sok; Jae, Moo Sung; Kim, Yong Wan
2011-01-01
The Passive safety system by very high temperature reactor which has attracted worldwide attention in the last century is the reliability safety system introduced for the improvement in the safety of the next generation nuclear power plant design. The Passive system functionality does not rely on an external source of energy, but on an intelligent use of the natural phenomena, such as gravity, conduction and radiation, which are always present. Because of these features, it is difficult to evaluate the passive safety on the risk analysis methodology having considered the existing active system failure. Therefore new reliability methodology has to be considered. In this study, the preliminary evaluation and conceptualization are tried, applying the concept of the load and capacity from the reliability physics model, designing the new passive system analysis methodology, and the trial applying to paper plant.
The socio-technical system and nuclear safety
International Nuclear Information System (INIS)
Stefanescu, Petre; Mihailescu, Nicolae; Dragusin, Octavian
1999-01-01
In the field of nuclear safety there have been defined notions like 'technical factors' and 'human factors'. The technical factors depend on designing and manufacturing of components/equipment, actually depend on the people's work. The study of human factors consists in analyzing and recommending the terms that allow an individual to be a reliable and safety agent. Accordingly, he/she is placed in working conditions corresponding to human abilities, associating the means of three levels: - designing, i.e. the action upon the technical system and upon work organization; - correction, i.e. the action upon the evolution of the technical system and organizing; - formation/training, i.e. action upon operators. The paper presents a characterization of the socio-technical system and on this basis discusses the issue of individual adjustment to the socio-technical system and reciprocally, the issue of the socio-technical system adjustment to the individual. Concepts as: ergonomics, physical medium, man/machine interface and support of the operator, man/machine task sharing, the work organizing are put in relation with the central subject, the nuclear safety
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
Recent advances in systems safety and security
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.
An Integrated Safety Assessment Methodology for Generation IV Nuclear Systems
International Nuclear Information System (INIS)
Leahy, Timothy J.
2010-01-01
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. Early work of the RSWG focused on defining a safety philosophy founded on lessons learned from current and prior generations of nuclear technologies, and on identifying technology characteristics that may help achieve Generation IV safety goals. More recent RSWG work has focused on the definition of an integrated safety assessment methodology for evaluating the safety of Generation IV systems. The methodology, tentatively called ISAM, is an integrated 'toolkit' consisting of analytical techniques that are available and matched to appropriate stages of Generation IV system concept development. 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.
[B-BS and occupational health and safety management systems].
Bacchetta, Adriano Paolo
2010-01-01
The objective of a SGSL is the "prevention" agreement as approach of "pro-active" toward the safety at work through the construction of an integrated managerial system in synergic an dynamic way with the business organization, according to continuous improvement principles. Nevertheless the adoption of a SGSL, not could guarantee by itself the obtainment of the full effectiveness than projected and every individual's adhesion to it, must guarantee it's personal involvement in proactive way, so that to succeed to actual really how much hypothesized to systemic level to increase the safety in firm. The objective of a behavioral safety process that comes to be integrated in a SGSL, it has the purpose to succeed in implementing in firm a process of cultural change that raises the workers social group fundamental safety value, producing an ample and full involvement of all in the activities of safety at work development. SGSL = Occupational Health and Safety Management System.
Robertson, Michael F
2018-01-01
Safety management systems (SMS) are becoming the industry standard for safety management throughout the aviation industry. As the Federal Aviation Administration continues to mandate SMS for different segments, the assessment of an organization’s safety culture becomes more important. An SMS can facilitate the development of a strong aviation safety culture. This study describes how safety culture and SMS are integrated. The purpose of this study was to examine the relationship between an ...
International Nuclear Information System (INIS)
Kao, G.C.; Daugherty, W.L.; Barnes, D.M.
1992-09-01
In order to extend the service life of safety related structures and systems in a logical manner, a Structural Enhancement Program was initiated to evaluate the structural integrity of eight (8) systems, namely: Cooling Water System, Emergency Cooling System, Moderator Recovery System supplementary Safety System, Water Removal System, Service Raw Water System, Service Clarified Water System, and River Water System. Since the level of importance of each system to reactor operations varies from one system to another, the scope of structural integrity evaluation for each system should be prioritized accordingly. This paper presents the assessment of system priority for structural evaluation based on a ranking methodology and specifies the level of structural evaluation consistent with the established priority. The effort was undertaken by a five-member panel representing four (4) major disciplines, including. structures, reactor engineering/operations, risk management and materials. The above systems were divided into a total of thirty-five (35) subsystem. These subsystems were then ranked with six (6) attributes, namely: Safety Classification, Degradation Mechanisms, Difficulty of Replacement, Failure Mode, Radiation Dose to Workers and Consequence of Failure. Each attribute was assigned a set of consequences or events with corresponding weighting scores. The results of the ranking process yielded two groups of subsystems, categorized as Priority I and II subsystems. The level of structural assessment was then formulated accordingly. The prioritized approach will allow more efficient allocation of resources, so that the Structural Enhancement Program can be implemented in a cost-effective and efficient manner
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)
A study on optimization of the nuclear safety system
International Nuclear Information System (INIS)
Lee, Sang Hoon; Koh, Byung Joon; Kim, Jin Soo; Kim, Byoung Do; Cho, Seong Won; Kwon, Seog Kwon; Choi, Kwang Sik
1986-12-01
The number of nuclear facilities (nuclear power plants, research reactors, nuclear fuel facilities) under construction or in operation in Korea continues to increase and this has brought about increased importance and concerns toward nuclear safety in Korea. Also, domestic nuclear related organizations are increasingly carrying out the design/construction of nuclear power plants and the development /supply of nuclear fuels. In order to flexibly respond to these changes and to suggest direction to take, it is necessary to re-examine the current nuclear safety regulation system. This study is carried out in two stages and this report describes the results of the analysis and the assessment of the nuclear licencing system of such foreign countries as sweden and German, as the first of the two. In this regard, this study includes the analysis on the backgrounds on the choice of nuclear licensing system, the analysis on the licensing procedures, the analysis on the safety inspection system and the enforcement laws, the analysis on the structure and function of the regulatory, business and research organizations as well as the analysis on the relationship between the safety research and the regulatory duties. In this study, the German safety inspection system and the enforcement procedures and the Swedish nuclear licensing system are analyzed in detail. By comparing and assessing the finding with the current Korea Nuclear Licensing System, this study points out some reform measures of the Korean system that needs to improved. With the changing situations in mind, this study aims to develop the nuclear safety regulation system optimized for Korean situation by re-examining the current regulation system. (Author)
Advancement on safety management system of nuclear power for safety and non-anxiety of society
International Nuclear Information System (INIS)
Yoshikawa, Hidekazu
2004-01-01
Advancement on safety management system is investigated to improve safety and non-anxiety of society for nuclear power, from the standpoint of human machine system research. First, the recent progress of R and D works of human machine interface technologies since 1980 s are reviewed and then the necessity of introducing a new approach to promote technical risk communication activity to foster safety culture in nuclear industries. Finally, a new concept of Offsite Operation and Maintenance Support Center (OMSC) is proposed as the core facility to assemble human resources and their expertise in all organizations of nuclear power, for enhancing safety and non-anxiety of society for nuclear power. (author)
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.
Progress in the development of methodology for fusion safety systems studies
International Nuclear Information System (INIS)
Ho, S.K.; Cambi, G.; Ciattaglia, S.; Fujii-e, Y.; Seki, Y.
1994-01-01
The development of fusion safety systems-study methodology, including the aspects of schematic classification of overall fusion safety system, qualitative assessment of fusion system for identification of critical accident scenarios, quantitative analysis of accident consequences and risk for safety design evaluation, and system-level analysis of accident consequences and risk for design optimization, by a consortium of international efforts is presented. The potential application of this methodology into reactor design studies will facilitate the systematic assessment of safety performance of reactor designs and enhance the impacts of safety considerations on the selection of design configurations
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
Effect Analysis of Digital I and C Systems on Plant Safety based on Fault-Tree Analysis
International Nuclear Information System (INIS)
Lee, Seung Jun; Jung, Wondea
2014-01-01
Deterioration and an inadequate supply of components of analog I and C systems have led to inefficient and costly maintenance. Moreover, since the fast evolution of digital technology has enabled more reliable functions to be designed for NPP safety, the transition from analog to digital has been accelerated. Owing to the distinguishable characteristics of digital I and C systems, a reliability analysis of digital systems has become an important element of a probabilistic safety assessment (PSA). Digital I and C systems have unique characteristics such as fault-tolerant techniques and software. However, these features have not been properly considered yet in most NPP PSA models. The effect of digital I and C systems should be evaluated by comparing them to that of analog I and C systems. Before installing a digital I and C system, even though it is expected that the plant safety can be improved through the advantageous features of digital I and C systems, it should be validated whether the total NPP safety is better than analog systems or is the same at least. In this work, the fault-tree (FT) technique, which is most widely used in a PSA, was used to compare the effects of analog and digital I and C systems. From a case study, the results of plant safety were compared. In this work, the effect of a digital RPS was evaluated by comparing it to that of an analog RPS based on the FT models. In the evaluation results, it was observed that digital RPS has a positive effect on reducing the system unavailability. The analysis results can be used for the development of a guide for evaluating digital I and C systems and reliability requirements
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...
Giannini, Andrea; Russo, Eleonora; Mannella, Paolo; Palla, Giulia; Pisaneschi, Silvia; Cecchi, Elena; Maremmani, Michele; Morelli, Luca; Perutelli, Alessandra; Cela, Vito; Melfi, Franca; Simoncini, Tommaso
2017-08-01
To present the first case series of total robotic hysterectomy (TRH), using integrated table motion (ITM), which is a new feature comprising a unique operating table by Trumpf Medical that communicates wirelessly with the da Vinci Xi surgical system. ITM has been specifically developed to improve multiquadrant robotic surgery such as that conducted in colorectal surgery. Between May and October 2015, a prospective post-market study was conducted on ITM in the EU in 40 cases from different specialties. The gynecological study group comprised 12 patients. Primary endpoints were ITM feasibility, safety and efficacy. Ten patients underwent TRH. Mean number of ITM moves was three during TRH; there were 31 instances of table moves in the ten procedures. Twenty-eight of 31 ITM moves were made to gain internal exposure. The endoscope remained inserted during 29 of the 31 table movements (94%), while the instruments remained inserted during 27 of the 31 moves (87%). No external instrument collisions or other problems related to the operating table were noted. There were no ITM safety-related observations and no adverse events. This preliminary study demonstrated the feasibility, safety and efficacy of ITM for the da Vinci Xi surgical system in TRH. ITM was safe, with no adverse events related to its use. Further studies will be useful to define the real role and potential benefit of ITM in gynecological surgery.
Research on the Evaluation System for Rural Public Safety Planning
Institute of Scientific and Technical Information of China (English)
Ming; SUN; Jianxin; YAN
2014-01-01
The indicator evaluation system is introduced to the study of rural public safety planning in this article.By researching the current rural public safety planning and environmental carrying capacity,we select some carrying capacity indicators influencing the rural public safety,such as land,population,ecological environment,water resources,infrastructure,economy and society,to establish the environmental carrying capacity indicator system.We standardize the indicators,use gray correlation analysis method to determine the weight of indicators,and make DEA evaluation of the indicator system,to obtain the evaluation results as the basis for decision making in rural safety planning,and provide scientific and quantified technical support for rural public safety planning.
Verification and validation issues for digitally-based NPP safety systems
International Nuclear Information System (INIS)
Ets, A.R.
1993-01-01
The trend toward standardization, integration and reduced costs has led to increasing use of digital systems in reactor protection systems. While digital systems provide maintenance and performance advantages, their use also introduces new safety issues, in particular with regard to software. Current practice relies on verification and validation (V and V) to ensure the quality of safety software. However, effective V and V must be done in conjunction with a structured software development process and must consider the context of the safety system application. This paper present some of the issues and concerns that impact on the V and V process. These include documentation of systems requirements, common mode failures, hazards analysis and independence. These issues and concerns arose during evaluations of NPP safety systems for advanced reactor designs and digital I and C retrofits for existing nuclear plants in the United States. The pragmatic lessons from actual systems reviews can provide a basis for further refinement and development of guidelines for applying V and V to NPP safety systems. (author). 14 refs
33 CFR 96.220 - What makes up a safety management system?
2010-07-01
... system? 96.220 Section 96.220 Navigation and Navigable Waters COAST GUARD, DEPARTMENT OF HOMELAND SECURITY VESSEL OPERATING REGULATIONS RULES FOR THE SAFE OPERATION OF VESSELS AND SAFETY MANAGEMENT SYSTEMS Company and Vessel Safety Management Systems § 96.220 What makes up a safety management system? (a) The...
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.
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.)
The Influence of Errors in Visualization Systems on the Level of Safety Threat in Air Traffic
Directory of Open Access Journals (Sweden)
Paweł Ferduła
2018-01-01
Full Text Available Air traffic management is carried out by air traffic controllers assisted by complex technical systems that provide them with visualization of the traffic situation. In practice, visualization systems errors sometimes occur. The purpose of this paper is to determine the impact of errors of different types on the safety of the air traffic. The assessment of the threat level is influenced by subjective factors and cannot be expressed precisely. Therefore, the fuzzy reasoning theory has been used. The developed fuzzy model has been used to obtain a tool for simulation of the impact of various factors on traffic safety assessment. The results obtained indicate that the most important determinants of safety are the time when the air traffic controller remains unaware of the breakdown and the total time he/she does not have full knowledge of the traffic situation. It has been found that the key role for the proper operation of the air traffic visualization system and the restoration of full situational awareness is played by self-diagnostic systems that can restore the system’s correct functioning without even the controller being aware of the error occurrence. Their role in ensuring safety might be even greater than redundancy which is commonly used.
Plutonium finishing plant safety systems and equipment list
International Nuclear Information System (INIS)
Bergquist, G.G.
1995-01-01
The Safety Equipment List (SEL) supports Analysis Report (FSAR), WHC-SD-CP-SAR-021 and the Plutonium Finishing Plant Operational Safety Requirements (OSRs), WHC-SD-CP-OSR-010. The SEL is a breakdown and classification of all Safety Class 1, 2, and 3 equipment, components, or system at the Plutonium Finishing Plant complex
Development of Non-safety System Architecture and Evaluation of Components/Systems
International Nuclear Information System (INIS)
Oh, I. S.; Lee, C. K.; Kim, D. H.; Lee, J. W.; Lee, D. Y.; Park, W. M.; Hwang, I. K.; Hur, S.; Kim, J. T.; Park, J. C.; Lee, J. W.
2007-10-01
We describe in this report the works performed for a technical evaluation of the non-safety digital control system of the KNICS, the non-safety process control system of the KNICS, a communication load analysis for the MMIS (including both the non-safety and the safety systems) of the KNICS, the development of MMI and an implementation of the logic for the CVCS, and the works performed to support writing a proposal needed for bidding an I and C system based on the KNICS. The technical evaluation results were aimed to be used by the designers to detect parts needed to be corrected or to be newly inserted, and also by the developers during the development phase. The requirement specifications and the data requirement characteristics have been identified for each subsystem of the determined KNICS structure. For each communication node, the specifications related to the data transfer including the data capacity for interfaces, delay time for the data transfer, and the marginal availability of its performance capabilities have been analyzed to identify the amount of data transfer and hence to verify that both of the designed structures for the safety related communications network and for the digital communications network are appropriate. The results of the supporting work performed for writing the technical specifications related to each subsystem of the KNICS structure, are expected to be useful in writing a proposal for the expected Uljin new units 1 and 2, and in the I and C upgrade for any of the existing nuclear power plants under operation. Also included in this report are the descriptions on a design of the chemical volume control system (CVCS), on the supporting work performed to draw the logic diagrams for CVCS using the tool ISaGRAF, and on the generation of a set of system displays to be used as references
Development of Non-safety System Architecture and Evaluation of Components/Systems
Energy Technology Data Exchange (ETDEWEB)
Oh, I. S.; Lee, C. K.; Kim, D. H.; Lee, J. W.; Lee, D. Y.; Park, W. M.; Hwang, I. K.; Hur, S.; Kim, J. T.; Park, J. C.; Lee, J. W
2007-10-15
We describe in this report the works performed for a technical evaluation of the non-safety digital control system of the KNICS, the non-safety process control system of the KNICS, a communication load analysis for the MMIS (including both the non-safety and the safety systems) of the KNICS, the development of MMI and an implementation of the logic for the CVCS, and the works performed to support writing a proposal needed for bidding an I and C system based on the KNICS. The technical evaluation results were aimed to be used by the designers to detect parts needed to be corrected or to be newly inserted, and also by the developers during the development phase. The requirement specifications and the data requirement characteristics have been identified for each subsystem of the determined KNICS structure. For each communication node, the specifications related to the data transfer including the data capacity for interfaces, delay time for the data transfer, and the marginal availability of its performance capabilities have been analyzed to identify the amount of data transfer and hence to verify that both of the designed structures for the safety related communications network and for the digital communications network are appropriate. The results of the supporting work performed for writing the technical specifications related to each subsystem of the KNICS structure, are expected to be useful in writing a proposal for the expected Uljin new units 1 and 2, and in the I and C upgrade for any of the existing nuclear power plants under operation. Also included in this report are the descriptions on a design of the chemical volume control system (CVCS), on the supporting work performed to draw the logic diagrams for CVCS using the tool ISaGRAF, and on the generation of a set of system displays to be used as references.
International Nuclear Information System (INIS)
Haapanen, P.; Korhonen, J.
1995-01-01
The safety assessment of programmable automation systems cannot be totally be based on conventional probabilistic methods because of the difficulties in quantification of the reliability of the software as well as the hardware. Additional means shall therefore be used to gain more confidence on the system dependability. One central confidence building measure is the independent dynamic testing of the completed system. An automated test harness is needed to run the required large amount of test cases in a restricted time span. This paper describes a prototype dynamic testing harness for programmable digital systems developed at VTT. (author). 12 refs, 2 figs, 2 tabs
International Nuclear Information System (INIS)
Yang Xiaohua; Liu Zhenghai; Liu Zhiming; Wan Yaping; Bai Xiaofeng
2012-01-01
We can improve the processing, the evaluation of capability and promote the user's trust by using system security engineering capability maturity model (SSE-CMM). SSE-CMM is the common method for organizing and implementing safety engineering, and it is a mature method for system safety engineering. Combining capability maturity model (CMM) with total quality management and statistic theory, SSE-CMM turns systems security engineering into a well-defined, mature, measurable, advanced engineering discipline. Lack of domain knowledge, the size of data, the diversity of evidences, the cumbersomeness of processes, and the complexity of matching evidences with problems are the main issues that SSE-CMM assessment has to face. To improve effectively the efficiency of assessment of spent fuel reprocessing system security engineering capability maturity model (SFR-SSE-CMM), in this paper we de- signed an intelligent assessment software based on domain ontology and that uses methods such as ontology, evidence theory, semantic web, intelligent information retrieval and intelligent auto-matching techniques. This software includes four subsystems, which are domain ontology creation and management system, evidence auto collection system, and a problem and evidence matching system. The architecture of the software is divided into five layers: a data layer, an oncology layer, a knowledge layer, a service layer arid a presentation layer. (authors)
The Management System for Nuclear Installations Safety Guide
International Nuclear Information System (INIS)
2009-01-01
This Safety Guide is applicable throughout the lifetime of a nuclear installation, including any subsequent period of institutional control, until there is no significant residual radiation hazard. For a nuclear installation, the lifetime includes site evaluation, design, construction, commissioning, operation and decommissioning. These stages in the lifetime of a nuclear installation may overlap. This Safety Guide may be applied to nuclear installations in the following ways: (a)To support the development, implementation, assessment and improvement of the management system of those organizations responsible for research, site evaluation, design, construction, commissioning, operation and decommissioning of a nuclear installation; (b)As an aid in the assessment by the regulatory body of the adequacy of the management system of a nuclear installation; (c)To assist an organization in specifying to a supplier, via contractual documentation, any specific element that should be included within the supplier's management system for the supply of products. This Safety Guide follows the structure of the Safety Requirements publication on The Management System for Facilities and Activities, whereby: (a)Section 2 provides recommendations on implementing the management system, including recommendations relating to safety culture, grading and documentation. (b)Section 3 provides recommendations on the responsibilities of senior management for the development and implementation of an effective management system. (c)Section 4 provides recommendations on resource management, including guidance on human resources, infrastructure and the working environment. (d)Section 5 provides recommendations on how the processes of the installation can be specified and developed, including recommendations on some generic processes of the management system. (e)Section 6 provides recommendations on the measurement, assessment and improvement of the management system of a nuclear installation. (f
Mitsutake, Yoshiaki; Ebner, Adrian; Yeung, Alan C; Taber, Mark D; Davidson, Charles J; Ikeno, Fumiaki
2015-02-15
To report our initial animal and human experience with a new multi-lumen catheter called MultiCross™ (Roxwood Medical, Inc.) in a porcine coronary model and patients with a chronic total occlusion (CTO). Preclinical safety study was done in the coronary vasculature of a porcine model. In a clinical setting, patients with a CTO of a coronary artery (n = 5) were enrolled. After an initial unsuccessful attempt using a conventional guidewire, operators could use the MultiCross system. The primary efficacy endpoint was successful recanalization (technical success) and the primary safety endpoint was serious adverse events through 30 days post-procedure. The MultiCross catheter was used for all patients after failure of the initial attempt with a guidewire. Successful recanalization was achieved in all CTOs attempted (100%). No patients reported any adverse events at 30 days post-procedure. In this first-in-man experience, the MultiCross catheter has the potential to enhance crossing of CTOs. © 2014 Wiley Periodicals, Inc.
Robertson, Mike Fuller
2017-01-01
Safety Management Systems (SMS) are becoming the industry standard for safety management throughout the aviation industry. As the Federal Aviation Administration (FAA) continues to mandate SMS for different segments, the assessment of an organization's safety culture becomes more important. An SMS can facilitate the development of a strong…
International Nuclear Information System (INIS)
2010-01-01
This Safety Guide was prepared under the IAEA programme for establishing safety standards for nuclear power plants. The basic requirements for the design of safety systems for nuclear power plants are established in the Safety Requirements publication, Safety Standards Series No. NS-R-1 on Safety of Nuclear Power Plants: Design, which it supplements. This Safety Guide describes how the requirements for the design of the reactor coolant system (RCS) and associated systems in nuclear power plants should be met. This publication is a revision and combination of two previous Safety Guides, Safety Series No. 50-SG-D6 on Ultimate Heat Sink and Directly Associated Heat Transport Systems for Nuclear Power Plants (1982), and Safety Series No. 50-SG-D13 on Reactor Coolant and Associated Systems in Nuclear Power Plants (1987), which are superseded by this new Safety Guide. The revision takes account of developments in the design of the RCS and associated systems in nuclear power plants since the earlier Safety Guides were published in 1982 and 1987, respectively. The other objectives of the revision are to ensure consistency with Ref., issued in 2004, and to update the technical content. In addition, an appendix on pressurized heavy water reactors (PHWRs) has been included.
European Workshop Industrical Computer Science Systems approach to design for safety
Zalewski, Janusz
1992-01-01
This paper presents guidelines on designing systems for safety, developed by the Technical Committee 7 on Reliability and Safety of the European Workshop on Industrial Computer Systems. The focus is on complementing the traditional development process by adding the following four steps: (1) overall safety analysis; (2) analysis of the functional specifications; (3) designing for safety; (4) validation of design. Quantitative assessment of safety is possible by means of a modular questionnaire covering various aspects of the major stages of system development.
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
FULCRUM - A dam safety management and alert system
Energy Technology Data Exchange (ETDEWEB)
Butt, Cameron; Greenaway, Graham [Knight Piesold Ltd., Vancouver, (Canada)
2010-07-01
Efficient management of instrumentation, monitoring and inspection data are the keys to safe performance and dam structure stability. This paper presented a data management system, FULCRUM, developed for dam safety management. FULCRUM is a secure web-based data management system which simplifies the process of data collection, processing and analysis of the information. The system was designed to organize and coordinate dam safety management requirements. Geotechnical instrumentation such as piezometers or inclinometers and operating data can be added to the database. Data from routine surveillance and engineering inspection can also be incorporated into the database. The system provides users with immediate access to historical and recent data. The integration of a GIS system allows for rapid assessment of the project site. Customisable alerting protocols can be set to identify and respond quickly to significant changes in operating conditions and potential impacts on dam safety.
International Nuclear Information System (INIS)
Torbioern, Ingemar; Mattson, Malin
2009-03-01
The aim of this pilot study has been to describe and analyse potential effects on safety-related behaviour and risks associated with the bonus systems currently used at Swedish nuclear plants. To this end and in order to establish a frame of reference several theories on motivation were consulted regarding the relevance of monetary rewards. In addition empirical evidence on effects upon behaviours in general and safety behaviours in particular was taken into consideration, as well as a systems and a rationalist perspective on organisations. The resulting frame of reference was used for a descriptive mapping of the bonus systems and for the formulation of a semi-structured interview schedule intended to capture the experiences of those concerned by the systems. A total of 15 interviews were performed with staff of different functions and organisational positions. Results of the study do not indicate any negative effects on safety-related behaviours. Rather they indicate that safety-behaviours may be promoted insofar as bonus rewards are linked to performance goals concerning safety. All of the bonus-systems may be characterised as low in incentive intensity, i.e. produce small effects on motivation and performance. Still, as the systems differ in design and in the way they are perceived, they also represent different challenges in order to function more efficiently as parameters
International Nuclear Information System (INIS)
Ryazanov, Boris G.; Sviridov, Victor I.; Frolov, Vladimir V.; Shvedov, Maxim O.; Mclaughlin, Thomas P.; Pruvost, Norman L.
2003-01-01
A considerable body of data has now been acquired about the principles, parameters and consequences of nuclear (criticality) accidents at facilities of the atomic industry in Russia, the United States, Great Britain and Japan. The total number of such accidents stands at 22. Russian and US specialists have prepared a rather extensive survey and analysis of these accidents. The final and important section of this survey is the lessons implied by the results of analysis of these 22 accidents. Among these lessons is the necessity of unconditional enforcement of control over the movement and transformations of special nuclear materials (SNM), and in particular fissile materials, (those SNMs with criticality accident concerns) during production and processing. Inadequacies in such control have been among the causes of most of the accidents that have occurred. Nuclear materials control and accounting (MC and A) for the purpose of ensuring storage reliability and nonproliferation safeguards is a major task of nuclear facilities in any nation. MC and A systems use the latest techniques and hardware for periodic control of SNM in specifically organized material balance areas. Immediate checking, periodic inventory of SNM, and measurements of the parameters of SNM at key points are the main sources of data for these systems. Data about the presence and sites of location of SNM in material balance areas that are acquired in inventories can be used for objective assessment of the status of nuclear safety. On the other hand, the inventory itself involves performance of operations that are unlike routine process engineering, and require special consideration of nuclear safety. Use of the techniques and hardware of MC and A systems not only for purposes of storage reliability, but also to ensure nuclear safety, will reduce the risk of nuclear accidents. This paper gives a concise overview of nuclear accidents that have occurred due to inadequacies in MC and A, and demonstrates
Nuclear-power-safety reporting system: feasibility analysis
International Nuclear Information System (INIS)
Finlayson, F.C.; Ims, J.
1983-04-01
The US Nuclear Regulatory Commission (NRC) is evaluating the possibility of instituting a data gathering system for identifying and quantifying the factors that contribute to the occurrence of significant safety problems involving humans in nuclear power plants. This report presents the results of a brief (6 months) study of the feasibility of developing a voluntary, nonpunitive Nuclear Power Safety Reporting System (NPSRS). Reports collected by the system would be used to create a data base for documenting, analyzing and assessing the significance of the incidents. Results of The Aerospace Corporation study are presented in two volumes. This document, Volume I, contains a summary of an assessment of the Aviation Safety Reporting System (ASRS). The FAA-sponsored, NASA-managed ASRS was found to be successful, relatively low in cost, generally acceptable to all facets of the aviation community, and the source of much useful data and valuable reports on human factor problems in the nation's airways. Several significant ASRS features were found to be pertinent and applicable for adoption into a NPSRS
Evaluating software for safety systems in nuclear power plants
International Nuclear Information System (INIS)
Lawrence, J.D.; Persons, W.L.; Preckshot, G.G.; Gallagher, J.
1994-01-01
In 1991, LLNL was asked by the NRC to provide technical assistance in various aspects of computer technology that apply to computer-based reactor protection systems. This has involved the review of safety aspects of new reactor designs and the provision of technical advice on the use of computer technology in systems important to reactor safety. The latter includes determining and documenting state-of-the-art subjects that require regulatory involvement by the NRC because of their importance in the development and implementation of digital computer safety systems. These subjects include data communications, formal methods, testing, software hazards analysis, verification and validation, computer security, performance, software complexity and others. One topic software reliability and safety is the subject of this paper
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
Tompa, Emile; Robson, Lynda; Sarnocinska-Hart, Anna; Klassen, Robert; Shevchenko, Anton; Sharma, Sharvani; Hogg-Johnson, Sheilah; Amick, Benjamin C; Johnston, David A; Veltri, Anthony; Pagell, Mark
2016-03-01
The aim of this study was to determine whether management system practices directed at both occupational health and safety (OHS) and operations (joint management system [JMS] practices) result in better outcomes in both areas than in alternative practices. Separate regressions were estimated for OHS and operational outcomes using data from a survey along with administrative records on injuries and illnesses. Organizations with JMS practices had better operational and safety outcomes than organizations without these practices. They had similar OHS outcomes as those with operations-weak practices, and in some cases, better outcomes than organizations with safety-weak practices. They had similar operational outcomes as those with safety-weak practices, and better outcomes than those with operations-weak practices. Safety and operations appear complementary in organizations with JMS practices in that there is no penalty for either safety or operational outcomes.
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
A hybrid approach to quantify software reliability in nuclear safety systems
International Nuclear Information System (INIS)
Arun Babu, P.; Senthil Kumar, C.; Murali, N.
2012-01-01
Highlights: ► A novel method to quantify software reliability using software verification and mutation testing in nuclear safety systems. ► Contributing factors that influence software reliability estimate. ► Approach to help regulators verify the reliability of safety critical software system during software licensing process. -- Abstract: Technological advancements have led to the use of computer based systems in safety critical applications. As computer based systems are being introduced in nuclear power plants, effective and efficient methods are needed to ensure dependability and compliance to high reliability requirements of systems important to safety. Even after several years of research, quantification of software reliability remains controversial and unresolved issue. Also, existing approaches have assumptions and limitations, which are not acceptable for safety applications. This paper proposes a theoretical approach combining software verification and mutation testing to quantify the software reliability in nuclear safety systems. The theoretical results obtained suggest that the software reliability depends on three factors: the test adequacy, the amount of software verification carried out and the reusability of verified code in the software. The proposed approach may help regulators in licensing computer based safety systems in nuclear reactors.
Safety and reliability of automatization software
Energy Technology Data Exchange (ETDEWEB)
Kapp, K; Daum, R [Karlsruhe Univ. (TH) (Germany, F.R.). Lehrstuhl fuer Angewandte Informatik, Transport- und Verkehrssysteme
1979-02-01
Automated technical systems have to meet very high requirements concerning safety, security and reliability. Today, modern computers, especially microcomputers, are used as integral parts of those systems. In consequence computer programs must work in a safe and reliable mannter. Methods are discussed which allow to construct safe and reliable software for automatic systems such as reactor protection systems and to prove that the safety requirements are met. As a result it is shown that only the method of total software diversification can satisfy all safety requirements at tolerable cost. In order to achieve a high degree of reliability, structured and modular programming in context with high level programming languages are recommended.
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)
Patient safety - the role of human factors and systems engineering.
Carayon, Pascale; Wood, Kenneth E
2010-01-01
Patient safety is a global challenge that requires knowledge and skills in multiple areas, including human factors and systems engineering. In this chapter, numerous conceptual approaches and methods for analyzing, preventing and mitigating medical errors are described. Given the complexity of healthcare work systems and processes, we emphasize the need for increasing partnerships between the health sciences and human factors and systems engineering to improve patient safety. Those partnerships will be able to develop and implement the system redesigns that are necessary to improve healthcare work systems and processes for patient safety.
Patient Safety: The Role of Human Factors and Systems Engineering
Carayon, Pascale; Wood, Kenneth E.
2011-01-01
Patient safety is a global challenge that requires knowledge and skills in multiple areas, including human factors and systems engineering. In this chapter, numerous conceptual approaches and methods for analyzing, preventing and mitigating medical errors are described. Given the complexity of healthcare work systems and processes, we emphasize the need for increasing partnerships between the health sciences and human factors and systems engineering to improve patient safety. Those partnerships will be able to develop and implement the system redesigns that are necessary to improve healthcare work systems and processes for patient safety. PMID:20543237
Safety assessment of complex engineered and natural systems: radioactive waste disposal
International Nuclear Information System (INIS)
McNeish, J.A.; Vallikat, V.; Atkins, J.; Balady, M.A.
1997-01-01
Evaluation of deep, geologic disposal of nuclear waste requires the probabilistic safety assessment of a complex system from the coupling of various processes and sub-systems, parameter and model uncertainties, spatial and temporal variabilities, and the multiplicity of designs and scenarios. Both the engineered and natural system are included in the evaluation. Each system has aspects with considerable uncertainty both in important parameters and in overall conceptual models. The study represented herein provides a probabilistic safety assessment of a potential respository system for multiple engineered barrier system (EBS) design and conceptual model configurations (CRWMS M and O, 1996a) and considers the effects of uncertainty on the overall results. The assessment is based on data and process models available at the time of the study and doesnt necessarily represent the current safety evaluation. In fact, the percolation flux through the repository system is now expected to be higher than the estimate used for this study. The potential effects of higher percolation fluxes are currently under study. The safety of the system was assessed for both 10,000 and 1,000,000 years. Use of alternative conceptual models also produced major improvement in safety. For example, use of a more realistic engineered system release model produced improvement of over an order of magnitude in safety. Alternative measurement locations for the safety assessment produced substantial increases in safety, through the results are based on uncertain dilution factors in the transporting groundwater. (Author)
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)
Behavior based safety process - a pragmatic approach
International Nuclear Information System (INIS)
Sharma, R.K.; Malaikar, N.L.; Belokar, S.G.; Arora, Yashpal
2009-01-01
Materials handling, processing and storage of hazardous chemicals has grown exponentially. The chemical industries has reacted to the situation by introducing numerous safety systems such as IS18001, 'HAZOP', safety audits, risk assessment, training etc, which has reduced hazards and improved safety performance, but has not totally eliminated exposure to the hazards. These safety systems aim to bring change in attitude of the persons which is difficult to change or control. However, behaviour of plant personnel can be controlled or improved upon, which should be our aim. (author)
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
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.
Software reliability and safety in nuclear reactor protection systems
Energy Technology Data Exchange (ETDEWEB)
Lawrence, J.D. [Lawrence Livermore National Lab., CA (United States)
1993-11-01
Planning the development, use and regulation of computer systems in nuclear reactor protection systems in such a way as to enhance reliability and safety is a complex issue. This report is one of a series of reports from the Computer Safety and Reliability Group, Lawrence Livermore that investigates different aspects of computer software in reactor National Laboratory, that investigates different aspects of computer software in reactor protection systems. There are two central themes in the report, First, software considerations cannot be fully understood in isolation from computer hardware and application considerations. Second, the process of engineering reliability and safety into a computer system requires activities to be carried out throughout the software life cycle. The report discusses the many activities that can be carried out during the software life cycle to improve the safety and reliability of the resulting product. The viewpoint is primarily that of the assessor, or auditor.
Software reliability and safety in nuclear reactor protection systems
International Nuclear Information System (INIS)
Lawrence, J.D.
1993-11-01
Planning the development, use and regulation of computer systems in nuclear reactor protection systems in such a way as to enhance reliability and safety is a complex issue. This report is one of a series of reports from the Computer Safety and Reliability Group, Lawrence Livermore that investigates different aspects of computer software in reactor National Laboratory, that investigates different aspects of computer software in reactor protection systems. There are two central themes in the report, First, software considerations cannot be fully understood in isolation from computer hardware and application considerations. Second, the process of engineering reliability and safety into a computer system requires activities to be carried out throughout the software life cycle. The report discusses the many activities that can be carried out during the software life cycle to improve the safety and reliability of the resulting product. The viewpoint is primarily that of the assessor, or auditor
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
International Nuclear Information System (INIS)
Ferrario, E.; Zio, E.
2014-01-01
We adopt a ‘system-of-systems’ framework of analysis, previously presented by the authors, to include the interdependent infrastructures which support a critical plant in the study of its safety with respect to the occurrence of an earthquake. We extend the framework to consider the recovery of the system of systems in which the plant is embedded. As a test system, we consider the impacts produced on a nuclear power plant (the critical plant) embedded in the connected power and water distribution, and transportation networks which support its operation. The Seismic Probabilistic Risk Assessment of such system of systems is carried out by Hierarchical modeling and Monte Carlo simulation. First, we perform a top-down analysis through a hierarchical model to identify the elements that at each level have most influence in restoring safety, adopting the criticality importance measure as a quantitative indicator. Then, we evaluate by Monte Carlo simulation the probability that the nuclear power plant enters in an unsafe state and the time needed to recover its safety. The results obtained allow the identification of those elements most critical for the safety and recovery of the nuclear power plant; this is relevant for determining improvements of their structural/functional responses and supporting the decision-making process on safety critical-issues. On the test system considered, under the given assumptions, the components of the external and internal water systems (i.e., pumps and pool) turn out to be the most critical for the safety and recovery of the plant. - Highlights: • We adopt a system-of-system framework to analyze the safety of a critical plant exposed to risk from external events, considering also the interdependent infrastructures that support the plant. • We develop a hierarchical modeling framework to represent the system of systems, accounting also for its recovery. • Monte Carlo simulation is used for the quantitative evaluation of the
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.
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
Safety implications of electronic driving support systems : an orientation.
Gundy, C.M. Steyvers, F.J.J.M. & Kaptein, N.A.
1995-01-01
This report focuses on traffic safety aspects of driving support systems. The report consists of two parts. First of all, the report discusses a number of topics, relevant for the implementation and evaluation of driving support systems. These topics include: (1) safety research into driving support systems: (2) the importance of research into driver models and the driving task; (3) horizontal integration of driving support systems; (4) vertical integration of driving support systems; (5) tas...
Systems Analysis of NASA Aviation Safety Program: Final Report
Jones, Sharon M.; Reveley, Mary S.; Withrow, Colleen A.; Evans, Joni K.; Barr, Lawrence; Leone, Karen
2013-01-01
A three-month study (February to April 2010) of the NASA Aviation Safety (AvSafe) program was conducted. This study comprised three components: (1) a statistical analysis of currently available civilian subsonic aircraft data from the National Transportation Safety Board (NTSB), the Federal Aviation Administration (FAA), and the Aviation Safety Information Analysis and Sharing (ASIAS) system to identify any significant or overlooked aviation safety issues; (2) a high-level qualitative identification of future safety risks, with an assessment of the potential impact of the NASA AvSafe research on the National Airspace System (NAS) based on these risks; and (3) a detailed, top-down analysis of the NASA AvSafe program using an established and peer-reviewed systems analysis methodology. The statistical analysis identified the top aviation "tall poles" based on NTSB accident and FAA incident data from 1997 to 2006. A separate examination of medical helicopter accidents in the United States was also conducted. Multiple external sources were used to develop a compilation of ten "tall poles" in future safety issues/risks. The top-down analysis of the AvSafe was conducted by using a modification of the Gibson methodology. Of the 17 challenging safety issues that were identified, 11 were directly addressed by the AvSafe program research portfolio.
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
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
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.)
Safety analysis of tritium processing system based on PHA
International Nuclear Information System (INIS)
Fu Wanfa; Luo Deli; Tang Tao
2012-01-01
Safety analysis on primary confinement of tritium processing system for TBM was carried out with Preliminary Hazard Analysis. Firstly, the basic PHA process was given. Then the function and safe measures with multiple confinements about tritium system were described and analyzed briefly, dividing the two kinds of boundaries of tritium transferring through, that are multiple confinement systems division and fluid loops division. Analysis on tritium releasing is the key of PHA. Besides, PHA table about tritium releasing was put forward, the causes and harmful results being analyzed, and the safety measures were put forward also. On the basis of PHA, several kinds of typical accidents were supposed to be further analyzed. And 8 factors influencing the tritium safety were analyzed, laying the foundation of evaluating quantitatively the safety grade of various nuclear facilities. (authors)
Software Safety Life cycle and Method of POSAFE-Q System
International Nuclear Information System (INIS)
Lee, Jang-Soo; Kwon, Kee-Choon
2006-01-01
This paper describes the relationship between the overall safety life cycle and the software safety life cycle during the development of the software based safety systems of Nuclear Power Plants. This includes the design and evaluation activities of components as well as the system. The paper also compares the safety life cycle and planning activities defined in IEC 61508 with those in IEC 60880, IEEE 7-4.3.2, and IEEE 1228. Using the KNICS project as an example, software safety life cycle and safety analysis methods applied to the POSAFE-Q are demonstrated. KNICS software safety life cycle is described by comparing to the software development, testing, and safety analysis process with international standards. The safety assessment of the software for POSAFE-Q is a joint Korean German project. The assessment methods applied in the project and the experiences gained from this project are presented
Design Information from the PSA for Digital Safety-Critical Systems
International Nuclear Information System (INIS)
Kang, Hyun Gook; Jang, Seung Cheol
2005-01-01
Many safety-critical applications such as nuclear field application usually adopt a similar design strategy for digital safety-critical systems. Their differences from the normal design for the non-safety-critical applications could be summarized as: multiple-redundancy, highly reliable components, strengthened monitoring mechanism, verified software, and automated test procedure. These items are focusing on maintaining the capability to perform the given safety function when it is requested. For the past several decades, probabilistic safety assessment (PSA) techniques are used in the nuclear industry to assess the relative effects of contributing events on plant risk and system reliability. They provide a unifying means of assessing physical faults, recovery processes, contributing effects, human actions, and other events that have a high degree of uncertainty. The applications of PSA provide not only the analysis results of already installed system but also the useful information for the system under design. The information could be derived from the PSA experience of the various safety-critical systems. Thanks to the design flexibility, the digital system is one of the most suitable candidates for risk-informed design (RID). In this article, we will describe the feedbacks for system design and try to develop a procedure for RID. Even though the procedure is not sophisticated enough now, it could be the start point of the further investigation for developing more complete and practical methodology
Guidelines for implementation of RCM on safety systems
International Nuclear Information System (INIS)
Kim, Tae Woon; Brijendra Singh.
1996-04-01
Reliability Centered Maintenance (RCM) methodology was originally developed by the commercial airlines industry in the early 1960s for identifying applicable and effective preventive maintenance tasks and as currently used in nuclear power industry. Effective maintenance of the systems at a nuclear power plant (NPP) is essential for its safe and reliable operation. Reliability Centered Maintenance at NPP is the program to assure that plant systems remain within an original design criteria and are not adversely affected during the plant life time. The aim of this report is to provide the guidelines to implement the RCM approach on NPP safety systems. Safety systems are usually standby and therefore, we need to periodically detect and repair failures that may have occurred since the previous activation or inspection the equipment. The RCM guidelines are intended to help identify the failure modes and related root causes and then decide the maintenance policies to achieve the high level of safety and reliability. The RCM is intended to improve or maintain high levels of system reliability and plant availability. Since the reliability of plant systems will be improved, the plant safety correspondingly will be increased. Another goal of RCM is to optimize the maintenance and surveillance tasks such that the overall level of resources required to accomplish essential tasks is kept to minimum. RCM also strives to eliminate unnecessary corrective maintenance and to select yet most cost-effective approach to maintenance, testing and inspection for system components. 9 refs. (Author) .new
An intelligent safety system concept for future CANDU reactors
International Nuclear Information System (INIS)
Hinds, H.W.
1980-01-01
A review of the current Regional Over-power Trip (ROPT) system employed on the Bruce NGS-A reactors confirmed the belief that future reactors should have an improved ROPT system. We are developing such an 'intelligent' safety system. It uses more of the available information on reactor status and employs modern computer technology. Fast triplicated safety computers compute maps of fuel channel power, based on readings from prompt-responding flux detectors. The coefficients for this calculation are downloaded periodically from a fourth supervisor computer. These coefficients are based on a detailed 3-D flux shape derived from physics data and other plant information. A demonstration of one of three safety channels of such a system is planned. (auth)
Safety system for child pillion riders of underbone motorcycles in Malaysia.
Sivasankar, S; Karmegam, K; Bahri, M T Shamsul; Naeini, H Sadeghi; Kulanthayan, S
2014-01-01
Motorcycles are a common mode of transport for most Malaysians. Underbone motorcycles are one of the most common types of motorcycle used in Malaysia due to their affordable price and ease of use, especially in heavy traffic in the major cities. In Malaysia, it is common to see a young or child pillion rider clinging on to an adult at the front of the motorcycle. One of the main issues facing young pillion riders is that their safety is often not taken into account when they are riding on a motorcycle. This article reviews the legally available systems in child safety for underbone motorcycles in Malaysia while putting forth the need for a safety system for child pillion riders. Various databases were searched for underbone motorcycle safety systems, related legislation, motorcycle accident data, and types of injuries and these were reviewed to put forth the need for a new safety system. In motorcycle-related accidents, children usually sustain lower limb injuries, which could temporarily or permanently inhibit the child's movements. Accident statistics in Malaysia, especially those involving motorcycles, reflect a pressing need for a reduction in the number of accidents. In Malaysia, the legislation does not go beyond the mandatory use of safety helmets for young pillion users. There is a pressing need for another safety system or mechanism(s) for young pillion riders of underbone motorcycles. Enforcement of laws to enforce the usage of passive safety systems such as helmets and protective gear is difficult in underdeveloped and developing countries. The intervention of new technology is inevitable. Therefore, this article highlights the need for a new safety backrest system for child pillion riders to ensure their safety.
Safety implications of electronic driving support systems : an orientation.
Gundy, C.M. Steyvers, F.J.J.M. & Kaptein, N.A.
1995-01-01
This report focuses on traffic safety aspects of driving support systems. The report consists of two parts. First of all, the report discusses a number of topics, relevant for the implementation and evaluation of driving support systems. These topics include: (1) safety research into driving support
New Automated System Available for Reporting Safety Concerns | Poster
A new system has been developed for reporting safety issues in the workplace. The Environment, Health, and Safety’s (EHS’) Safety Inspection and Issue Management System (SIIMS) is an online resource where any employee can report a problem or issue, said Siobhan Tierney, program manager at EHS.
Safety and the environment: Key to total quality
International Nuclear Information System (INIS)
Scott, J.N.
1991-01-01
The author briefly discusses the principles that guide the safety and environmental efforts of the petroleum industry. Specific examples are given of things that the industry is doing to carry out those principles. There are 12 principles which say that the petroleum industry will be responsive to the communities in which they operate. In general they say that waste will try to be eliminated; the industry will handle raw materials, products and waste in a safe manner; and that the petroleum industry will work with government to draft regulations that will improve the safety of the workplace and the environment
Diversity requirements for safety critical software-based automation systems
International Nuclear Information System (INIS)
Korhonen, J.; Pulkkinen, U.; Haapanen, P.
1998-03-01
System vendors nowadays propose software-based systems even for the most critical safety functions in nuclear power plants. Due to the nature and mechanisms of influence of software faults new methods are needed for the safety and reliability evaluation of these systems. In the research project 'Programmable automation systems in nuclear power plants (OHA)' various safety assessment methods and tools for software based systems are developed and evaluated. This report first discusses the (common cause) failure mechanisms in software-based systems, then defines fault-tolerant system architectures to avoid common cause failures, then studies the various alternatives to apply diversity and their influence on system reliability. Finally, a method for the assessment of diversity is described. Other recently published reports in OHA-report series handles the statistical reliability assessment of software based (STUK-YTO-TR 119), usage models in reliability assessment of software-based systems (STUK-YTO-TR 128) and handling of programmable automation in plant PSA-studies (STUK-YTO-TR 129)
Electronic clinical safety reporting system: a benefits evaluation.
Elliott, Pamela; Martin, Desmond; Neville, Doreen
2014-06-11
Eastern Health, a large health care organization in Newfoundland and Labrador (NL), started a staged implementation of an electronic occurrence reporting system (used interchangeably with "clinical safety reporting system") in 2008, completing Phase One in 2009. The electronic clinical safety reporting system (CSRS) was designed to replace a paper-based system. The CSRS involves reporting on occurrences such as falls, safety/security issues, medication errors, treatment and procedural mishaps, medical equipment malfunctions, and close calls. The electronic system was purchased from a vendor in the United Kingdom that had implemented the system in the United Kingdom and other places, such as British Columbia. The main objective of the new system was to improve the reporting process with the goal of improving clinical safety. The project was funded jointly by Eastern Health and Canada Health Infoway. The objectives of the evaluation were to: (1) assess the CSRS on achieving its stated objectives (particularly, the benefits realized and lessons learned), and (2) identify contributions, if any, that can be made to the emerging field of electronic clinical safety reporting. The evaluation involved mixed methods, including extensive stakeholder participation, pre/post comparative study design, and triangulation of data where possible. The data were collected from several sources, such as project documentation, occurrence reporting records, stakeholder workshops, surveys, focus groups, and key informant interviews. The findings provided evidence that frontline staff and managers support the CSRS, identifying both benefits and areas for improvement. Many benefits were realized, such as increases in the number of occurrences reported, in occurrences reported within 48 hours, in occurrences reported by staff other than registered nurses, in close calls reported, and improved timelines for notification. There was also user satisfaction with the tool regarding ease of use
International Nuclear Information System (INIS)
Han, Seok Jung; Lim, Ho Gon; Yang, Joon Eon
2003-01-01
To support the development of a Probabilistic Safety Assessment (PSA) model usable in Riskinformed Applications (RIA) for Korea Standard Nuclear power Plants (KSNP), we have performed a thermal hydraulic analysis of Aggressive Secondary Cooldown (ASC) in a 2-inch Small Break Loss Of Coolant Accident (SBLOCA) with a total loss of High Pressure Safety Injection (HPSI). The present study focuses on the estimation of the success criteria of ASC, and the enhanced understanding of the detailed thermal hydraulic behavior and phenomena. The results have shown that the Reactor Coolant System (RCS) pressure can be reduced to the Low Pressure Safety Injection (LPSI) operation conditions without core damage. It was also shown that more relaxed success criteria compared to those in the previous PSA models of KSNP could be used in the new PSA model. However, it was found that the results could be affected by various parameters related with ASC operation, i.e., reference temperature for the calculation of the cooldown rate and its control method
Organizational and methodological aspects for contemporary health and safety management system
Directory of Open Access Journals (Sweden)
Sugak Evgeny
2017-01-01
Full Text Available Industrial injuries and work-related disorders considerable lowering we are facing in developed countries may be due to switching to a new health and safety management system entitled “Occupational Safety and Health Management System”. The Russian Federation has prepared certain regulatory documents prescribing some suggestions regarding implementing the contemporary system for industrial injuries prevention based upon the methods for professional risks management. However, despite the efforts made by the Russian Government, reformation of the health and safety management system at various companies is being performed rather slowly that may be as well owing to poor competence of managers and specialists regarding contemporary labor safety model content, methodical and organizational novations in the sphere of occupational safety and health management.. The article refers to a number of principal issues distinguishing the new health and safety management system from conventional approach.
Workshop on development and view on digital safety system of KNICS
Energy Technology Data Exchange (ETDEWEB)
NONE
2006-05-15
The contents of this workshop are vision of KNICS, introduction of development of safety system of KNICS, development situation of safety class of PLC, view of software for safety-critical system in PLC, RTOS development by shaping, quality assurance and attestation of PLC, development situation of nuclear reactor system and development situation of ESF-CCS.
Workshop on development and view on digital safety system of KNICS
International Nuclear Information System (INIS)
2006-05-01
The contents of this workshop are vision of KNICS, introduction of development of safety system of KNICS, development situation of safety class of PLC, view of software for safety-critical system in PLC, RTOS development by shaping, quality assurance and attestation of PLC, development situation of nuclear reactor system and development situation of ESF-CCS
Risk-based reconfiguration of safety monitoring system using dynamic Bayesian network
International Nuclear Information System (INIS)
Kohda, Takehisa; Cui Weimin
2007-01-01
To prevent an abnormal event from leading to an accident, the role of its safety monitoring system is very important. The safety monitoring system detects symptoms of an abnormal event to mitigate its effect at its early stage. As the operation time passes by, the sensor reliability decreases, which implies that the decision criteria of the safety monitoring system should be modified depending on the sensor reliability as well as the system reliability. This paper presents a framework for the decision criteria (or diagnosis logic) of the safety monitoring system. The logic can be dynamically modified based on sensor output data monitored at regular intervals to minimize the expected loss caused by two types of safety monitoring system failure events: failed-dangerous (FD) and failed-safe (FS). The former corresponds to no response under an abnormal system condition, while the latter implies a spurious activation under a normal system condition. Dynamic Bayesian network theory can be applied to modeling the entire system behavior composed of the system and its safety monitoring system. Using the estimated state probabilities, the optimal decision criterion is given to obtain the optimal diagnosis logic. An illustrative example of a three-sensor system shows the merits and characteristics of the proposed method, where the reasonable interpretation of sensor data can be obtained
International Nuclear Information System (INIS)
Haapanen, P.; Korhonen, J.
1995-01-01
The safety assessment of programmable automation systems can not totally be based on conventional probabilistic methods because of the difficulties in quantification of the reliability of the software as well as the hardware. Additional means shall therefore be used to gain more confidence on the system dependability. One central confidence building measure is the independent dynamic testing of the completed system. An automated test harness is needed to run the required large amount of test cases in a restricted time span. The prototype dynamic testing harness for programmable digital systems developed at the Technical Research Centre of Finland (VTT) is described in the presentation. (12 refs., 2 figs., 2 tabs.)
Safety Metrics for Human-Computer Controlled Systems
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.
Nuclear Reactor RA Safety Report, Vol. 8, Auxiliary system
International Nuclear Information System (INIS)
1986-11-01
This volume describes RA reactor auxiliary systems, as follows: special ventilation system, special drainage system, hot cells, systems for internal transport. Ventilation system is considered as part of the reactor safety and protection system. Its role is eliminate possible radioactive particles dispersion in the environment. Special drainage system includes pipes and reservoirs with the safety role, meaning absorption or storage of possible radioactive waste water from the reactor building. Hot cells existing in the RA reactor building are designed for production of sealed radioactive sources, including packaging and transport [sr
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
Smith, David J
2010-01-01
Electrical, electronic and programmable electronic systems increasingly carry out safety functions to guard workers and the public against injury or death and the environment against pollution. The international functional safety standard IEC 61508 was revised in 2010, and this is the first comprehensive guide available to the revised standard. As functional safety is applicable to many industries, this book will have a wide readership beyond the chemical and process sector, including oil and gas, power generation, nuclear, aircraft, and automotive industries, plus project, instrumentation, design, and control engineers. * The only comprehensive guide to IEC 61508, updated to cover the 2010 amendments, that will ensure engineers are compliant with the latest process safety systems design and operation standards* Helps readers understand the process required to apply safety critical systems standards* Real-world approach helps users to interpret the standard, with case studies and best practice design examples...
Proceedings of the Digital Systems Reliability and Nuclear Safety Workshop
Energy Technology Data Exchange (ETDEWEB)
Wallace, D. R.; Cuthill, B. B.; Ippolito, L. M. [National Inst. of Standards and Technology, Gaithersburg, MD (United States); Beltracchi, L. [Nuclear Regulatory Commission, Washington, DC (United States) ed.
1994-03-01
The United States Nuclear Regulatory Commission (NRC), in cooperation with the National Institute of Standards and Technology conducted the.Digital Systems Reliability and Nuclear Safety Workshop on September 13--14, 1993, in Rockville, Maryland. The workshop provided a forum for the exchange of information among experts within the nuclear industry, experts from other industries, regulators and academia. The information presented at this workshop provided in-depth exposure of the NRC staff and the nuclear industry to digital systems design safety issues and also provided feedback to the NRC from outside experts regarding identified safety issues, proposed regulatory positions, and intended research associated with the use of digital systems in nuclear power plants. Technical presentations provided insights on areas where current software engineering practices may be inadequate for safety-critical systems, on potential solutions for development issues, and on methods for reducing risk in safety-critical systems. This report contains an analysis of results of the workshop, the papers presented panel presentations, and summaries of, discussions at this workshop. The individual papers have been cataloged separately.
Patient Safety Learning Systems: A Systematic Review and Qualitative Synthesis.
2017-01-01
A patient safety learning system (sometimes called a critical incident reporting system) refers to structured reporting, collation, and analysis of critical incidents. To inform a provincial working group's recommendations for an Ontario Patient Safety Event Learning System, a systematic review was undertaken to determine design features that would optimize its adoption into the health care system and would inform implementation strategies. The objective of this review was to address two research questions: (a) what are the barriers to and facilitators of successful adoption of a patient safety learning system reported by health professionals and (b) what design components maximize successful adoption and implementation? To answer the first question, we used a published systematic review. To answer the second question, we used scoping study methodology. Common barriers reported in the literature by health care professionals included fear of blame, legal penalties, the perception that incident reporting does not improve patient safety, lack of organizational support, inadequate feedback, lack of knowledge about incident reporting systems, and lack of understanding about what constitutes an error. Common facilitators included a non-accusatory environment, the perception that incident reporting improves safety, clarification of the route of reporting and of how the system uses reports, enhanced feedback, role models (such as managers) using and promoting reporting, legislated protection of those who report, ability to report anonymously, education and training opportunities, and clear guidelines on what to report. Components of a patient safety learning system that increased successful adoption and implementation were emphasis on a blame-free culture that encourages reporting and learning, clear guidelines on how and what to report, making sure the system is user-friendly, organizational development support for data analysis to generate meaningful learning outcomes
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
Licensing process for safety-critical software-based systems
Energy Technology Data Exchange (ETDEWEB)
Haapanen, P. [VTT Automation, Espoo (Finland); Korhonen, J. [VTT Electronics, Espoo (Finland); Pulkkinen, U. [VTT Automation, Espoo (Finland)
2000-12-01
System vendors nowadays propose software-based technology even for the most critical safety functions in nuclear power plants. Due to the nature of software faults and the way they cause system failures new methods are needed for the safety and reliability evaluation of these systems. In the research project 'Programmable automation systems in nuclear power plants (OHA)', financed together by the Radiation and Nuclear Safety Authority (STUK), the Ministry of Trade and Industry (KTM) and the Technical Research Centre of Finland (VTT), various safety assessment methods and tools for software based systems are developed and evaluated. As a part of the OHA-work a reference model for the licensing process for software-based safety automation systems is defined. The licensing process is defined as the set of interrelated activities whose purpose is to produce and assess evidence concerning the safety and reliability of the system/application to be licensed and to make the decision about the granting the construction and operation permissions based on this evidence. The parties of the licensing process are the authority, the licensee (the utility company), system vendors and their subcontractors and possible external independent assessors. The responsibility about the production of the evidence in first place lies at the licensee who in most cases rests heavily on the vendor expertise. The evaluation and gauging of the evidence is carried out by the authority (possibly using external experts), who also can acquire additional evidence by using their own (independent) methods and tools. Central issue in the licensing process is to combine the quality evidence about the system development process with the information acquired through tests, analyses and operational experience. The purpose of the licensing process described in this report is to act as a reference model both for the authority and the licensee when planning the licensing of individual applications
Licensing process for safety-critical software-based systems
International Nuclear Information System (INIS)
Haapanen, P.; Korhonen, J.; Pulkkinen, U.
2000-12-01
System vendors nowadays propose software-based technology even for the most critical safety functions in nuclear power plants. Due to the nature of software faults and the way they cause system failures new methods are needed for the safety and reliability evaluation of these systems. In the research project 'Programmable automation systems in nuclear power plants (OHA)', financed together by the Radiation and Nuclear Safety Authority (STUK), the Ministry of Trade and Industry (KTM) and the Technical Research Centre of Finland (VTT), various safety assessment methods and tools for software based systems are developed and evaluated. As a part of the OHA-work a reference model for the licensing process for software-based safety automation systems is defined. The licensing process is defined as the set of interrelated activities whose purpose is to produce and assess evidence concerning the safety and reliability of the system/application to be licensed and to make the decision about the granting the construction and operation permissions based on this evidence. The parties of the licensing process are the authority, the licensee (the utility company), system vendors and their subcontractors and possible external independent assessors. The responsibility about the production of the evidence in first place lies at the licensee who in most cases rests heavily on the vendor expertise. The evaluation and gauging of the evidence is carried out by the authority (possibly using external experts), who also can acquire additional evidence by using their own (independent) methods and tools. Central issue in the licensing process is to combine the quality evidence about the system development process with the information acquired through tests, analyses and operational experience. The purpose of the licensing process described in this report is to act as a reference model both for the authority and the licensee when planning the licensing of individual applications. Many of the
Nuclear power plants. Electrical equipment of the safety system. Qualification
International Nuclear Information System (INIS)
2001-01-01
This International Standard applies to electrical parts of safety systems employed at nuclear power plants, including components and equipment of any interface whose failure could affect unfavourably properties of the safety system. The standard also applies to non-electrical safety-related interfaces. Furthermore, the standard describes the generic process of qualification certification procedures and methods of qualification testing and related documentation. (P.A.)
Reliability estimation of safety-critical software-based systems using Bayesian networks
International Nuclear Information System (INIS)
Helminen, A.
2001-06-01
Due to the nature of software faults and the way they cause system failures new methods are needed for the safety and reliability evaluation of software-based safety-critical automation systems in nuclear power plants. In the research project 'Programmable automation system safety integrity assessment (PASSI)', belonging to the Finnish Nuclear Safety Research Programme (FINNUS, 1999-2002), various safety assessment methods and tools for software based systems are developed and evaluated. The project is financed together by the Radiation and Nuclear Safety Authority (STUK), the Ministry of Trade and Industry (KTM) and the Technical Research Centre of Finland (VTT). In this report the applicability of Bayesian networks to the reliability estimation of software-based systems is studied. The applicability is evaluated by building Bayesian network models for the systems of interest and performing simulations for these models. In the simulations hypothetical evidence is used for defining the parameter relations and for determining the ability to compensate disparate evidence in the models. Based on the experiences from modelling and simulations we are able to conclude that Bayesian networks provide a good method for the reliability estimation of software-based systems. (orig.)
Safety of the medical gas pipeline system
Directory of Open Access Journals (Sweden)
Sushmita Sarangi
2018-01-01
Full Text Available Medical gases are nowadays being used for a number of diverse clinical applications and its piped delivery is a landmark achievement in the field of patient care. Patient safety is of paramount importance in the design, installation, commissioning, and operation of medical gas pipeline systems (MGPS. The system has to be operational round the clock, with practically zero downtime and its failure can be fatal if not restored at the earliest. There is a lack of awareness among the clinicians regarding the medico-legal aspect involved with the MGPS. It is a highly technical field; hence, an in-depth knowledge is a must to ensure safety with the system.
Optimization of maintenance periodicity of complex of NPP safety systems
International Nuclear Information System (INIS)
Kolykhanov, V.; Skalozubov, V.; Kovrigkin, Y.
2006-01-01
The analysis of the positive and negative aspects connected to maintenance of the safety systems equipment which basically is in a standby state is executed. Tests of systems provide elimination of the latent failures and raise their reliability. Poor quality of carrying out the tests can be a source of the subsequent failures. Therefore excess frequency of tests can result in reducing reliability of safety systems. The method of optimization of maintenance periodicity of the equipment taking into account factors of its reliability and restoration procedures quality is submitted. The unavailability factor is used as a criterion of optimization of maintenance periodicity. It is offered to use parameters of reliability of the equipment and each of safety systems of NPPs received at developing PSA. And it is offered to carry out the concordance of maintenance periodicity of systems within the NPP maintenance program taking into account a significance factor of the system received on the basis of the contribution of system in CDF. Basing on the submitted method the small computer code is developed. This code allows to calculate reliability factors of a separate safety system and to determine optimum maintenance periodicity of its equipment. Optimization of maintenance periodicity of a complex of safety systems is stipulated also. As an example results of optimization of maintenance periodicity at Zaporizhzhya NPP are presented. (author)
Modeling for safety in a synthesis-centric systems engineering framework
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
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)
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
The NASA Aviation Safety Reporting System
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.
Regulatory system reform of occupational health and safety in China.
Wu, Fenghong; Chi, Yan
2015-01-01
With the explosive economic growth and social development, China's regulatory system of occupational health and safety now faces more and more challenges. This article reviews the history of regulatory system of occupational health and safety in China, as well as the current reform of this regulatory system in the country. Comprehensive, a range of laws, regulations and standards that promulgated by Chinese government, duties and responsibilities of the regulatory departments are described. Problems of current regulatory system, the ongoing adjustments and changes for modifying and improving regulatory system are discussed. The aim of reform and the incentives to drive forward more health and safety conditions in workplaces are also outlined.
Human-system safety methods for development of advanced air traffic management systems
International Nuclear Information System (INIS)
Nelson, William R.
1999-01-01
The Idaho National Engineering and Environmental Laboratory (INEEL) is supporting the National Aeronautics and Space Administration in the development of advanced air traffic management (ATM) systems as part of the Advanced Air Transportation Technologies program. As part of this program INEEL conducted a survey of human-system safety methods that have been applied to complex technical systems, to identify lessons learned from these applications and provide recommendations for the development of advanced ATM systems. The domains that were surveyed included offshore oil and gas, commercial nuclear power, commercial aviation, and military. The survey showed that widely different approaches are used in these industries, and that the methods used range from very high-level, qualitative approaches to very detailed quantitative methods such as human reliability analysis (HRA) and probabilistic safety assessment (PSA). In addition, the industries varied widely in how effectively they incorporate human-system safety assessment in the design, development, and testing of complex technical systems. In spite of the lack of uniformity in the approaches and methods used, it was found that methods are available that can be combined and adapted to support the development of advanced air traffic management systems (author) (ml)
Can cyclist safety be improved with intelligent transport systems?
Silla, Anne; Leden, Lars; Rämä, Pirkko; Scholliers, Johan; Van Noort, Martijn; Bell, Daniel
2017-08-01
In recent years, Intelligent Transport Systems (ITS) have assisted in the decrease of road traffic fatalities, particularly amongst passenger car occupants. Vulnerable Road Users (VRUs) such as pedestrians, cyclists, moped riders and motorcyclists, however, have not been that much in focus when developing ITS. Therefore, there is a clear need for ITS which specifically address VRUs as an integrated element of the traffic system. This paper presents the results of a quantitative safety impact assessment of five systems that were estimated to have high potential to improve the safety of cyclists, namely: Blind Spot Detection (BSD), Bicycle to Vehicle communication (B2V), Intersection safety (INS), Pedestrian and Cyclist Detection System+Emergency Braking (PCDS+EBR) and VRU Beacon System (VBS). An ex-ante assessment method proposed by Kulmala (2010) targeted to assess the effects of ITS for cars was applied and further developed in this study to assess the safety impacts of ITS specifically designed for VRUs. The main results of the assessment showed that all investigated systems affect cyclist safety in a positive way by preventing fatalities and injuries. The estimates considering 2012 accident data and full penetration showed that the highest effects could be obtained by the implementation of PCDS+EBR and B2V, whereas VBS had the lowest effect. The estimated yearly reduction in cyclist fatalities in the EU-28 varied between 77 and 286 per system. A forecast for 2030, taking into accounts the estimated accident trends and penetration rates, showed the highest effects for PCDS+EBR and BSD. Copyright © 2016 Elsevier Ltd. All rights reserved.
Kushniruk, Andre; Senathirajah, Yalini; Borycki, Elizabeth
2017-01-01
The usability and safety of health information systems have become major issues in the design and implementation of useful healthcare IT. In this paper we describe a multi-phased multi-method approach to integrating usability engineering methods into system testing to ensure both usability and safety of healthcare IT upon widespread deployment. The approach involves usability testing followed by clinical simulation (conducted in-situ) and "near-live" recording of user interactions with systems. At key stages in this process, usability problems are identified and rectified forming a usability and technology-induced error "safety net" that catches different types of usability and safety problems prior to releasing systems widely in healthcare settings.
The impact of using an intravenous workflow management system (IVWMS) on cost and patient safety.
Lin, Alex C; Deng, Yihong; Thaibah, Hilal; Hingl, John; Penm, Jonathan; Ivey, Marianne F; Thomas, Mark
2018-07-01
The aim of this study was to determine the financial costs associated with wasted and missing doses before and after the implementation of an intravenous workflow management system (IVWMS) and to quantify the number and the rate of detected intravenous (IV) preparation errors. A retrospective analysis of the sample hospital information system database was conducted using three months of data before and after the implementation of an IVWMS System (DoseEdge ® ) which uses barcode scanning and photographic technologies to track and verify each step of the preparation process. The financial impact associated with wasted and missing >IV doses was determined by combining drug acquisition, labor, accessory, and disposal costs. The intercepted error reports and pharmacist detected error reports were drawn from the IVWMS to quantify the number of errors by defined error categories. The total number of IV doses prepared before and after the implementation of the IVWMS system were 110,963 and 101,765 doses, respectively. The adoption of the IVWMS significantly reduced the amount of wasted and missing IV doses by 14,176 and 2268 doses, respectively (p system was $144,019 over 3 months. The total number of errors detected was 1160 (1.14%) after using the IVWMS. The implementation of the IVWMS facilitated workflow changes that led to a positive impact on cost and patient safety. The implementation of the IVWMS increased patient safety by enforcing standard operating procedures and bar code verifications. Published by Elsevier B.V.
ICT support safety, health and environment management system (e-SHEMS)
International Nuclear Information System (INIS)
Amy Hamijah Ab Hamid; Hasfazilah Hassan; Siti Massari Amran; Norzalina Nasirudin; Azimawati Ahmad; Mohd Suhaimi Kassim; Shaharum Ramli; Musa Ibrahim; Mohd Sidek Othman
2009-01-01
Safety program is compulsory for a nuclear technology related research and development institution like Nuclear Malaysia. It has been implemented in various safety standard systems including Act 514, Act 304, ISO 14000, OSHAS 18001 and IAEA. This paper began with Nuclear Malaysia history in initiating our own safety standard system since 1982. Currently, Nuclear Malaysia's Safety Health and Environment Management System (SHE-MS) was stipulated for similar purpose. Furthermore, it has implemented guidelines by AELB, IAEA, DOSH, Fire Brigade and Police Force. This paper briefly describes the overall structure of SHE-MS, how it functions and being managed, and lessons learned. The findings which are based on the issues and challenges, then it can be analysed to propose a development of SHE-MS ICT-support application for future improvement and enhancement in inculcating and nurturing safety culture among Nuclear Malaysia staff. (Author)
EC-sponsored research activities on innovative passive safety systems
International Nuclear Information System (INIS)
Bermejo, J.M.; Goethem, G. van
2000-01-01
On April 26th 1994, the European Union (EU) adopted via a Council Decision a EURATOM Multiannual Programme for community activities in the field of Nuclear Fission Safety (NFS) Research for the period 1994 to 1998. An area of work having, as an objective, to 'explore innovative approaches' to improve the safety of future and existing reactors, was introduced in this programme. Most of the projects selected in this area, which have been grouped under a common cluster known as 'INNO', are currently being carried out on a 'cost-shared' basis, i.e. contribution of the European Commission is up to 50% of the total cost. At present, the 'INNO' cluster is composed of 10 projects in which 25 different organisations, representing research centres, universities, regulators, utilities and vendors from 7 EU member states and Switzerland, are involved. These projects are proving to be an efficient means to gain the necessary phenomenological knowledge and to solve the challenging problems, many times of generic nature, posed among others by the characteristically small driving forces of the systems studied and by the lack of really prototypical test facilities. (author)
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…
Establishment of Safety Analysis System and Technology for CANDU Reactors
International Nuclear Information System (INIS)
Min, Byung Joo; Kim, W. Y.; Kim, H. T.; Rhee, B. W.; Yoon, C.; Kang, H. S.; Yoo, K. J.
2005-03-01
To improve the CANDU design/operation safety analysis codes and the CANDU safety analysis methodology, the following works have been done. From the development of the lattice codes (WIMS/CANDU), the lattice model simulates the real core lattice geometry and the effect of the pressure tube creep to the core lattice parameter has been evaluated. From the development of the 3-dimensional thermal-hydraulic analysis model of the moderator behavior (CFX4-CAMO), validation of the model against STERN Lab experiment has been executed. The butterfly-shaped grid structure and the 3-dimensional flow resistance model for porous media were developed and applied to the moderator analysis for Wolsong units 2/3/4. The single fuel channel analysis codes for blowdown and post-blowdown were unified by CATHENA. The 3-dimensional fuel channel analysis model (CFX-CACH) has been developed for validation of CATHENA fuel channel analysis model. The interlinking analysis system (CANVAS) of the thermal-hydraulic safety analysis codes for the primary heat transport system and containment system has been executed. The database system of core physics and thermal-hydraulics experimental data for safety analysis has been established on the URL: http://CANTHIS.kaeri.re.kr. For documentation and Standardization of the general safety analysis procedure, the general safety analysis procedure is developed and applied to a large break LOCA. The present research results can be utilized for establishment of the independent safety analysis technology and acquisition of the optimal safety analysis technology
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)
Development of safety review advisory system for nuclear power plants
International Nuclear Information System (INIS)
Kim, M. W.; Lee, H. C.; Park, S. O.; Park, W. J.; Lee, J. I.; Hur, K. Y.; Choi, S. S.; Lee, S. J.; Kang, C. M.
2001-01-01
For the development of an expert system supporting the safety review of nuclear power plants, the application program was implemented after gathering necessary theoretical background and practical requirements. The general and the detail functional specifications were established, and they were investigated by the safety review experts at KINS. Safety Review Advisory System (SRAS), the windows application on client-server environment was developed according to the above specifications. Reviewers can do their safety reviewing regardless of speciality or reviewing experiences because SRAS is operated by the safety review plans which are converted to standardized format. When the safety reviewing is carried out by using SRAS, the results of safety reviewing are accumulated in the database and may be utilized later usefully, and we can grasp safety reviewing progress. Users of SRAS are categorized into three groups, administrator, project manager, and reviewer. Each user group has appropriate access capability. The function and some screen shots of SRAS are described in this paper
Decision support systems and expert systems for risk and safety analysis
International Nuclear Information System (INIS)
Baybutt, P.
1986-01-01
During the last 1-2 years, rapid developments have occurred in the development of decision support systems and expert systems to aid in decision making related to risk and safety of industrial plants. These activities are most noteworthy in the nuclear industry where numerous systems are under development with implementation often being made on personal computers. An overview of some of these developments is provided, and an example of one recently developed decision support system is given. This example deals with CADET, a system developed to aid the U.S. Nuclear Regulatory Commission in making decisions related to the topical issue of source terms resulting from degraded core accidents in light water reactors. The paper concludes with some comments on the likely directions of future developments in decision support systems and expert systems to aid in the management of risk and safety in industrial plants. (author)
NRC confirmatory safety system testing in support of AP600 design review
International Nuclear Information System (INIS)
Rhee, G.S.; Bessette, D.E.; Shotkin, L.M.
1994-01-01
Westinghouse Electric Corporation has submitted the Advanced Passive 600 MWe (AP600) nuclear power plant design to the NRC for design certification. The Office of Nuclear Regulatory Research is proceeding to conduct confirmatory testing to help the NRC staff evaluate the AP600 safety system design. For confirmatory testing, it was determined that the cost-effective route was to modify an existing full-height, full-pressure test facility rather than build a new one. Thus, all the existing integral effects test facilities, both in the US and abroad, were screened to select the best candidate. As a result, the ROSA-V (Rig of Safety Assessment-V) test facility located in the Japan Atomic Energy Research Institute (JAERI) was chosen. However, because of some differences in design between the existing ROSA-V facility and the AP600, the ROSA-V is being modified to conform to the AP600 safety system design. The modification work will be completed by the end of this year. A series of facility characterization tests will then be performed in January 1994 for the modified part of the facility before the main test series is initiated in February 1994. A total of 12 tests will be performed in 1994 under Phase I of this cooperative program with JAERI. Phase II testing is being considered to be conducted in 1995 mainly for beyond-design-basis accident evaluation
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.
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
New reactor technology: safety improvements in nuclear power systems.
Corradini, M L
2007-11-01
Almost 450 nuclear power plants are currently operating throughout the world and supplying about 17% of the world's electricity. These plants perform safely, reliably, and have no free-release of byproducts to the environment. Given the current rate of growth in electricity demand and the ever growing concerns for the environment, nuclear power can only satisfy the need for electricity and other energy-intensive products if it can demonstrate (1) enhanced safety and system reliability, (2) minimal environmental impact via sustainable system designs, and (3) competitive economics. The U.S. Department of Energy with the international community has begun research on the next generation of nuclear energy systems that can be made available to the market by 2030 or earlier, and that can offer significant advances toward these challenging goals; in particular, six candidate reactor system designs have been identified. These future nuclear power systems will require advances in materials, reactor physics, as well as thermal-hydraulics to realize their full potential. However, all of these designs must demonstrate enhanced safety above and beyond current light water reactor systems if the next generation of nuclear power plants is to grow in number far beyond the current population. This paper reviews the advanced Generation-IV reactor systems and the key safety phenomena that must be considered to guarantee that enhanced safety can be assured in future nuclear reactor systems.
Advanced Range Safety System for High Energy Vehicles
Claxton, Jeffrey S.; Linton, Donald F.
2002-01-01
The advanced range safety system project is a collaboration between the National Aeronautics and Space Administration and the United States Air Force to develop systems that would reduce costs and schedule for safety approval for new classes of unmanned high-energy vehicles. The mission-planning feature for this system would yield flight profiles that satisfy the mission requirements for the user while providing an increased quality of risk assessment, enhancing public safety. By improving the speed and accuracy of predicting risks to the public, mission planners would be able to expand flight envelopes significantly. Once in place, this system is expected to offer the flexibility of handling real-time risk management for the high-energy capabilities of hypersonic vehicles including autonomous return-from-orbit vehicles and extended flight profiles over land. Users of this system would include mission planners of Space Launch Initiative vehicles, space planes, and other high-energy vehicles. The real-time features of the system could make extended flight of a malfunctioning vehicle possible, in lieu of an immediate terminate decision. With this improved capability, the user would have more time for anomaly resolution and potential recovery of a malfunctioning vehicle.
Safety parameter display system (SPDS) for Russian-designed NPPs
International Nuclear Information System (INIS)
Anikanov, S.S.; Catullo, W.J.; Pelusi, J.L.
1997-01-01
As part of the programs aimed at improving the safety of Russian-designed reactors, the US DoE has sponsored a project of providing a safety parameter display system (SPDS) for nuclear power plants with such reactors. The present paper is focused mostly on the system architecture design features of SPDS systems for WWER-1000 and RBMK-1000 reactors. The function and the operating modes of the SPDS are outlined, and a description of the display system is given. The system architecture and system design of both an integrated and a stand-alone IandC system is explained. (A.K.)
Nuclear power plant systems, structures and components and their safety classification
International Nuclear Information System (INIS)
2000-01-01
The assurance of a nuclear power plant's safety is based on the reliable functioning of the plant as well as on its appropriate maintenance and operation. To ensure the reliability of operation, special attention shall be paid to the design, manufacturing, commissioning and operation of the plant and its components. To control these functions the nuclear power plant is divided into structural and functional entities, i.e. systems. A systems safety class is determined by its safety significance. Safety class specifies the procedures to be employed in plant design, construction, monitoring and operation. The classification document contains all documentation related to the classification of the nuclear power plant. The principles of safety classification and the procedures pertaining to the classification document are presented in this guide. In the Appendix of the guide, examples of systems most typical of each safety class are given to clarify the safety classification principles
Safety assessment of envisaged systems for automotive hydrogen supply and utilization
Energy Technology Data Exchange (ETDEWEB)
Landucci, Gabriele [Dipartimento di Ingegneria Chimica, Chimica Industriale e Scienza dei Materiali, Universita di Pisa, via Diotisalvi n.2, 56126 Pisa (Italy); Tugnoli, Alessandro; Cozzani, Valerio [Dipartimento di Ingegneria Chimica, Mineraria e delle Tecnologie Ambientali, Alma Mater Studiorum - Universita di Bologna, via Terracini n.28, 40131 Bologna (Italy)
2010-02-15
A novel consequence-based approach was applied to the inherent safety assessment of the envisaged hydrogen production, distribution and utilization systems, in the perspective of the widespread hydrogen utilization as a vehicle fuel. Alternative scenarios were assessed for the hydrogen system chain from large scale production to final utilization. Hydrogen transportation and delivery was included in the analysis. The inherent safety fingerprint of each system was quantified by a set of Key Performance Indicators (KPIs). Rules for KPIs aggregation were considered for the overall assessment of the system chains. The final utilization stage resulted by large the more important for the overall expected safety performance of the system. Thus, comparison was carried out with technologies proposed for the use of other low emission fuels, as LPG and natural gas. The hazards of compressed hydrogen-fueled vehicles resulted comparable, while reference innovative hydrogen technologies evidenced a potentially higher safety performance. Thus, switching to the inherently safer technologies currently under development may play an important role in the safety enhancement of hydrogen vehicles, resulting in a relevant improvement of the overall safety performance of the entire hydrogen system. (author)
System and safety studies of accelerator driven systems for transmutation. Annual report 2007
International Nuclear Information System (INIS)
Arzhanov, Vasily; Fokau, Andrei; Persson, Calle; Runevall, Odd; Sandberg, Nils; Tesinsky, Milan; Wallenius, Janne; Youpeng Zhang
2008-05-01
Within the project 'System and safety studies of accelerator driven systems for transmutation', research on design and safety of sub-critical reactors for recycling of minor actinides is performed. During 2007, the reactor physics division at KTH has calculated safety parameters for EFIT-400 with cermet fuel, permitting to start the transient safety analysis. The accuracy of different reactivity meters applied to the YALINA facility was assessed and neutron detection studies were performed. A model to address deviations from point kinetic behaviour was developed. Studies of basic radiation damage physics included calculations of vacancy formation and activation enthalpies in bcc niobium. In order to predict the oxygen potential of inert matrix fuels, a thermo-chemical model for mixed actinide oxides was implemented in a phase equilibrium code
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)
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
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
Yorio, Patrick L; Willmer, Dana R; Haight, Joel M
2014-08-01
Since the late 1980s, the U.S. Department of Labor has considered regulating a systems approach to occupational health and safety management. Recently, a health and safety management systems (HSMS) standard has returned to the regulatory agenda of both the Occupational Safety and Health Administration (OSHA) and the Mine Safety and Health Administration (MSHA). Because a mandated standard has implications for both industry and regulating bodies alike, it is imperative to gain a greater understanding of the potential effects that an HSMS regulatory approach can have on establishment-level injuries and illnesses. Through the lens of MSHA's regulatory framework, we first explore how current enforcement activities align with HSMS elements. Using MSHA data for the years 2003-2010, we then analyze the relationship between various types of enforcement activities (e.g., total number of citations, total penalty amount, and HSMS-aligned citations) and mine reportable injuries. Our findings show that the reduction in mine reportable injuries predicted by increases in MSHA enforcement ranges from negligible to 18%. The results suggest that the type and focus of the enforcement activity may be more important for accident reduction than the total number of citations issued and the associated penalty amount. © 2014 Society for Risk Analysis.
Evaluation of food safety management systems in Serbian dairy industry
Directory of Open Access Journals (Sweden)
Igor Tomašević
2016-01-01
Full Text Available This paper reports incentives, costs, difficulties and benefits of food safety management systems implementation in the Serbian dairy industry. The survey involved 27 food business operators with the national milk and dairy market share of 65 %. Almost two thirds of the assessed dairy producers (70.4 % claimed that they had a fully operational and certified HACCP system in place, while 29.6 % implemented HACCP, but had no third party certification. ISO 22000 was implemented and certified in 29.6 % of the companies, while only 11.1 % had implemented and certified IFS standard. The most important incentive for implementing food safety management systems for Serbian dairy producers was to increase and improve safety and quality of dairy products. The cost of product investigation/analysis and hiring external consultants were related to the initial set-up of food safety management system with the greatest importance. Serbian dairy industry was not greatly concerned by the financial side of implementing food safety management systems due to the fact that majority of prerequisite programmes were in place and regularly used by almost 100 % of the producers surveyed. The presence of competency gap between the generic knowledge for manufacturing food products and the knowledge necessary to develop and implement food safety management systems was confirmed, despite the fact that 58.8 % of Serbian dairy managers had university level of education. Our study brings about the innovation emphasizing the attitudes and the motivation of the food production staff as the most important barrier for the development and implementation of HACCP. The most important identified benefit was increased safety of dairy products with the mean rank scores of 6.85. The increased customer confidence and working discipline of staff employed in food processing were also found as important benefits of implementing/operating HACCP. The study shows that the level of HACCP
Impact of Passive Safety on FHR Instrumentation Systems Design and Classification
International Nuclear Information System (INIS)
Holcomb, David Eugene
2015-01-01
Fluoride salt-cooled high-temperature reactors (FHRs) will rely more extensively on passive safety than earlier reactor classes. 10CFR50 Appendix A, General Design Criteria for Nuclear Power Plants, establishes minimum design requirements to provide reasonable assurance of adequate safety. 10CFR50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors, provides guidance on how the safety significance of systems, structures, and components (SSCs) should be reflected in their regulatory treatment. The Nuclear Energy Institute (NEI) has provided 10 CFR 50.69 SSC Categorization Guideline (NEI-00-04) that factors in probabilistic risk assessment (PRA) model insights, as well as deterministic insights, through an integrated decision-making panel. Employing the PRA to inform deterministic requirements enables an appropriately balanced, technically sound categorization to be established. No FHR currently has an adequate PRA or set of design basis accidents to enable establishing the safety classification of its SSCs. While all SSCs used to comply with the general design criteria (GDCs) will be safety related, the intent is to limit the instrumentation risk significance through effective design and reliance on inherent passive safety characteristics. For example, FHRs have no safety-significant temperature threshold phenomena, thus enabling the primary and reserve reactivity control systems required by GDC 26 to be passively, thermally triggered at temperatures well below those for which core or primary coolant boundary damage would occur. Moreover, the passive thermal triggering of the primary and reserve shutdown systems may relegate the control rod drive motors to the control system, substantially decreasing the amount of safety-significant wiring needed. Similarly, FHR decay heat removal systems are intended to be running continuously to minimize the amount of safety-significant instrumentation needed to initiate
The complexity of patient safety reporting systems in UK dentistry.
Renton, T; Master, S
2016-10-21
Since the 'Francis Report', UK regulation focusing on patient safety has significantly changed. Healthcare workers are increasingly involved in NHS England patient safety initiatives aimed at improving reporting and learning from patient safety incidents (PSIs). Unfortunately, dentistry remains 'isolated' from these main events and continues to have a poor record for reporting and learning from PSIs and other events, thus limiting improvement of patient safety in dentistry. The reasons for this situation are complex.This paper provides a review of the complexities of the existing systems and procedures in relation to patient safety in dentistry. It highlights the conflicting advice which is available and which further complicates an overly burdensome process. Recommendations are made to address these problems with systems and procedures supporting patient safety development in dentistry.
21 CFR 862.1110 - Bilirubin (total or direct) test system.
2010-04-01
... 21 Food and Drugs 8 2010-04-01 2010-04-01 false Bilirubin (total or direct) test system. 862.1110... Systems § 862.1110 Bilirubin (total or direct) test system. (a) Identification. A bilirubin (total or direct) test system is a device intended to measure the levels of bilirubin (total or direct) in plasma...
System analysis of vehicle active safety problem
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.
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
Information systems in food safety management.
McMeekin, T A; Baranyi, J; Bowman, J; Dalgaard, P; Kirk, M; Ross, T; Schmid, S; Zwietering, M H
2006-12-01
Information systems are concerned with data capture, storage, analysis and retrieval. In the context of food safety management they are vital to assist decision making in a short time frame, potentially allowing decisions to be made and practices to be actioned in real time. Databases with information on microorganisms pertinent to the identification of foodborne pathogens, response of microbial populations to the environment and characteristics of foods and processing conditions are the cornerstone of food safety management systems. Such databases find application in: Identifying pathogens in food at the genus or species level using applied systematics in automated ways. Identifying pathogens below the species level by molecular subtyping, an approach successfully applied in epidemiological investigations of foodborne disease and the basis for national surveillance programs. Predictive modelling software, such as the Pathogen Modeling Program and Growth Predictor (that took over the main functions of Food Micromodel) the raw data of which were combined as the genesis of an international web based searchable database (ComBase). Expert systems combining databases on microbial characteristics, food composition and processing information with the resulting "pattern match" indicating problems that may arise from changes in product formulation or processing conditions. Computer software packages to aid the practical application of HACCP and risk assessment and decision trees to bring logical sequences to establishing and modifying food safety management practices. In addition there are many other uses of information systems that benefit food safety more globally, including: Rapid dissemination of information on foodborne disease outbreaks via websites or list servers carrying commentary from many sources, including the press and interest groups, on the reasons for and consequences of foodborne disease incidents. Active surveillance networks allowing rapid dissemination
Regulatory system reform of occupational health and safety in China
WU, Fenghong; CHI, Yan
2015-01-01
With the explosive economic growth and social development, China’s regulatory system of occupational health and safety now faces more and more challenges. This article reviews the history of regulatory system of occupational health and safety in China, as well as the current reform of this regulatory system in the country. Comprehensive, a range of laws, regulations and standards that promulgated by Chinese government, duties and responsibilities of the regulatory departments are described. Problems of current regulatory system, the ongoing adjustments and changes for modifying and improving regulatory system are discussed. The aim of reform and the incentives to drive forward more health and safety conditions in workplaces are also outlined. PMID:25843565
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...
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...
System safety program plan for the Isotope Brayton Ground Demonstration System (phase I)
International Nuclear Information System (INIS)
1976-01-01
The safety engineering effort to be undertaken in achieving an acceptable level of safety in the Brayton Isotope Power System (BIPS) development program is discussed. The safety organizational relationships, the methods to be used, the tasks to be completed, and the documentation to be published are described. The plan will be updated periodically as the need arises
Failure and factors of safety in piping system design
International Nuclear Information System (INIS)
Antaki, G.A.
1993-01-01
An important body of test and performance data on the behavior of piping systems has led to an ongoing reassessment of the code stress allowables and their safety margin. The codes stress allowables, and their factors of safety, are developed from limits on the incipient yield (for ductile materials), or incipient rupture (for brittle materials), of a test specimen loaded in simple tension. In this paper, we examine the failure theories introduced in the B31 and ASME III codes for piping and their inherent approximations compared to textbook failure theories. We summarize the evolution of factors of safety in ASME and B31 and point out that, for piping systems, it is appropriate to reconsider the concept and definition of factors of safety
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)
33 CFR 96.230 - What objectives must a safety management system meet?
2010-07-01
... management system meet? 96.230 Section 96.230 Navigation and Navigable Waters COAST GUARD, DEPARTMENT OF HOMELAND SECURITY VESSEL OPERATING REGULATIONS RULES FOR THE SAFE OPERATION OF VESSELS AND SAFETY MANAGEMENT SYSTEMS Company and Vessel Safety Management Systems § 96.230 What objectives must a safety...
Engineering systems reliability, safety, and maintenance an integrated approach
Dhillon, B S
2017-01-01
Today, engineering systems are an important element of the world economy and each year billions of dollars are spent to develop, manufacture, operate, and maintain various types of engineering systems around the globe. Many of these systems are highly sophisticated and contain millions of parts. For example, a Boeing jumbo 747 is made up of approximately 4.5 million parts including fasteners. Needless to say, reliability, safety, and maintenance of systems such as this have become more important than ever before. Global competition and other factors are forcing manufacturers to produce highly reliable, safe, and maintainable engineering products. Therefore, there is a definite need for the reliability, safety, and maintenance professionals to work closely during design and other phases. Engineering Systems Reliability, Safety, and Maintenance: An Integrated Approach eliminates the need to consult many different and diverse sources in the hunt for the information required to design better engineering syste...
Developing a fuzzy rule based cognitive map for total system safety assessment
International Nuclear Information System (INIS)
Lemos, Francisco Luiz de; Sullivan, Terry
2007-01-01
Total System Performance Assessment, TSPA, for radioactive waste disposal is a multi and interdisciplinary task that is characterized by complex interactions between parameters and processes; lack of data; and ignorance regarding natural processes and conditions. The vagueness in the determination of ranges of values of parameters and identification of interacting processes pose further difficulties to the analysts with regard to the establishment of the relations between processes and parameters. More specifically the vagueness makes uncertainty propagation and sensitivity analysis challenging to analyze. To cope with these difficulties experts often use simplifications and linguistic terms to express their state of knowledge about a certain situation. For example, experts use terms such as 'low pH', 'very unlikely', etc to describe their perception about natural processes or conditions. In this work we propose the use of Fuzzy Cognitive Maps, FCM, for representation of interrelation between processes and parameters as well as to promote a better understanding of the system performance. Fuzzy cognitive maps are suited for the case where the causal relations are not clearly defined and, therefore, can not be represented by crisp values. In other words, instead of representing the quality of the interactions by crisp values, they are assigned degrees of truth. For example, we can assign values to the effect of one process on another such that (+) 1 corresponds to positive, (-) 1 to negative and 0 to neutral effects respectively. In this case the effect of a process A, on a process, B, can be depicted as function of the membership to the fuzzy set 'causal effect' of the cause process to the target one. One of the main advantages of this methodology would be that it allows one to aggregate the linguistic expressions as descriptions of processes. For example, a process can be known to have a 'very strong' positive effect on another one, or using fuzzy sets terminology
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
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
76 FR 55825 - Federal Motor Vehicle Safety Standards, Child Restraint Systems
2011-09-09
... [Docket No. NHTSA-2011-0139] RIN 2127-AJ44 Federal Motor Vehicle Safety Standards, Child Restraint Systems..., amends a provision in Federal Motor Vehicle Safety Standard No. 213, ``Child restraint systems,'' that... provision: When a motor vehicle safety standard is in effect under this chapter, a State or a political...
Environmental safety of the disposal system for radioactive substance-contaminated wastes
International Nuclear Information System (INIS)
Oosako, Masahiro
2012-01-01
In accordance with the full-scale enforcement of 'The Act on Special Measures concerning the Handling of Radioactive Pollution' in 2012, the collective efforts of entire Japan for dealing with radioactive pollutants began. The most important item for dealing with radioactive pollution is to control radioactive substances that polluted the global environment and establish a contaminated waste treatment system for risk reduction. On the incineration system and landfill disposal system of radioactive waste, this paper arranges the scientific information up to now, and discusses the safety of the treatment / disposal systems of contaminated waste. As for 'The Act on Special Measures concerning the Handling of Radioactive Pollution,' this paper discusses the points of the Act and basic policy, roadmap for the installation of interim storage facilities, and enforcement regulations (Ordinance of the Ministry of the Environment). About the safety of waste treatment system, it discusses the safety level of technical standards at waste treatment facilities, safety of incineration facilities, and safety of landfill disposal sites. (O.A.)
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)
Editorial: emerging issues in sociotechnical systems thinking and workplace safety.
Noy, Y Ian; Hettinger, Lawrence J; Dainoff, Marvin J; Carayon, Pascale; Leveson, Nancy G; Robertson, Michelle M; Courtney, Theodore K
2015-01-01
The burden of on-the-job accidents and fatalities and the harm of associated human suffering continue to present an important challenge for safety researchers and practitioners. While significant improvements have been achieved in recent decades, the workplace accident rate remains unacceptably high. This has spurred interest in the development of novel research approaches, with particular interest in the systemic influences of social/organisational and technological factors. In response, the Hopkinton Conference on Sociotechnical Systems and Safety was organised to assess the current state of knowledge in the area and to identify research priorities. Over the course of several months prior to the conference, leading international experts drafted collaborative, state-of-the-art reviews covering various aspects of sociotechnical systems and safety. These papers, presented in this special issue, cover topics ranging from the identification of key concepts and definitions to sociotechnical characteristics of safe and unsafe organisations. This paper provides an overview of the conference and introduces key themes and topics. Sociotechnical approaches to workplace safety are intended to draw practitioners' attention to the critical influence that systemic social/organisational and technological factors exert on safety-relevant outcomes. This paper introduces major themes addressed in the Hopkinton Conference within the context of current workplace safety research and practice challenges.
Testing Challenges of Maritime Safety and Security Systems-of-Systems
Gonzalez, A.; Piel, E.; Gross, H.G.
2008-01-01
Preprint of paper published in: TAIC PART 2008: Testing: Academic & Industrial Conference, Practice and Research Techniques, 29-31 August 2008; doi:10.1109/TAIC-PART.2008.14 Maritime Safety and Security systems represent a novel kind of large-scale distributed component-based systems in which the
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.
Directory of Open Access Journals (Sweden)
Kołowrocki Krzysztof
2017-06-01
Full Text Available The paper presents an integrated general model of complex technical system, linking its multistate safety model and the model of its operation process including operating environment threats and considering variable at different operation states its safety structures and its components safety parameters. Under the assumption that the system has exponential safety function, the safety characteristics of the port oil piping transportation system are determined.
Kołowrocki, Krzysztof; Kuligowska, Ewa; Soszyńska-Budny, Joanna
2017-01-01
The paper presents an integrated general model of complex technical system, linking its multistate safety model and the model of its operation process including operating environment threats and considering variable at different operation states its safety structures and its components safety parameters. Under the assumption that the system has exponential safety function, the safety characteristics of the port oil piping transportation system are determined.
International Nuclear Information System (INIS)
Song, Tae Young
2007-01-01
At present, the 10-year Periodic Safety Review(PSR) has been performing to confirm all the aspects of safety issues for all the operating plants in compliance with domestic nuclear law of article 23, subarticle 3. For each plant, in addition, Probabilistic Safety Assessment(PSA) and Severe Accident Management Guideline(SAMG) are being implemented and revised periodically to reflect the latest safety level according to principle fulfillment of severe accident policy statement. The assessment reports, as one of outcomes from these activities, are submitted into and reviewed by domestic regulatory body. During reviewing (in-office duty) and licensing (regulatory duty) process, a large number of outcomes of which most are the formal technical reports and licensing materials, are inevitably produced. Moreover, repeated review process over the plants can make them accumulated and produce a variety of documents additionally. This circumstance motivates to develop effective tool or system for the management of these reports and related technical documents for the future use in licensing process and for subsequent plant assessments. This paper presents the development status of Safety Assessment Information System(SAIS) which manages safety-related documents of PSR, PSA and SAMG for practical use for experienced engineers in charge of these areas
Energy Technology Data Exchange (ETDEWEB)
Song, Tae Young [Nuclear Engineering and Technology Institute, Daejeon (Korea, Republic of)
2007-07-01
At present, the 10-year Periodic Safety Review(PSR) has been performing to confirm all the aspects of safety issues for all the operating plants in compliance with domestic nuclear law of article 23, subarticle 3. For each plant, in addition, Probabilistic Safety Assessment(PSA) and Severe Accident Management Guideline(SAMG) are being implemented and revised periodically to reflect the latest safety level according to principle fulfillment of severe accident policy statement. The assessment reports, as one of outcomes from these activities, are submitted into and reviewed by domestic regulatory body. During reviewing (in-office duty) and licensing (regulatory duty) process, a large number of outcomes of which most are the formal technical reports and licensing materials, are inevitably produced. Moreover, repeated review process over the plants can make them accumulated and produce a variety of documents additionally. This circumstance motivates to develop effective tool or system for the management of these reports and related technical documents for the future use in licensing process and for subsequent plant assessments. This paper presents the development status of Safety Assessment Information System(SAIS) which manages safety-related documents of PSR, PSA and SAMG for practical use for experienced engineers in charge of these areas.
A regulatory frame for safety digital systems in nuclear power plants
International Nuclear Information System (INIS)
Mozas Garcia, A.
1998-01-01
The paper focuses on Spanish experience regarding software based systems for safety applications from the regulator's point of view. It describes the actual situation in Spain, number and models of reactors, modernization projects, digital systems implemented and licensing documentation and processes already followed by some upgrading projects. The paper wonders what documents should be required for safety and reliability demonstration of a safety system, when they should be reviewed, and what other activities may be necessary to acquire confidence on a particular system. It describes Spanish laws regarding nuclear safety under which, national standards from the NPP design original country apply to nuclear reactors in Spain. It finally suggests that an international standard jointly used by system manufacturers, nuclear licensees and nuclear safety authorities, both from the country where the NPP is installed, and from the original design country, should be developed so that rapid and easy agreement on licensing issues is reached among all parties. The last part of the paper describes the licensing approach proposed by CSN (Spanish Nuclear Safety Authority). It is still under development and it is based on previous experience on digital systems for non-safety applications. It consists of constructing several frames: 1) databases of existing software based systems, 2) guides for inspection and 3) questionnaires for helping in verification and validation activities evaluation. The scope is to establish a well defined procedure that helps in evaluating the particular system. However, in order for such a procedure to be useful, both regulators and utilities and, perhaps also system manufacturers, should agree on it. Joint CSN-utilities working groups may be suitable for such a purpose. (author)
System and safety studies of accelerator driven systems for transmutation. Annual report 2007
Energy Technology Data Exchange (ETDEWEB)
Arzhanov, Vasily; Fokau, Andrei; Persson, Calle; Runevall, Odd; Sandberg, Nils; Tesinsky, Milan; Wallenius, Janne; Youpeng Zhang (Div. of Reactor Physics, Royal Institute of Technology, Stockholm (Sweden))
2008-05-15
Within the project 'System and safety studies of accelerator driven systems for transmutation', research on design and safety of sub-critical reactors for recycling of minor actinides is performed. During 2007, the reactor physics division at KTH has calculated safety parameters for EFIT-400 with cermet fuel, permitting to start the transient safety analysis. The accuracy of different reactivity meters applied to the YALINA facility was assessed and neutron detection studies were performed. A model to address deviations from point kinetic behaviour was developed. Studies of basic radiation damage physics included calculations of vacancy formation and activation enthalpies in bcc niobium. In order to predict the oxygen potential of inert matrix fuels, a thermo-chemical model for mixed actinide oxides was implemented in a phase equilibrium code
Software for the occupational health and safety integrated management system
International Nuclear Information System (INIS)
Vătăsescu, Mihaela
2015-01-01
This paper intends to present the design and the production of a software for the Occupational Health and Safety Integrated Management System with the view to a rapid drawing up of the system documents in the field of occupational health and safety
Software for the occupational health and safety integrated management system
Energy Technology Data Exchange (ETDEWEB)
Vătăsescu, Mihaela [University Politehnica Timisoara, Department of Engineering and Management, 5 Revolutiei street, 331128 Hunedoara (Romania)
2015-03-10
This paper intends to present the design and the production of a software for the Occupational Health and Safety Integrated Management System with the view to a rapid drawing up of the system documents in the field of occupational health and safety.
International Nuclear Information System (INIS)
Chang, Soon Heung; Kim, Sang Ho; Choi, Jae Young
2013-01-01
Highlights: • We newly propose the design concept of integrated passive safety system (IPSS). • It has five safety functions for decay heat removal and severe accident mitigation. • Simulations for IPSS show that core melt does not occur in accidents with SBO. • IPSS can achieve the passive in-vessel retention and ex-vessel cooling strategy. • The applicability of IPSS is high due to the installation outside the containment. -- Abstract: The design concept of integrated passive safety system (IPSS) which can perform various passive safety functions is proposed in this paper. It has the various functions of passive decay heat removal system, passive safety injection system, passive containment cooling system, passive in-vessel retention and cavity flooding system, and filtered venting system with containment pressure control. The objectives of this paper are to propose the conceptual design of an IPSS and to estimate the design characters of the IPSS with accident simulations using MARS code. Some functions of the IPSS are newly proposed and the other functions are reviewed with the integration of the functions. Consequently, all of the functions are modified and integrated for simplicity of the design in preparation for beyond design based accidents (BDBAs) focused on a station black out (SBO). The simulation results with the IPSS show that the decay heat can be sufficiently removed in accidents that occur with a SBO. Also, the molten core can be retained in a vessel via the passive in-vessel retention strategy of the IPSS. The actual application potential of the IPSS is high, as numerous strong design characters are evaluated. The installation of the IPSS into the original design of a nuclear power plant requires minimal design change using the current penetrations of the containment. The functions are integrated in one or two large tanks outside the containment. Furthermore, the operation time of the IPSS can be increased by refilling coolant from the
Safety review on unit testing of safety system software of nuclear power plant
International Nuclear Information System (INIS)
Liu Le; Zhang Qi
2013-01-01
Software unit testing has an important place in the testing of safety system software of nuclear power plants, and in the wider scope of the verification and validation. It is a comprehensive, systematic process, and its documentation shall meet the related requirements. When reviewing software unit testing, attention should be paid to the coverage of software safety requirements, the coverage of software internal structure, and the independence of the work. (authors)
Performance Test Results of Safety I and C Systems of SMART MMIS
International Nuclear Information System (INIS)
Suh, Yong Suk; Keum, Jong Yong; Jeong, Kwang Il; Lee, Joon Ku; Lee, Sang Seok; Kim, Kwan Woong
2011-01-01
KAERI has developed SMART (System-integrated Modular Advanced ReacTor), a 330MWt integral pressurized light water reactor that integrates four reactor coolant pumps, one pressurizer, eight steam generators, and one reactor core into a reactor vessel, since 1997 and submitted a SSAR (Standard design Safety Analysis Report) to Korea institute of nuclear safety (KINS) at the end of 2010 for the purpose of achieving the standard design approval (SDA) by the end of 2011. SMART MMIS has been designed with fully digitalized systems. Non-safety instrumentation and control (I and C) systems are designed based on the commercial distributed control systems. The safety I and C systems are designed using a new platform that was developed and validated by KAERI. Safety I and C systems are modularized using the platform. In the protection systems (PSs), datalinks are used to transmit data in a one-way direction in order to meet the independency requirement. In the engineered safety features-component control system (ESF-CCS), network switch devices (NSDs) are used to connect the group and loop controllers. The NSD was also newly developed and validated by KAERI. After validating the platform and NSD, a test facility was developed using the platform and NSDs to validate the performance of safety I and C systems. This paper presents the development and test results from the test facility
Feasibility and Safety of Endovascular Stripping of Totally Implantable Venous Access Devices
International Nuclear Information System (INIS)
Heye, Sam; Maleux, Geert; Goossens, G. A.; Vaninbroukx, Johan; Jerôme, M.; Stas, M.
2012-01-01
Purpose: To evaluate the safety and feasibility of percutaneous stripping of totally implantable venous access devices (TIVAD) in case of catheter-related sleeve and to report a technique to free the catheter tip from vessel wall adherence. Materials and Methods: A total of 37 stripping procedures in 35 patients (14 men, 40%, and 21 women, 60%, mean age 53 ± 14 years) were reviewed. Totally implantable venous access devices were implanted because of malignancy in most cases (85.7%). Catheter-related sleeve was confirmed as cause of persistent catheter dysfunction despite instillation of thrombolytics. A technique to mobilize the catheter tip from the vessel wall was used when stripping with the snare catheter was impossible. Technical success, complication rate, and outcome were noted. Results: A total of 55.9% (n = 19) of the 34 technically successful procedures (91.9%) could be done with the snare catheter. In 15 cases (44.1%), additional maneuvers to free the TIVAD’s tip from the vessel wall were needed. Success rate was not significantly lower before (72.4%) than after (96.7%) implementation of the new technique (P = 0.09). No complications were observed. Follow-up was available in 67.6% of cases. Recurrent catheter dysfunction was found in 17 TIVADs (78.3%) at a mean of 137.7 days and a median of 105 days. Conclusions: Stripping of TIVADs is technically feasible and safe, with an overall success rate of 91.9%. Additional endovascular techniques to mobilize the distal catheter tip from the wall of the superior vena cava or right atrium to allow encircling the TIVAD tip with the snare catheter may be needed in 44.1% of cases.
Development of FPGA-based safety-related I and C systems
Energy Technology Data Exchange (ETDEWEB)
Goto, Y.; Oda, N.; Miyazaki, T.; Hayashi, T.; Sato, T.; Igawa, S. [08, Shinsugita-cho, Isogo-ku, Yokohama 235-8523 (Japan); 1, Toshiba-cho, Fuchu, Tokyo 183-8511 (Japan)
2006-07-01
Toshiba has developed Non-rewritable (NRW) Field Programmable Gate Array (FPGA)-based safety-related Instrumentation and Control (I and C) system [1]. 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 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 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 Monitor (PRM). 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)
Feng, Zhitao; Xu, Juan; He, Guochao; Cao, Meiqun; Duan, Lihong; Chen, Liguo; Wu, Zhengzhi
2016-01-01
Objective. To evaluate the efficacy and safety of the total glucosides of peony (TGP) and leflunomide (LEF) for the treatment of rheumatoid arthritis (RA). Methods. Randomized controlled trials (RCTs) on the efficacy and safety of the combination of TGP and LEF versus LEF alone for the treatment of RA were retrieved by searching PubMed, EMBASE, Cochrane Library, the China National Knowledge Infrastructure database, and Wanfang database. Results. Eight RCTs including 643 RA patients were included in the present meta-analysis. The quality of included studies was poor. The levels of ESR (P TGP and LEF were significantly lower than RA patients who received LEF therapy alone. The pooled results suggest that the combination of TGP and LEF caused less abnormal liver function than LEF alone (P = 0.02). No significant difference in the gastrointestinal discomfort was identified between the combination of TGP and LEF and LEF alone groups (P = 0.18). Conclusion. The combination of TGP and LEF in treatment of RA presented the characteristics of notably decreasing the levels of laboratory indexes and higher safety in terms of liver function. However, this conclusion should be further investigated based on a larger sample size.
Employee Engagement and a Culture of Safety in the Intensive Care Unit.
Collier, Susan L; Fitzpatrick, Joyce J; Siedlecki, Sandra L; Dolansky, Mary A
2016-01-01
A descriptive, retrospective design was used to explore the relationship between employee engagement and culture of safety in ICUs within a large Midwestern healthcare system. Results demonstrated a strong positive relationship between total engagement score and total patient safety score (r = 0.645, P engagement score and the 12 safety culture dimensions. These findings have implications for improving managerial strategies relative to employee engagement that may ultimately impact perceptions of a safety culture.
TOTAL QUALITY MANAGEMENT SYSTEM (TQMS IN SMALL WINERY AND BAKERY IN CYPRUS. A CASE STUDY.
Directory of Open Access Journals (Sweden)
ANTONIS A. ZORPAS
2012-06-01
Full Text Available Quality of products and production affects the welfare, efficiency and competitiveness of the business entity. Total Quality Management System (TQMS represent the organizational structure of enterprise, application of methods, tools and processes used for the implantation, maintenance and improvement of activities and their outcomes as well as the organization takes all the necessary actions for its environmental performance and safety.The aim of the implementing TQMS in company is to create a good relation of employees to quality and safety as a fundamental category of the market economy and make them produce this quality and on the other and to provide good relationbetween SMEs and governmental authorities. QMS determine the prior characteristics of products, through which the company meets the identified needs of the market. The paper presents a case study of the implementation of aTQMS in small winery and bakery industry from Cyprus. The TQMS include the implementation of ISO 9001:2008, ISO 14001:2004 and ISO 22000:2005.
Modernization of safety system for the radiation facility for industrial sterilization
International Nuclear Information System (INIS)
Drndarevic, V.; Djuric, D.; Koturovic, A.; Arandjelovic, M.; Mikic, R.
1995-01-01
Modernization of the existing safety system of the radiation facility for industrial sterilization at the Vinca Institute of nuclear science is done. In order to improve radiation safety of the facility, the latest recommendations and requirements of IAEA have been implemented. Concept and design of the modernized system are presented. The new elements of the safety system are described and the improvements achieved by means of this modernization are pointed out. (author)
Fiyanzar, Adin Eka; Nusraningrum, Dewi; Arofat, Osman
2016-01-01
This study aimed to analyze the effect of the implementation of Safety Management System (SMS) and the use of information system on the Flight Safety in the Indonesian Air Navigation Services Organization both partially and simultaneously. The research uses quantitative methods, and the data are analyzed using linear regression, simple correlation both partially and simultaneously and path analysis. The result shows; implementation of Safety Management System (X1) as measured by the Flight Sa...
Total Analysis System for Ship Structural Strength
Takuya, Yoneya; Hiroyuki, Kobayashi; Abdul M., Rahim; Yoshimichi, Sasaki; Masaki, Irisawa; Technical Investigation and Information Department, Research Center; Technical Investigation and Information Department, Research Center; Singapore Office; Technical Investigation and Information Department, Research Center; Technical Investigation and Information Department, Research Center
2001-01-01
This paper outlines a total analysis system for ship hull structures, which integrates a wide variety of analysis functions to realise practical applications of rational methods for assessing ship structural strength. It is based on direct calculation of wave-induced loads as well as three-dimensional structural analysis of an entire-ship or hold structure. Three major analysis functions of the total system are ship motion and wave load analysis, ship structural analysis and statistical analy...
Safety study of PCC 2140 and ALILOG 21 used as part of safety measurement systems
International Nuclear Information System (INIS)
Meriaux, Pierre; Adnot, Serge; Rayrolles, Catherine.
1978-03-01
The PCC 2140 and ALILOG 21 equipment may be used at C.E.A. or E.D.F., as part of safety measurement systems. In a study of a similar, but earlier equipment, it was noticed that certain types of failures caused the system to switch to the least sensitive measurement range, which was detrimental to safety. This report analyses failure modes leading to unsafe failures and evaluates the risks ran into taking in account tests during use [fr
Noise problems in PFR safety system power supplies
Energy Technology Data Exchange (ETDEWEB)
Hodgson, D
1978-08-15
Reactor power on PFR is controlled by eleven neutron absorbing rods. They are held vertically above the core by electromagnets which, in turn, are mounted on motorised lead screws. Five of the rods are used as power control elements; five are used as shut off rods and the eleventh is a safety rod, positioned vertically above the centre of the reactor core. All rods fall under gravity into the reactor when the electro magnets are de-energised. In addition the safety rod can be driven into the reactor by a pneumatic piston and cylinder mechanism. This provides some degree of diversity from the common design of mechanism used by the control and shut off rods. Also being in the center it can be armed and ready for use when the above core shield is rotated. The automatic protective system for the plant provides the electro magnet currents and hence the reactor trip signals via two separate and diverse safety systems, namely a ''Relay System'' and a ''Solid State System.'' The two systems are completely separate and independent. The overall probability of failure of the complete system to release at least 3 rods when called upon to do so is less than 10/sup -6/.
Digital Signal Processing for In-Vehicle Systems and Safety
Boyraz, Pinar; Takeda, Kazuya; Abut, Hüseyin
2012-01-01
Compiled from papers of the 4th Biennial Workshop on DSP (Digital Signal Processing) for In-Vehicle Systems and Safety this edited collection features world-class experts from diverse fields focusing on integrating smart in-vehicle systems with human factors to enhance safety in automobiles. Digital Signal Processing for In-Vehicle Systems and Safety presents new approaches on how to reduce driver inattention and prevent road accidents. The material addresses DSP technologies in adaptive automobiles, in-vehicle dialogue systems, human machine interfaces, video and audio processing, and in-vehicle speech systems. The volume also features: Recent advances in Smart-Car technology – vehicles that take into account and conform to the driver Driver-vehicle interfaces that take into account the driving task and cognitive load of the driver Best practices for In-Vehicle Corpus Development and distribution Information on multi-sensor analysis and fusion techniques for robust driver monitoring and driver recognition ...
NASA aviation safety reporting system
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.
Safety management of a complex R and D ground operating system
Connors, J. F.; Maurer, R. A.
1975-01-01
A perspective on safety program management was developed for a complex R&D operating system, such as the NASA-Lewis Research Center. Using a systems approach, hazardous operations are subjected to third-party reviews by designated-area safety committees and are maintained under safety permit controls. To insure personnel alertness, emergency containment forces and employees are trained in dry-run emergency simulation exercises. The keys to real safety effectiveness are top management support and visibility of residual risks.
Ex-ante assessment of the safety effects of intelligent transport systems.
Kulmala, Risto
2010-07-01
There is a need to develop a comprehensive framework for the safety assessment of Intelligent Transport Systems (ITS). This framework should: (1) cover all three dimensions of road safety-exposure, crash risk and consequence, (2) cover, in addition to the engineering effect, also the effects due to behavioural adaptation and (3) be compatible with the other aspects of state of the art road safety theories. A framework based on nine ITS safety mechanisms is proposed and discussed with regard to the requirements set to the framework. In order to illustrate the application of the framework in practice, the paper presents a method based on the framework and the results from applying that method for twelve intelligent vehicle systems in Europe. The framework is also compared to two recent frameworks applied in the safety assessment of intelligent vehicle safety systems. Copyright 2010 Elsevier Ltd. All rights reserved.
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....
A sensor monitoring system for telemedicine, safety and security applications
Vlissidis, Nikolaos; Leonidas, Filippos; Giovanis, Christos; Marinos, Dimitrios; Aidinis, Konstantinos; Vassilopoulos, Christos; Pagiatakis, Gerasimos; Schmitt, Nikolaus; Pistner, Thomas; Klaue, Jirka
2017-02-01
A sensor system capable of medical, safety and security monitoring in avionic and other environments (e.g. homes) is examined. For application inside an aircraft cabin, the system relies on an optical cellular network that connects each seat to a server and uses a set of database applications to process data related to passengers' health, safety and security status. Health monitoring typically encompasses electrocardiogram, pulse oximetry and blood pressure, body temperature and respiration rate while safety and security monitoring is related to the standard flight attendance duties, such as cabin preparation for take-off, landing, flight in regions of turbulence, etc. In contrast to previous related works, this article focuses on the system's modules (medical and safety sensors and associated hardware), the database applications used for the overall control of the monitoring function and the potential use of the system for security applications. Further tests involving medical, safety and security sensing performed in an real A340 mock-up set-up are also described and reference is made to the possible use of the sensing system in alternative environments and applications, such as health monitoring within other means of transport (e.g. trains or small passenger sea vessels) as well as for remotely located home users, over a wired Ethernet network or the Internet.
International Nuclear Information System (INIS)
Nishimoto, Masae; Umezawa, Shigemitsu; Okabe, Kazuharu; Matsuoka, Tsuyoshi
1996-01-01
A Hybrid Safety System, which is an optimum combination of active and passive safety systems, has been developed in order to improve the safety, reliability and economic features of the next generation of PWRs. The passive safety systems include Automatic primary Depressurization System (ADS), Secondary Automatic Depressurization System (SADS), advanced accumulators, gravity injection system and so on. In this study the authors have improved the actuation logic of the passive safety systems. The original logic in the previous study actuates ADS at an early stage of an event such as a Loss-of-Coolant Accident (LOCA), and this is followed by the actuation of SADS. In this study they divide SADS into two systems. The first, small SADS, uses small valves corresponding to the relief valves of the conventional PWR plants. The second, large SADS, corresponds to the original SADS using multiple valves of large capacity. With the new logic, the passive systems are actuated during a typical small LOCA. Small LOCA analyses using several break areas were performed for a 1,400 MWe PWR plant with a Hybrid Safety System. The results predict that core uncovery does not occur in the case of a relatively small break area and that core heat removal during a small LOCA is improved in comparison with the analyses for conventional PWR plants, where the secondary pressure remains higher during the event. The results also predict that this new logic make it possible to reduce the ADS valve size and the actuation pressure setpoint of the passive safety systems
SWR 1000: the main design features of the advanced boiling water reactor with passive safety systems
International Nuclear Information System (INIS)
Carsten, Pasler
2007-01-01
The SWR-1000 (1000 MW) is a boiling water reactor whose economic efficiency in comparison with large-capacity designs is achieved by deploying very simple passive safety equipment, simplified systems for plant operation, and a very simple plant configuration in which systems engineering is optimized and dependence on electrical and instrumentation and control systems is reduced. In addition, systems and components that require protection against natural and external man-made hazards are accommodated in such a way that as few buildings as possible have to be designed to withstand the loads from such events. The fuel assemblies have been enlarged from a 10*10 rod array to a 12*12 array. This reduces the total number of fuel assemblies in the core and thus also the number of control rods and control rod drives, as well as in-core neutron flux monitors. The design owes its competitiveness to the fact that investment costs, maintenance costs and fuel cycle costs are all lower. In addition, refueling outages are shorter, thanks to the reduced scope of outage activities. The larger fuel assemblies have been extensively and successfully tested, as have all of the other new components and systems incorporated into the plant design. As in existing plants, the forced coolant circulation method is deployed, ensuring problem-free startup, and enabling plant operators to adjust power rapidly in the high power range (70%-100%) without moving the control rods, as well as allowing spectral-shift and stretch-out operation. The plant safety concept is based on a combination of passive safety systems and a reduced number of active safety systems. All postulated accidents can be controlled using passive systems alone. Control of a postulated core melt accident is assured with considerable safety margins thanks to passive flooding of the containment for in-vessel melt retention. The SWR-1000 is compliant with international nuclear codes and standards, and is also designed to withstand
Innovation of Supervision System for Quality and Safety of Edible Agricultural Products
Institute of Scientific and Technical Information of China (English)
Xingxing; MEI; Zhongchao; FENG
2014-01-01
This paper elaborated multidimensional characteristics of quality and safety of agricultural products,introduced current situation of quality and safety supervision of edible agricultural products in China,analyzed existing problems of quality and safety supervision system and corresponding reasons,and finally came up with recommendations for innovation of supervision system for quality and safety of agricultural products.
Visualization of Safety Assessment Result Using GIS in SITES
International Nuclear Information System (INIS)
Yun, Bong-Yo; Park, Joo Wan; Park, Se-Moon; Kim, Chang-Lak
2006-01-01
Site Information and Total Environmental database management System (SITES) is an integrated program for overall data analysis, environmental monitoring, and safety analysis that are produced from the site investigation and environmental assessment of the relevant nuclear facility. SITES is composed of three main modules such as Site Environment Characterization database for Unified and Reliable Evaluation system (SECURE), Safety Assessment INTegration system (SAINT) and Site Useful Data Analysis and ALarm system (SUDAL). The visualization function of safety assessment and environmental monitoring results is designed. This paper is to introduce the visualization design method using Geographic Information System (GIS) for SITES
Application of Safety Instrumented System (SIS) approach in older nuclear power plants
Energy Technology Data Exchange (ETDEWEB)
Nasimi, Elnara; Gabbar, Hossam A., E-mail: hossam.gabbar@uoit.ca
2016-05-15
Highlights: • Study Safety Instrumented System (SIS) design for older nuclear power plant. • Apply SIS on Reheater Drains (RD) system. • Apply IEC 61508/61511 to design safety system. • Evaluate risk reduction based on proposed SIS design. - Abstract: In order to remain economically effective and financially profitable, the modern industries have to take their safety culture to a higher level and consider production losses in addition to simple accident prevention techniques. Ideally, compliance with safety requirements start during early design stages, but in some older facilities provisions for Safety Instrumented Systems (SIS) may not have been originally included. In this paper, a case study of a Reheater Drains (RD) system is used to illustrate such an example. Frequent failures of tank level controller lead to transients where the operation of shutting down RD pumps requires operators to manually isolate the quenching water and to close the main steam admission valves. Water in this system is at saturation temperature for the reheater steam side pressure, and any manual operation of the system is highly undesirable due to hazards of working with wet steam at approximately 758 kPa(g) pressure, preheated to 237 °C. Additionally, losses of inventory are highly undesirable as well and challenge other systems in the plant. In this paper, it is suggested that RD system can benefit from installation of an independent SIS system in order to address current challenges. This idea is being explored using IEC 61508 framework for “Functional safety of electrical/electronic/programmable electronic safety-related systems” to provide assurance that the SIS will offer the necessary risk reduction required to achieve required safety for the equipment.
Knowledge management and safety compliance in a high-risk distributed organizational system.
Gressgård, Leif Jarle
2014-06-01
In a safety perspective, efficient knowledge management is important for learning purposes and thus to prevent errors from occurring repeatedly. The relationship between knowledge exchange among employees and safety behavior may be of particular importance in distributed organizational systems where similar high-risk activities take place at several locations. This study develops and tests hypotheses concerning the relationship between knowledge exchange systems usage, knowledge exchange in the organizational system, and safety compliance. The operational context of the study is petroleum drilling and well operations involving distributed high-risk activities. The hypotheses are tested by use of survey data collected from a large petroleum operator company and eight of its main contractors. The results show that safety compliance is influenced by use of knowledge exchange systems and degree of knowledge exchange in the organizational system, both within and between units. System usage is the most important predictor, and safety compliance seems to be more strongly related to knowledge exchange within units than knowledge exchange between units. Overall, the study shows that knowledge management is central for safety behavior.
On the safety performance of the advanced nuclear energy systems
International Nuclear Information System (INIS)
Li Shounan
1999-01-01
Some features on the safety performances of the Advanced Nuclear Energy Systems are discussed. The advantages and some peculiar problems on the safety of Advanced Nuclear Energy Systems with subcritical nuclear reactor driven by external neutron sources are also pointed out in comparison with conventional nuclear reactors
23 CFR 973.212 - Indian lands safety management system (SMS).
2010-04-01
... implementation of public information and education activities on safety needs, programs, and countermeasures... 23 Highways 1 2010-04-01 2010-04-01 false Indian lands safety management system (SMS). 973.212... HIGHWAYS MANAGEMENT SYSTEMS PERTAINING TO THE BUREAU OF INDIAN AFFAIRS AND THE INDIAN RESERVATION ROADS...
Assessment of Safety Standards for Automotive Electronic Control Systems
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 - ...
Energy Technology Data Exchange (ETDEWEB)
NONE
1985-03-01
The automatic sewing system technique research association has been commissioned by the Agency of Industrial Science and Technology for (research and development of automatic sewing systems). This program covers R and D of the elementary techniques for total systems and sewing preparation/processing, sewing/assembling, cloth handling, and system management/control. This report describes the results of the R and D efforts for the total systems. For producing diversified products in small quantities, which is the object of this R and D program, the target is set to reduce processing time by at least 50% from the current level. The efforts to achieve this target by solving the problems involved therein are made from the side of developing FMS and FA procedures, and apparel side. The program for the former investigates the required functions of individual sub-systems that constitute the total system, to clarify the technical targets of the core elementary techniques for each sub-system. The program for the latter analyzes the sewing steps for producing diversified clothes types, specifications for automatic operation, integration, standardization and coding of the steps, and so on, based on the results of studies on basic apparel types. (NEDO)
Effect of change in half-life of Se-79 on the safety of HLW geological disposal system
International Nuclear Information System (INIS)
Ishihara, Yoshinao; Ishiguro, Katsuhiko; Umeki, Hiroyuki
1999-11-01
Se-79 is one of key radionuclides in the performance assessment of the geological disposal system. Based on recent measurements, it is possible that the half-life of Se-79 will be changed longer than the present value in most handbooks and tables of isotopes. This study presents performance assessment calculations to investigate the overall effect of change in half-life of Se-79 on the repository system safety. The total system performance analyses for Se-79 were carried out, which focussed on the Reference-Case of the safety assessment in the H12 Project. As results, the maximum release rate in Becquerel unit of Se-79 from the engineered barrier system with new half-life decreases about one order of magnitude than that with half-life used so far. It is, however, that the maximum release rate in Becquerel unit of Se-79 from the natural barrier system is almost same for both half-life because of the channelling effects of groundwater flow. Consequently, the calculated maximum dose rate of Se-79 with new half-life does not change. It can be concluded that the change in half-life of Se-79 does not affect overall safety of the H12 disposal concept. (author)
Design of reactor containment systems for nuclear power plants. Safety guide
International Nuclear Information System (INIS)
2008-01-01
This Safety Guide was prepared under the IAEA programme for safety standards for nuclear power plants. It is a revision of the Safety Guide on Design of the Reactor Containment Systems in Nuclear Power Plants (Safety Series No. 50-Sg-D1) issued in 1985 and supplements the Safety Requirements publication on Safety of Nuclear Power Plants: Design. The present Safety Guide was prepared on the basis of a systematic review of the relevant publications, including the Safety of Nuclear Power Plants: Design, the Safety fundamentals publication on The Safety of Nuclear Installations, Safety Guides, INSAG Reports, a Technical Report and other publications covering the safety of nuclear power plants. 1.2. The confinement of radioactive material in a nuclear plant, including the control of discharges and the minimization of releases, is a fundamental safety function to be ensured in normal operational modes, for anticipated operational occurrences, in design basis accidents and, to the extent practicable, in selected beyond design basis accidents. In accordance with the concept of defence in depth, this fundamental safety function is achieved by means of several barriers and levels of defence. In most designs, the third and fourth levels of defence are achieved mainly by means of a strong structure enveloping the nuclear reactor. This structure is called the 'containment structure' or simply the 'containment'. This definition also applies to double wall containments. 1.3. The containment structure also protects the reactor against external events and provides radiation shielding in operational states and accident conditions. The containment structure and its associated systems with the functions of isolation, energy management, and control of radionuclides and combustible gases are referred to as the containment systems
Design of reactor containment systems for nuclear power plants. Safety guide
International Nuclear Information System (INIS)
2004-01-01
This Safety Guide was prepared under the IAEA programme for safety standards for nuclear power plants. It is a revision of the Safety Guide on Design of the Reactor Containment Systems in Nuclear Power Plants (Safety Series No. 50-Sg-D1) issued in 1985 and supplements the Safety Requirements publication on Safety of Nuclear Power Plants: Design. The present Safety Guide was prepared on the basis of a systematic review of the relevant publications, including the Safety of Nuclear Power Plants: Design, the Safety fundamentals publication on The Safety of Nuclear Installations, Safety Guides, INSAG Reports, a Technical Report and other publications covering the safety of nuclear power plants. 1.2. The confinement of radioactive material in a nuclear plant, including the control of discharges and the minimization of releases, is a fundamental safety function to be ensured in normal operational modes, for anticipated operational occurrences, in design basis accidents and, to the extent practicable, in selected beyond design basis accidents. In accordance with the concept of defence in depth, this fundamental safety function is achieved by means of several barriers and levels of defence. In most designs, the third and fourth levels of defence are achieved mainly by means of a strong structure enveloping the nuclear reactor. This structure is called the 'containment structure' or simply the 'containment'. This definition also applies to double wall containments. 1.3. The containment structure also protects the reactor against external events and provides radiation shielding in operational states and accident conditions. The containment structure and its associated systems with the functions of isolation, energy management, and control of radionuclides and combustible gases are referred to as the containment systems
International Nuclear Information System (INIS)
Dahll, Gustav; Liwang, Bo; Wainwright, Norman
2006-01-01
The introduction of new software based technology in the safety systems in nuclear power plants also makes it necessary to develop new strategies for regulatory review and assessment of these new systems that is more focused on reviewing the processes at the different phases in design phases during the system life cycle. It is a general requirement that the licensee shall perform different kinds of reviews. From a regulatory point of view it is more cost effective to assess that the design activities at the suppliers and the review activities within the development project are performed with good quality. But the change from more technical reviews over to the development process oriented approach also cause problems. When reviewing development and quality aspects there are no 'hard facts' that can be judged against some specified criteria, the issues are more 'soft' and are more to build up structure of arguments and evidences that the requirements are met. The regulatory review strategy must therefore change to follow the development process over the whole life cycle from concept phase until installation and operation. Even if we know what factors that is of interest we need some guidance on how to interpret and judge the information.For that purpose SKl started research activities in this area at the end of the 1990s. In the first phase, in co-operation with Gustav Dahll at the Halden project, a life cycle model was selected. For the different phases a qualitative influence net was constructed of the type that is used in Bayesian Believe Network together with a discussion on different issues involved. In the second phase of the research work, in co-operation with Norman Wainwright, a former NII inspector, information from a selection of the most important sources as guidelines, IAEA and EC reports etc, was mapped into the influence net structure (the total list on used sources are in the report). The result is presented in the form of questions (Q) and a
Energy Technology Data Exchange (ETDEWEB)
Dahll, Gustav (OECD Halden Project, Halden (NO)); Liwaang, Bo (Swedish Nuclear Power Inspectorate, Stockholm (Sweden)); Wainwright, Norman (Wainwright Safety Advice (GB))
2006-07-01
The introduction of new software based technology in the safety systems in nuclear power plants also makes it necessary to develop new strategies for regulatory review and assessment of these new systems that is more focused on reviewing the processes at the different phases in design phases during the system life cycle. It is a general requirement that the licensee shall perform different kinds of reviews. From a regulatory point of view it is more cost effective to assess that the design activities at the suppliers and the review activities within the development project are performed with good quality. But the change from more technical reviews over to the development process oriented approach also cause problems. When reviewing development and quality aspects there are no 'hard facts' that can be judged against some specified criteria, the issues are more 'soft' and are more to build up structure of arguments and evidences that the requirements are met. The regulatory review strategy must therefore change to follow the development process over the whole life cycle from concept phase until installation and operation. Even if we know what factors that is of interest we need some guidance on how to interpret and judge the information.For that purpose SKl started research activities in this area at the end of the 1990s. In the first phase, in co-operation with Gustav Dahll at the Halden project, a life cycle model was selected. For the different phases a qualitative influence net was constructed of the type that is used in Bayesian Believe Network together with a discussion on different issues involved. In the second phase of the research work, in co-operation with Norman Wainwright, a former NII inspector, information from a selection of the most important sources as guidelines, IAEA and EC reports etc, was mapped into the influence net structure (the total list on used sources are in the report). The result is presented in the form of
Nitric Acid Revamp and Upgrading of the Alarm & Protection Safety System at Petrokemija, Croatia
Directory of Open Access Journals (Sweden)
Hoško, I.
2012-04-01
Full Text Available Every industrial production, particularly chemical processing, demands special attention in conducting the technological process with regard to the security requirements. For this reason, production processes should be continuously monitored by means of control and alarm safety instrumented systems. In the production of nitric acid at Petrokemija d. d., the original alarm safety system was designed as a combination of an electrical relay safety system and transistorized alarm module system. In order to increase safety requirements and modernize the technological process of nitric acid production, revamping and upgrading of the existing alarm safety system was initiated with a new microprocessor system. The newly derived alarm safety system, Simatic PCS 7, links the function of "classically" distributed control (DCS and logical systems in a common hardware and software platform with integrated engineering tools and operator interface to meet the minimum safety standards with safety integrity level 2 (SIL2 up to level 3 (SIL3, according to IEC 61508 and IEC 61511. This professional paper demonstrates the methodology of upgrading the logic of the alarm safety system in the production of nitric acid in the form of a logical diagram, which was the basis for a further step in its design and construction. Based on the mentioned logical diagram and defined security requirements, the project was implemented in three phases: analysis and testing, installation of the safety equipment and system, and commissioning. Developed also was a verification system of all safety conditions, which could be applied to other facilities for production of nitric acid. With the revamped and upgraded interlock alarm safety system, a new and improved safety boundary in the production of nitric acid was set, which created the foundation for further improvement of the production process in terms of improved analysis.
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.
European BWR R and D cluster for innovative passive safety systems
International Nuclear Information System (INIS)
Hicken, E.F.; Lensa, W. von
1996-01-01
The main technological innovation trends for future nuclear power plants tend towards a broader use of passive safety systems for the prevention, mitigation and managing of severe accident scenarios. Several approaches have been undertaken in a number of European countries to study and demonstrate the feasibility and charateristics of innovative passive safety systems. The European BWR R and D Cluster combines those experimental and analytical efforts that are mainly directed to the introduction of passive safety systems into boiling water reactor technology. The Cluster is grouped around thermohydraulic test facilities in Europe for the qualification of innovative BWR safety systems, also taking into account especially the operating experience of the nuclear power plant Dodewaard and other BWRs, which already incorporated some passive safety features. The background, the objectives, the structure of the project and the work programme are presented in this paper as well as an outline of the significance of the expected results. (orig.) [de
Effect of national cultural values on safety climate, and safety management system
International Nuclear Information System (INIS)
Ali, T.H.; Memon, N.A.
2008-01-01
This paper investigates the critical role played by the national culture in influencing how workers safely or otherwise behave (mainly in risky situations) on construction sites, and how site managers implement safety management processes and practices. The paper presents the findings of an empirical research study based on a questionnaire survey, administered in Pakistan, targeting construction site managers and workers to gauge the effect national culture has on managers preferences for and perceptions of safety management systems (policies and practices) and than linking this effect to predict workers attitudes and intentional behaviors. (author)
Energy Technology Data Exchange (ETDEWEB)
Lee, Jang Soo; Cheon, Se Woo; Kim, Chang Hoi; Sim, Yun Sub
2001-02-01
This report has described the software safety analysis techniques and the engineering guidelines for developing safety critical software to identify the state of the art in this field and to give the software safety engineer a trail map between the code and standards layer and the design methodology and documents layer. We have surveyed the management aspects of software safety activities during the software lifecycle in order to improve the safety. After identifying the conventional safety analysis techniques for systems, we have surveyed in details the software safety analysis techniques, software FMEA(Failure Mode and Effects Analysis), software HAZOP(Hazard and Operability Analysis), and software FTA(Fault Tree Analysis). We have also surveyed the state of the art in the software reliability assessment techniques. The most important results from the reliability techniques are not the specific probability numbers generated, but the insights into the risk importance of software features. To defend against potential common-mode failures, high quality, defense-in-depth, and diversity are considered to be key elements in digital I and C system design. To minimize the possibility of CMFs and thus increase the plant reliability, we have provided D-in-D and D analysis guidelines.
International Nuclear Information System (INIS)
Lee, Jang Soo; Cheon, Se Woo; Kim, Chang Hoi; Sim, Yun Sub
2001-02-01
This report has described the software safety analysis techniques and the engineering guidelines for developing safety critical software to identify the state of the art in this field and to give the software safety engineer a trail map between the code and standards layer and the design methodology and documents layer. We have surveyed the management aspects of software safety activities during the software lifecycle in order to improve the safety. After identifying the conventional safety analysis techniques for systems, we have surveyed in details the software safety analysis techniques, software FMEA(Failure Mode and Effects Analysis), software HAZOP(Hazard and Operability Analysis), and software FTA(Fault Tree Analysis). We have also surveyed the state of the art in the software reliability assessment techniques. The most important results from the reliability techniques are not the specific probability numbers generated, but the insights into the risk importance of software features. To defend against potential common-mode failures, high quality, defense-in-depth, and diversity are considered to be key elements in digital I and C system design. To minimize the possibility of CMFs and thus increase the plant reliability, we have provided D-in-D and D analysis guidelines
Application of the management system for facilities and activities. Safety guide
International Nuclear Information System (INIS)
2006-01-01
This Safety Guide supports the Safety Requirements publication on The Management System for Facilities and Activities. It provides generic guidance to aid in establishing, implementing, assessing and continually improving a management system that complies with the requirements established. In addition to this Safety Guide, there are a number of Safety Guides for specific technical areas. Together these provide all the guidance necessary for implementing these requirements. This publication supersedes Safety Series No. 50-SG-Q1-Q7 (1996). The guidance provided here may be used by organizations in the following ways: - To assist in the development of the management systems of organizations directly responsible for operating facilities and activities and providing services for: Nuclear facilities; Activities using sources of ionizing radiation; Radioactive waste management; The transport of radioactive material; Radiation protection activities; Any other practices or circumstances in which people may be exposed to radiation from naturally occurring or artificial sources; The regulation of such facilities and activities; - To assist in the development of the management systems of the relevant regulatory bodies; - By the operator, to specify to a supplier, via contractual documentation, any guidance of this Safety Guide that should be included in the supplier's management system for the supply and delivery of products
Hsu, Yueh-Ling
2004-01-01
The systemic origins of many accidents have led to heightened interest in the way in which organisations identify and manage risks within the airline industry. The activities which are thought to represent the term "organisational accident", "safety culture" and "proactive approach" are documented and seek to explain the fact that airlines differ in their willingness and ability to conduct safety management. However, an important but yet relatively undefined task in the airline...
Inherent and passive safety measures in accelerator driven systems: a safety strategy for ADS
International Nuclear Information System (INIS)
Maschek, W.; Rineiski, A.; Morita, K.; Flad, M.
2001-01-01
The efficiency of Accelerator Driven Systems (ADSs) for the transmutation and incineration of nuclear waste is strongly related to the utilization of so-called dedicated fuels. In the ideal case these fuels should consist of pure TRUs without fertile materials as 238 U or 232 Th to achieve highest incineration/transmutation rates. Dedicated fuels still have to be developed and programs are under way for their fabrication, irradiation and testing. These fertile-free fuels may suffer from deteriorated thermal or thermo-mechanical properties, as a reduced melting point, reduced thermal conductivity or even thermal instability. First analyses have shown that the use of dedicated fuels may lead to a strong deterioration of the safety parameters of the reactor core as e.g. the void worth, the Doppler or the kinetics quantities as neutron generation time and β eff . In addition, a dedicated core may contain multiple ''critical'' fuel masses, resulting in a considerable recriticality potential. Current knowledge on these dedicated fuels suggests that ''critical'' reactors may not be feasible, because of safety reasons. However, for ADSs, the salient hope has been promoted that due to the subcriticality of the system the poor safety features of such fuels could be coped with. Analyses are presented which show potential safety problems for such dedicated cores. Respecting the results of these analyses a safety strategy is proposed along the lines of defense approach in analogy with ideas formerly developed for fast reactors. Inherent and passive safety measures are integrated into the various defense lines. (author)
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
Qualitative safety analysis in accelerator based systems
International Nuclear Information System (INIS)
Sarkar, P.K.; Chowdhury, Lekha M.
2006-01-01
In recent developments connected to high energy and high current accelerators, the accelerator driven systems (ADS) and the Radioactive Ion Beam (RIB) facilities come in the forefront of application. For medical and industrial applications high current accelerators often need to be located in populated areas. These facilities pose significant radiological hazard during their operation and accidental situations. We have done a qualitative evaluation of radiological safety analysis using the probabilistic safety analysis (PSA) methods for accelerator-based systems. The major contribution to hazard comes from a target rupture scenario in both ADS and RIB facilities. Other significant contributors to hazard in the facilities are also discussed using fault tree and event tree methodologies. (author)
The Advanced Light Source (ALS) Radiation Safety System
International Nuclear Information System (INIS)
Ritchie, A.; Oldfather, D.; Lindner, A.
1993-05-01
The Advanced Light Source (ALS) at the Lawrence Berkeley Laboratory (LBL) is a 1.5 GeV synchrotron light source facility consisting of a 120 keV electron gun, 50 MeV linear accelerator, 1.5 Gev booster synchrotron, 200 meter circumference electron storage ring, and many photon beamline transport systems for research. The Radiation Safety System for the ALS has been designed and built with a primary goal of providing protection against inadvertent personnel exposure to gamma and neutron radiation and, secondarily, to enhance the electrical safety of select magnet power supplies
Institutionalization of safety re-assessment system for operating nuclear power plants
International Nuclear Information System (INIS)
Kim, H. J.; Cho, J. C.; Min, B. K.; Park, J. S.; Jung, H. D.; Oh, K. M.; Kim, W. K.; Lim, J. H.
1999-01-01
In this study, in-depth reviews of the foreign countries' experiences and practices in applications of the periodic safety review (PSR), backfitting and license renewal systems as well as the current status of nuclear power safety assurance programs and activities in Korea have been performed to investigate the necessity and feasibility of the application of the systems for the domestic operating nuclear power plants and to establish effective strategy and methodology for the institutionalization of a periodic safety re-assessment system appropriate to both the domestic and international nuclear power environments by incorporating the PSR with the backfitting and license renewal systems. For these purposes, the regulatory policy, fundamental principles and detailed requirements for the institutionalization of the safety re-assessment system and the effective measures for active implementation of the backfitting program have been developed and then a comparative study of benefits and shortcomings has been conducted for the three different models of the periodic safety re-assessment system incorporated with either the license renewal or life extension process, which have been considered as practicable ones in the domestic situation. The model chosen in this study as the most appropriate safety re-assessment system is the one that the re-assessments are performed at the interval of ten years throughout the service life of nuclear power plant and the ten-year license renewal or life extension after the expiration of design life can be permitted based on the regulatory review of the re-assessment results and follow-up measures. Finally, this paper has discussed on the details of the requirements, approach and procedures established for the institutionalization of the periodic safety re-assessment system chosen as the most appropriate one for domestic applications
Safety assessment of a robotic system handling nuclear material
International Nuclear Information System (INIS)
Atcitty, C.B.; Robinson, D.G.
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 at the Department of Energy facility at Pantex by which nuclear material is inspected for weight and leakage. Failure Modes and Effects Analyses were completed for the robotics process to ensure that safety goals for the system had been meet. These analyses showed that the risks to people and the internal and external environment were acceptable
75 FR 15620 - Federal Motor Vehicle Safety Standards; Air Brake Systems
2010-03-30
... fully develop improved brake systems and also to ensure vehicle control and stability while braking... [Docket No. NHTSA 2009-0175] RIN 2127-AK62 Federal Motor Vehicle Safety Standards; Air Brake Systems... Federal motor vehicle safety standard for air brake systems by requiring substantial improvements in...
78 FR 9623 - Federal Motor Vehicle Safety Standards; Air Brake Systems
2013-02-11
... initial speeds, vehicle manufacturers will need to develop unique or complicated braking systems to comply... [Docket No. NHTSA-2013-0011] RIN 2127-AL11 Federal Motor Vehicle Safety Standards; Air Brake Systems... rule that amended the Federal motor vehicle safety standard for air brake systems by requiring...
Analysis of developed transition road safety barrier systems.
Soltani, Mehrtash; Moghaddam, Taher Baghaee; Karim, Mohamed Rehan; Sulong, N H Ramli
2013-10-01
Road safety barriers protect vehicles from roadside hazards by redirecting errant vehicles in a safe manner as well as providing high levels of safety during and after impact. This paper focused on transition safety barrier systems which were located at the point of attachment between a bridge and roadside barriers. The aim of this study was to provide an overview of the behavior of transition systems located at upstream bridge rail with different designs and performance levels. Design factors such as occupant risk and vehicle trajectory for different systems were collected and compared. To achieve this aim a comprehensive database was developed using previous studies. The comparison showed that Test 3-21, which is conducted by impacting a pickup truck with speed of 100 km/h and angle of 25° to transition system, was the most severe test. Occupant impact velocity and ridedown acceleration for heavy vehicles were lower than the amounts for passenger cars and pickup trucks, and in most cases higher occupant lateral impact ridedown acceleration was observed on vehicles subjected to higher levels of damage. The best transition system was selected to give optimum performance which reduced occupant risk factors using the similar crashes in accordance with Test 3-21. Copyright © 2013 Elsevier Ltd. All rights reserved.
Analysing supercritical water reactor's (SCWR's) special safety systems using probabilistic tools
International Nuclear Information System (INIS)
Ituen, I.; Novog, D.R.
2011-01-01
The next generation of reactors, termed Generation IV, has very attractive features -- its superior safety characteristics, high thermal efficiency, and fuel cycle sustainability. A key element of the Generation IV designs is the improvement in safety, which in turn requires improvements in safety system performance and reliability, as well as a reduction in initiating event frequencies. This study compares the response of the systems important to safety in the CANDU-Supercritical Water Reactor to those of the generic CANDU under a main steamline break accident and loss of forced circulation events -- to quantify the improvements in safety for the pre-conceptual CANDU SCWR design. Probabilistic safety analysis is the tool used in this study to test the behavior of the pre- conceptual design during these events. (author)
Lithium-thionyl chloride cell system safety hazard analysis
Dampier, F. W.
1985-03-01
This system safety analysis for the lithium thionyl chloride cell is a critical review of the technical literature pertaining to cell safety and draws conclusions and makes recommendations based on this data. The thermodynamics and kinetics of the electrochemical reactions occurring during discharge are discussed with particular attention given to unstable SOCl2 reduction intermediates. Potentially hazardous reactions between the various cell components and discharge products or impurities that could occur during electrical or thermal abuse are described and the most hazardous conditions and reactions identified. Design factors influencing the safety of Li/SOCl2 cells, shipping and disposal methods and the toxicity of Li/SOCl2 battery components are additional safety issues that are also addressed.
Nuclear power safety reporting system feasibility analysis and concept description
International Nuclear Information System (INIS)
Finlayson, F.C.; Ims, J.R.; Hussman, T.A.
1984-01-01
The Aerospace Corporation is assisting the US Nuclear Regulatory Commission (NRC) in the evaluation of the potential attributes of a voluntary, nonpunitive data gathering system for identifying and quantifying the factors that contribute to the occurrence of significant safety problems involving humans in nuclear power plants. The objectives of the Aerospace Administration (FAA)/National Aeronautics and Space Administration (NASA) Aviation Safety Reporting System (ASRS) in order to determine whether it would be feasible to apply part (or all) of the ASRS concepts for collecting data on human factor related incidents to the nuclear industry; and (2) to identify and define the basic elements and requirements of a Nuclear Power Safety Reporting System (NPSRS), assuming the feasibility of implementing such a system was established
Evaluation and review of the safety management system implementation in the Royal Thai Air Force
Chaiwan, Sakkarin
This study was designed to determine situation and effectiveness of the safety management system currently implemented in the Royal Thai Air Force. Reviewing the ICAO's SMS and the RTAF's SMS was conducted to identify similarities and differences between the two safety management systems. Later, the researcher acquired safety statistics from the RTAF Safety Center to investigate effectiveness of its safety system. The researcher also collected data to identify other factors affecting effectiveness of the safety system during conducting in-depth interviews. Findings and Conclusions: The study shows that the Royal Thai Air Force has never applied the International Civil Aviation Organization's Safety management System to its safety system. However, the RTAF's SMS and the ICAO's SMS have been developed based on the same concepts. These concepts are from Richard H. Woods's book, Aviation safety programs: A management handbook. However, the effectiveness of the Royal Thai Air Force's safety system is in good stance. An accident rate has been decreasing regularly but there are no known factors to describe the increasing rate, according to the participants' opinion. The participants have informed that there are many issues to be resolved to improve the RTAF's safety system. Those issues are cooperation among safety center's staffs, attitude toward safety of the RTAF senior commanders, and safety standards.
Rapid prototyping of the Central Safety System for Nuclear Risk in ITER
Energy Technology Data Exchange (ETDEWEB)
Scibile, L. [ITER Organization, CS 90 046, St. Paul-lez-Durance, Cedex (France); Ambrosino, G. [Consorzio CREATE, Universita degli Studi di Napoli Federico II, via Claudio 21, 80125, Napoli (Italy); De Tommasi, G., E-mail: detommas@unina.i [Consorzio CREATE, Universita degli Studi di Napoli Federico II, via Claudio 21, 80125, Napoli (Italy); Pironti, A. [Consorzio CREATE, Universita degli Studi di Napoli Federico II, via Claudio 21, 80125, Napoli (Italy)
2010-07-15
The Central Safety System for Nuclear Risk (CSS-N) coordinates the safety control systems to ensure nuclear safety for the ITER complex. Since the CSS-N is a safety critical system, its validation and commissioning play a very important role; in particular the required level of reliability must be demonstrated. In such a scenario, it is strongly recommended to use modeling and simulation tools since the early design phase. Indeed, the modeling tools will help in the definition of the control system requirements. Furthermore the models can than be used for the rapid prototyping of the safety system. Hardware-in-the-loop simulations can also be performed in order to assess the performance of the control hardware against a plant simulator. The proposed approach relies on the availability of a plant simulator to develop the prototype of the control system. This paper introduces the methodology used to design and develop both the CSS-N Oriented Plant Simulator and the CSS-N Prototype.
Quality and Safety Education for Nurses (QSEN): The Key is Systems Thinking.
Dolansky, Mary A; Moore, Shirley M
2013-09-30
Over a decade has passed since the Institute of Medicine's reports on the need to improve the American healthcare system, and yet only slight improvement in quality and safety has been reported. The Quality and Safety Education for Nurses (QSEN) initiative was developed to integrate quality and safety competencies into nursing education. The current challenge is for nurses to move beyond the application of QSEN competencies to individual patients and families and incorporate systems thinking in quality and safety education and healthcare delivery. This article provides a history of QSEN and proposes a framework in which systems thinking is a critical aspect in the application of the QSEN competencies. We provide examples of how using this framework expands nursing focus from individual care to care of the system and propose ways to teach and measure systems thinking. The conclusion calls for movement from personal effort and individual care to a focus on care of the system that will accelerate improvement of healthcare quality and safety.
Rapid prototyping of the Central Safety System for Nuclear Risk in ITER
International Nuclear Information System (INIS)
Scibile, L.; Ambrosino, G.; De Tommasi, G.; Pironti, A.
2010-01-01
The Central Safety System for Nuclear Risk (CSS-N) coordinates the safety control systems to ensure nuclear safety for the ITER complex. Since the CSS-N is a safety critical system, its validation and commissioning play a very important role; in particular the required level of reliability must be demonstrated. In such a scenario, it is strongly recommended to use modeling and simulation tools since the early design phase. Indeed, the modeling tools will help in the definition of the control system requirements. Furthermore the models can than be used for the rapid prototyping of the safety system. Hardware-in-the-loop simulations can also be performed in order to assess the performance of the control hardware against a plant simulator. The proposed approach relies on the availability of a plant simulator to develop the prototype of the control system. This paper introduces the methodology used to design and develop both the CSS-N Oriented Plant Simulator and the CSS-N Prototype.
Laboratory safety and the WHO World Alliance for Patient Safety.
McCay, Layla; Lemer, Claire; Wu, Albert W
2009-06-01
Laboratory medicine has been a pioneer in the field of patient safety; indeed, the College of American Pathology first called attention to the issue in 1946. Delivering reliable laboratory results has long been considered a priority, as the data produced in laboratory medicine have the potential to critically influence individual patients' diagnosis and management. Until recently, most attention on laboratory safety has focused on the analytic stage of laboratory medicine. Addressing this stage has led to significant and impressive improvements in the areas over which laboratories have direct control. However, recent data demonstrate that pre- and post-analytical phases are at least as vulnerable to errors; to further improve patient safety in laboratory medicine, attention must now be focused on the pre- and post-analytic phases, and the concept of patient safety as a multi-disciplinary, multi-stage and multi-system concept better understood. The World Alliance for Patient Safety (WAPS) supports improvement of patient safety globally and provides a potential framework for considering the total testing process.
Validation of risk-based performance indicators: Safety system function trends
International Nuclear Information System (INIS)
Boccio, J.L.; Vesely, W.E.; Azarm, M.A.; Carbonaro, J.F.; Usher, J.L.; Oden, N.
1989-10-01
This report describes and applies a process for validating a model for a risk-based performance indicator. The purpose of the risk-based indicator evaluated, Safety System Function Trend (SSFT), is to monitor the unavailability of selected safety systems. Interim validation of this indicator is based on three aspects: a theoretical basis, an empirical basis relying on statistical correlations, and case studies employing 25 plant years of historical data collected from five plants for a number of safety systems. Results using the SSFT model are encouraging. Application of the model through case studies dealing with the performance of important safety systems shows that statistically significant trends in, and levels of, system performance can be discerned which thereby can provide leading indications of degrading and/or improving performances. Methods for developing system performance tolerance bounds are discussed and applied to aid in the interpretation of the trends in this risk-based indicator. Some additional characteristics of the SSFT indicator, learned through the data-collection efforts and subsequent data analyses performed, are also discussed. The usefulness and practicality of other data sources for validation purposes are explored. Further validation of this indicator is noted. Also, additional research is underway in developing a more detailed estimator of system unavailability. 9 refs., 18 figs., 5 tabs
Quality and safety implications of emergency department information systems.
Farley, Heather L; Baumlin, Kevin M; Hamedani, Azita G; Cheung, Dickson S; Edwards, Michael R; Fuller, Drew C; Genes, Nicholas; Griffey, Richard T; Kelly, John J; McClay, James C; Nielson, Jeff; Phelan, Michael P; Shapiro, Jason S; Stone-Griffith, Suzanne; Pines, Jesse M
2013-10-01
The Health Information Technology for Economic and Clinical Health Act of 2009 and the Centers for Medicare & Medicaid Services "meaningful use" incentive programs, in tandem with the boundless additional requirements for detailed reporting of quality metrics, have galvanized hospital efforts to implement hospital-based electronic health records. As such, emergency department information systems (EDISs) are an important and unique component of most hospitals' electronic health records. System functionality varies greatly and affects physician decisionmaking, clinician workflow, communication, and, ultimately, the overall quality of care and patient safety. This article is a joint effort by members of the Quality Improvement and Patient Safety Section and the Informatics Section of the American College of Emergency Physicians. The aim of this effort is to examine the benefits and potential threats to quality and patient safety that could result from the choice of a particular EDIS, its implementation and optimization, and the hospital's or physician group's approach to continuous improvement of the EDIS. Specifically, we explored the following areas of potential EDIS safety concerns: communication failure, wrong order-wrong patient errors, poor data display, and alert fatigue. Case studies are presented that illustrate the potential harm that could befall patients from an inferior EDIS product or suboptimal execution of such a product in the clinical environment. The authors have developed 7 recommendations to improve patient safety with respect to the deployment of EDISs. These include ensuring that emergency providers actively participate in selection of the EDIS product, in the design of processes related to EDIS implementation and optimization, and in the monitoring of the system's ongoing success or failure. Our recommendations apply to emergency departments using any type of EDIS: custom-developed systems, best-of-breed vendor systems, or enterprise systems
Context-aware system for pre-triggering irreversible vehicle safety actuators.
Böhmländer, Dennis; Dirndorfer, Tobias; Al-Bayatti, Ali H; Brandmeier, Thomas
2017-06-01
New vehicle safety systems have led to a steady improvement of road safety and a reduction in the risk of suffering a major injury in vehicle accidents. A huge leap forward in the development of new vehicle safety systems are actuators that have to be activated irreversibly shortly before a collision in order to mitigate accident consequences. The triggering decision has to be based on measurements of exteroceptive sensors currently used in driver assistance systems. This paper focuses on developing a novel context-aware system designed to detect potential collisions and to trigger safety actuators even before an accident occurs. In this context, the analysis examines the information that can be collected from exteroceptive sensors (pre-crash data) to predict a certain collision and its severity to decide whether a triggering is entitled or not. A five-layer context-aware architecture is presented, that is able to collect contextual information about the vehicle environment and the actual driving state using different sensors, to perform reasoning about potential collisions, and to trigger safety functions upon that information. Accident analysis is used in a data model to represent uncertain knowledge and to perform reasoning. A simulation concept based on real accident data is introduced to evaluate the presented system concept. Copyright © 2017 Elsevier Ltd. All rights reserved.