Development of integrated D/B system for the safety-related structures in nuclear power plant

International Nuclear Information System (INIS)

Cho, M. S.; Song, Y. C.; Lee, J. S.; Choi, W. S.

2002-01-01

The integrated D/B system is developed for digitalizing the history of the safety-related structures of nuclear power plant. It have 5 database which are consist of Generals, Structural and Design, Materials, Construction, Aging and repair information D/B. For efficient operation of the system, we are to set up the outline of the system, find out data field for target structures, and develop utilities. Utilities will be the aging and repair data management program, the close examination management program, the data search engine with various options which help users to find the information quickly, and the data management program restoring, updating and exchanging input data. Development of the integrated D/B system of the safety-related structures will contribute to management of the structures of nuclear power plant with advanced technology

Introduction of structural health and safety monitoring warning systems for Shenzhen-Hong Kong Western Corridor Shenzhen Bay Bridge

Science.gov (United States)

Li, N.; Zhang, X. Y.; Zhou, X. T.; Leng, J.; Liang, Z.; Zheng, C.; Sun, X. F.

2008-03-01

Though the brief introduction of the completed structural health and safety monitoring warning systems for Shenzhen-Hongkong western corridor Shenzhen bay highway bridge (SZBHMS), the self-developed system frame, hardware and software scheme of this practical research project are systematically discussed in this paper. The data acquisition and transmission hardware and the basic software based on the NI (National Instruments) Company virtual instruments technology were selected in this system, which adopted GPS time service receiver technology and so on. The objectives are to establish the structural safety monitoring and status evaluation system to monitor the structural responses and working conditions in real time and to analyze the structural working statue using information obtained from the measured data. It will be also provided the scientific decision-making bases for the bridge management and maintenance. Potential technical approaches to the structural safety warning systems, status identification and evaluation method are presented. The result indicated that the performance of the system has achieved the desired objectives, ensure the longterm high reliability, real time concurrence and advanced technology of SZBHMS. The innovate achievement which is the first time to implement in domestic, provide the reference for long-span bridge structural health and safety monitoring warning systems design.

Safety significance of steam generator tube degradation mechanisms

Energy Technology Data Exchange (ETDEWEB)

Roussel, G; Mignot, P [AIB-Vincotte Nuclear - AVN, Brussels (Belgium)

1991-07-01

Steam generator (SG) tube bundle is a part of the Reactor Coolant Pressure Boundary (RCPB): this means that its integrity must be maintained. However, operating experience shows various types of tube degradation to occur in the SG tubing, which may lead to SG tube leaks or SG tube ruptures and create a loss of primary system coolant through the SG, therefore providing a direct path to the environment outside the primary containment structure. In this paper, the major types of known SG tube degradations are described and analyzed in order to assess their safety significance with regard to SG tube integrity. In conclusion: The operational reliability and the safety of the PWR steam generator s requires a sufficient knowledge of the degradation mechanisms to determine the amount of degradation that a tube can withstand and the time that it may remain in operation. They also require the availability of inspection techniques to accurately detect and characterize the various degradations. The status of understanding of the major types of degradation summarized in this paper shows and justifies why efforts are being performed to improve the management of the steam generator tube defects.

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

Audit Report The Procurement of Safety Class/Safety-Significant Items at the Savannah River Site

International Nuclear Information System (INIS)

2009-01-01

The Department of Energy operates several nuclear facilities at its Savannah River Site, and several additional facilities are under construction. This includes the National Nuclear Security Administration's Tritium Extraction Facility (TEF) which is designated to help maintain the reliability of the U.S. nuclear stockpile. The Mixed Oxide Fuel Fabrication Facility (MOX Facility) is being constructed to manufacture commercial nuclear reactor fuel assemblies from weapon-grade plutonium oxide and depleted uranium. The Interim Salt Processing (ISP) project, managed by the Office of Environmental Management, will treat radioactive waste. The Department has committed to procuring products and services for nuclear-related activities that meet or exceed recognized quality assurance standards. Such standards help to ensure the safety and performance of these facilities. To that end, it issued Departmental Order 414.1C, Quality Assurance (QA Order). The QA Order requires the application of Quality Assurance Requirements for Nuclear Facility Applications (NQA-1) for nuclear-related activities. The NQA-1 standard provides requirements and guidelines for the establishment and execution of quality assurance programs during the siting, design, construction, operation, and decommissioning of nuclear facilities. These requirements, promulgated by the American Society of Mechanical Engineers, must be applied to 'safety-class' and 'safety-significant' structures, systems and components (SSCs). Safety-class SSCs are defined as those necessary to prevent exposure off site and to protect the public. Safety-significant SSCs are those whose failure could irreversibly impact worker safety such as a fatality, serious injury, or significant radiological or chemical exposure. Due to the importance of protecting the public, workers, and environment, we initiated an audit to determine whether the Department of Energy procured safety-class and safety-significant SSCs that met NQA-1 standards at

Human and organization factors: engineering operating safety into offshore structures

International Nuclear Information System (INIS)

Bea, Robert G.

1998-01-01

History indicates clearly that the safety of offshore structures is determined primarily by the humans and organizations responsible for these structures during their design, construction, operation, maintenance, and decommissioning. If the safety of offshore structures is to be preserved and improved, then attention of engineers should focus on to how to improve the reliability of the offshore structure 'system,' including the people that come into contact with the structure during its life-cycle. This article reviews and discusss concepts and engineering approaches that can be used in such efforts. Two specific human factor issues are addressed: (1) real-time management of safety during operations, and (2) development of a Safety Management Assessment System to help improve the safety of offshore structures

Keeping patients safe in healthcare organizations: a structuration theory of safety culture.

Science.gov (United States)

Groves, Patricia S; Meisenbach, Rebecca J; Scott-Cawiezell, Jill

2011-08-01

This paper presents a discussion of the use of structuration theory to facilitate understanding and improvement of safety culture in healthcare organizations. Patient safety in healthcare organizations is an important problem worldwide. Safety culture has been proposed as a means to keep patients safe. However, lack of appropriate theory limits understanding and improvement of safety culture. The proposed structuration theory of safety culture was based on a critique of available English-language literature, resulting in literature published from 1983 to mid-2009. CINAHL, Communication and Mass Media Complete, ABI/Inform and Google Scholar databases were searched using the following terms: nursing, safety, organizational culture and safety culture. When viewed through the lens of structuration theory, safety culture is a system involving both individual actions and organizational structures. Healthcare organization members, particularly nurses, share these values through communication and enact them in practice, (re)producing an organizational safety culture system that reciprocally constrains and enables the actions of the members in terms of patient safety. This structurational viewpoint illuminates multiple opportunities for safety culture improvement. Nurse leaders should be cognizant of competing value-based culture systems in the organization and attend to nursing agency and all forms of communication when attempting to create or strengthen a safety culture. Applying structuration theory to the concept of safety culture reveals a dynamic system of individual action and organizational structure constraining and enabling safety practice. Nurses are central to the (re)production of this safety culture system. © 2011 Blackwell Publishing Ltd.

Design of concrete structures important to safety of nuclear facilities

International Nuclear Information System (INIS)

2001-10-01

Civil engineering structures in nuclear installations form an important feature having implications to safety performance of these installations. The objective and minimum requirements for the design of civil engineering buildings/structures to be fulfilled to provide adequate assurance for safety of nuclear installations in India (such as pressurised heavy water reactor and related systems) are specified in the Safety standard for civil engineering structures important to safety of nuclear facilities. This standard is written by AERB to specify guidelines for implementation of the above civil engineering safety standard in the design of concrete structures important to safety

System analysis of vehicle active safety problem

Science.gov (United States)

Buznikov, S. E.

2018-02-01

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

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.

10 CFR 50.69 - Risk-informed categorization and treatment of structures, systems and components for nuclear...

Science.gov (United States)

2010-01-01

..., systems and components for nuclear power reactors. (a) Definitions. Risk-Informed Safety Class (RISC)-1... functions. Risk-Informed Safety Class (RISC)-2 structures, systems and components (SSCs) means nonsafety-related SSCs that perform safety significant functions. Risk-Informed Safety Class (RISC)-3 structures...

Structural analysis of a rehabilitative training system based on a ceiling rail for safety of hemiplegia patients.

Science.gov (United States)

Kim, Kyong; Song, Won Kyung; Chong, Woo Suk; Yu, Chang Ho

2018-04-17

The body-weight support (BWS) function, which helps to decrease load stresses on a user, is an effective tool for gait and balance rehabilitation training for elderly people with weakened lower-extremity muscular strength, hemiplegic patients, etc. This study conducts structural analysis to secure user safety in order to develop a rail-type gait and balance rehabilitation training system (RRTS). The RRTS comprises a rail, trolley, and brain-machine interface. The rail (platform) is connected to the ceiling structure, bearing the loads of the RRTS and of the user and allowing locomobility. The trolley consists of a smart drive unit (SDU) that assists the user with forward and backward mobility and a body-weight support (BWS) unit that helps the user to control his/her body-weight load, depending on the severity of his/her hemiplegia. The brain-machine interface estimates and measures on a real-time basis the body-weight (load) of the user and the intended direction of his/her movement. Considering the weight of the system and the user, the mechanical safety performance of the system frame under an applied 250-kg static load is verified through structural analysis using ABAQUS (6.14-3) software. The maximum stresses applied on the rail and trolley under the given gravity load of 250 kg, respectively, are 18.52 MPa and 48.44 MPa. The respective safety factors are computed to be 7.83 and 5.26, confirming the RRTS's mechanical safety. An RRTS with verified structural safety could be utilized for gait movement and balance rehabilitation and training for patients with hemiplegia.

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)

Categorization of safety related motor operated valve safety significance for Ulchin Unit 3

International Nuclear Information System (INIS)

Kang, D. I.; Kim, K. Y.

2002-03-01

We performed a categorization of safety related Motor Operated Valve (MOV) safety significance for Ulchin Unit 3. The safety evaluation of MOV of domestic nuclear power plants affects the generic data used for the quantification of MOV common cause failure ( CCF) events in Ulchin Units 3 PSA. Therefore, in this study, we re-estimated the MGL(Multiple Greek Letter) parameter used for the evaluation of MOV CCF probabilities in Ulchin Units 3 Probabilistic Safety Assessment (PSA) and performed a classification of the MOV safety significance. The re-estimation results of the MGL parameter show that its value is decreased by 30% compared with the current value in Ulchin Unit 3 PSA. The categorization results of MOV safety significance using the changed value of MGL parameter shows that the number of HSSCs(High Safety Significant Components) is decreased by 54.5% compared with those using the current value of it in Ulchin Units 3 PSA

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

Science.gov (United States)

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

2016-11-01

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

A quantitative approach for risk-informed safety significance categorization in option-2

International Nuclear Information System (INIS)

Ha, Jun Su; Seong, Poong Hyun

2004-01-01

OPTION-2 recommends that Structures, Systems, or Components (SSCs) of Nuclear Power Plants (NPPs) should be categorized into four groups according to their safety significance as well as whether they are safety-related or not. With changes to the scope of SSCs covered by 10 CFR 50, safety-related components which categorized into low safety significant SSC (RISC-3 SSC) can be exempted from the existing conservative burden (or requirements). As OPTION-2 paradigm is applied, a lot of SSCs may be categorized into RISC-3 SSCs. Changes in treatment of the RISC-3 SSCs will be recommended and then finally the recommended changes shall be evaluated. Consequently, before recommending the changes in treatment, probable candidate SSCs for the changes in treatment need to be identified for efficient risk-informed regulation and application (RIRA). Hence, in this work, a validation focused on the RISC-3 SSCs is proposed to identify probable candidate SSCs. Burden to Importance Ratio (BIR) is utilized as a quantitative measure for the validation. BIR is a measure representing the extent of resources or requirements imposed on a SSC with respect to the value of the importance measure of the SSC. Therefore SSCs having high BIR can be considered as probable candidate SSCs for the changes in treatment. In addition, the final decision whether RISC-3 SSCs can be considered as probable candidate SSCs or not should be made by an expert panel. For the effective decision making, a structured mathematical decision-making process is constructed based on Belief Networks (BBN) to overcome demerits of conventional group meeting based on unstructured discussion for decision-making. To demonstrate the usefulness of the proposed approach, the approach is applied to 22 components selected from 512 In-Service Test (IST) components of Ulchin unit 3. The results of the application show that the proposed approach can identify probable candidate SSCs for changes in treatment. The identification of the

Safety of huge systems

International Nuclear Information System (INIS)

Kondo, Jiro.

1995-01-01

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

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

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.

Comparing performance level estimation of safety functions in three distributed structures

International Nuclear Information System (INIS)

Hietikko, Marita; Malm, Timo; Saha, Heikki

2015-01-01

The capability of a machine control system to perform a safety function is expressed using performance levels (PL). This paper presents the results of a study where PL estimation was carried out for a safety function implemented using three different distributed control system structures. Challenges relating to the process of estimating PLs for safety related distributed machine control functions are highlighted. One of these examines the use of different cabling schemes in the implementation of a safety function and its effect on the PL evaluation. The safety function used as a generic example in PL calculations relates to a mobile work machine. It is a safety stop function where different technologies (electrical, hydraulic and pneumatic) can be utilized. It was detected that by replacing analogue cables with digital communication the system structure becomes simpler with less number of failing components, which can better the PL of the safety function. - Highlights: • Integration in distributed systems enables systems with less components. • It offers high reliability and diagnostic properties. • Analogue signals create uncertainty in signal reliability and difficult diagnostics

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

A method for risk-informed safety significance categorization using the analytic hierarchy process and bayesian belief networks

International Nuclear Information System (INIS)

Ha, Jun Su; Seong, Poong Hyun

2004-01-01

A risk-informed safety significance categorization (RISSC) is to categorize structures, systems, or components (SSCs) of a nuclear power plant (NPP) into two or more groups, according to their safety significance using both probabilistic and deterministic insights. In the conventional methods for the RISSC, the SSCs are quantitatively categorized according to their importance measures for the initial categorization. The final decisions (categorizations) of SSCs, however, are qualitatively made by an expert panel through discussions and adjustments of opinions by using the probabilistic insights compiled in the initial categorization process and combining the probabilistic insights with the deterministic insights. Therefore, owing to the qualitative and linear decision-making process, the conventional methods have the demerits as follows: (1) they are very costly in terms of time and labor, (2) it is not easy to reach the final decision, when the opinions of the experts are in conflict and (3) they have an overlapping process due to the linear paradigm (the categorization is performed twice - first, by the engineers who propose the method, and second, by the expert panel). In this work, a method for RISSC using the analytic hierarchy process (AHP) and bayesian belief networks (BBN) is proposed to overcome the demerits of the conventional methods and to effectively arrive at a final decision (or categorization). By using the AHP and BBN, the expert panel takes part in the early stage of the categorization (that is, the quantification process) and the safety significance based on both probabilistic and deterministic insights is quantified. According to that safety significance, SSCs are quantitatively categorized into three categories such as high safety significant category (Hi), potentially safety significant category (Po), or low safety significant category (Lo). The proposed method was applied to the components such as CC-V073, CV-V530, and SI-V644 in Ulchin Unit

Seismic Safety Margins Research Program. Phase 1. Project V. Structural sub-system response: subsystem response review

International Nuclear Information System (INIS)

Fogelquist, J.; Kaul, M.K.; Koppe, R.; Tagart, S.W. Jr.; Thailer, H.; Uffer, R.

1980-03-01

This project is directed toward a portion of the Seismic Safety Margins Research Program which includes one link in the seismic methodology chain. The link addressed here is the structural subsystem dynamic response which consists of those components and systems whose behavior is often determined decoupled from the major structural response. Typically the mathematical model utilized for the major structural response will include only the mass effects of the subsystem and the main model is used to produce the support motion inputs for subsystem seismic qualification. The main questions addressed in this report have to do with the seismic response uncertainty of safety-related components or equipment whose seismic qualification is performed by (a) analysis, (b) tests, or (c) combinations of analysis and tests, and where the seismic input is assumed to have no uncertainty

The Evolution of System Safety at NASA

Science.gov (United States)

Dezfuli, Homayoon; Everett, Chris; Groen, Frank

2014-01-01

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

Identification of structures, systems, and components important to safety at the potential repository at Yucca Mountain

International Nuclear Information System (INIS)

Hartman, D.J.; Miller, D.D.; Klamerus, L.J.

1991-10-01

This study recommends which structures, systems, and components of the potential repository at Yucca Mountain are important to safety. The assessment was completed in April 1990 and uses the reference repository configuration in the Site Characterization Plan Conceptual Design Report and follows the methodology required at that time by DOE Procedure AP6.10-Q. Failures of repository items during the preclosure period are evaluated to determine the potential offsite radiation doses and associated probabilities. Items are important to safety if, in the event they fail to perform their intended function, an accident could result which causes a dose commitment greater than 0.5 rem to the whole body or any organ of an individual in an unrestricted area. This study recommends that these repository items include the structures that house spent fuel and high-level waste, the associated filtered ventilation exhaust systems, certain waste- handling equipment, the waste containers, the waste treatment building structure, the underground waste transporters, and other items listed in this report. This work was completed April 1990. 27 refs., 7 figs., 9 tabs

Food safety performance indicators to benchmark food safety output of food safety management systems.

Science.gov (United States)

Jacxsens, L; Uyttendaele, M; Devlieghere, F; Rovira, J; Gomez, S Oses; Luning, P A

2010-07-31

There is a need to measure the food safety performance in the agri-food chain without performing actual microbiological analysis. A food safety performance diagnosis, based on seven indicators and corresponding assessment grids have been developed and validated in nine European food businesses. Validation was conducted on the basis of an extensive microbiological assessment scheme (MAS). The assumption behind the food safety performance diagnosis is that food businesses which evaluate the performance of their food safety management system in a more structured way and according to very strict and specific criteria will have a better insight in their actual microbiological food safety performance, because food safety problems will be more systematically detected. The diagnosis can be a useful tool to have a first indication about the microbiological performance of a food safety management system present in a food business. Moreover, the diagnosis can be used in quantitative studies to get insight in the effect of interventions on sector or governmental level. Copyright 2010 Elsevier B.V. All rights reserved.

Improved safety of the system 80+TM standard plants design through increased diversity and redundancy of safety systems

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

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)

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.

Design, fabrication and erection of steel structures important to safety of nuclear facilities

International Nuclear Information System (INIS)

2001-10-01

Civil engineering structures in nuclear installations form an important feature having implications to safety performance of these installations. The objective and minimum requirements for the design of civil engineering buildings/structures to be fulfilled to provide adequate assurance for safety of nuclear installations in India (such as pressurised heavy water reactor and related systems) are specified in the Safety Standard for Civil Engineering Structures Important to Safety of Nuclear Facilities. This standard is written by AERB to specify guidelines for implementation of the above civil engineering safety standard in the design, fabrication and erection of steel structures important to safety

Reliability Analysis and Calibration of Partial Safety Factors for Redundant Structures

DEFF Research Database (Denmark)

Sørensen, John Dalsgaard

1998-01-01

Redundancy is important to include in the design and analysis of structural systems. In most codes of practice redundancy is not directly taken into account. In the paper various definitions of a deterministic and reliability based redundancy measure are reviewed. It is described how reundancy can...... be included in the safety system and how partial safety factors can be calibrated. An example is presented illustrating how redundancy is taken into account in the safety system in e.g. the Danish codes. The example shows how partial safety factors can be calibrated to comply with the safety level...

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

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

Development of Applicabe Methods for Evaluating the Safety of Offsshore Structures

DEFF Research Database (Denmark)

Sørensen, John Dalsgaard; Thoft-Christensen, Palle; Sigurdsson, Gudfinnur

] was published in January 1985 and describes the computer programme package UNZIP by which the reliability of 3-dimensional structures can be evaluated. The present paper describes progress made during the first four months of 1985. A new program package called MEKBETA is described. By this program fundamental......The research project "Development of applicable methods for evaluating the safety of offshore structures" was initiated in August 1984. It is in 1985 supported by Danish Ministry of Energy with D. kr. 380.000. The results obtained will be published in a series of papers. The first paper [1...... mechanisms can be identified automatically. Also significant mechanisms in plane and space frame and lattice structures are identified automatically. Finally, the reliability of the structural system is estimated on the basis of the identified significant mechanisms. The automatic generation of fundamental...

SAFETY CRITERION IN ASSESSING THE IMPORTANCE OF AN ELEMENT IN THE COMPLEX TECHNOLOGICAL SYSTEM RELIABILITY STRUCTURE

Directory of Open Access Journals (Sweden)

Leszek CHYBOWSKI

2012-01-01

Full Text Available The paper presents the need to develop a description of the importance of the technological systems reliability structure elements in terms of security of the system. Basic issues related to the exploration of weak links and important elements in the system as well as a proposal to develop the current approach to assessing the importance of the system components have been presented. Moreover, the differences between the unreliability of suitability and unreliability of safety have been pointed out.

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

An intelligent hybrid system for surface coal mine safety analysis

Energy Technology Data Exchange (ETDEWEB)

Lilic, N.; Obradovic, I.; Cvjetic, A. [University of Belgrade, Belgrade (Serbia)

2010-06-15

Analysis of safety in surface coal mines represents a very complex process. Published studies on mine safety analysis are usually based on research related to accidents statistics and hazard identification with risk assessment within the mining industry. Discussion in this paper is focused on the application of AI methods in the analysis of safety in mining environment. Complexity of the subject matter requires a high level of expert knowledge and great experience. The solution was found in the creation of a hybrid system PROTECTOR, whose knowledge base represents a formalization of the expert knowledge in the mine safety field. The main goal of the system is the estimation of mining environment as one of the significant components of general safety state in a mine. This global goal is subdivided into a hierarchical structure of subgoals where each subgoal can be viewed as the estimation of a set of parameters (gas, dust, climate, noise, vibration, illumination, geotechnical hazard) which determine the general mine safety state and category of hazard in mining environment. Both the hybrid nature of the system and the possibilities it offers are illustrated through a case study using field data related to an existing Serbian surface coal mine.

Concrete structures. Contribution to the safety assessment of existing structures

Directory of Open Access Journals (Sweden)

D. COUTO

Full Text Available The safety evaluation of an existing concrete structure differs from the design of new structures. The partial safety factors for actions and resistances adopted in the design phase consider uncertainties and inaccuracies related to the building processes of structures, variability of materials strength and numerical approximations of the calculation and design processes. However, when analyzing a finished structure, a large number of unknown factors during the design stage are already defined and can be measured, which justifies a change in the increasing factors of the actions or reduction factors of resistances. Therefore, it is understood that safety assessment in existing structures is more complex than introducing security when designing a new structure, because it requires inspection, testing, analysis and careful diagnose. Strong knowledge and security concepts in structural engineering are needed, as well as knowledge about the materials of construction employed, in order to identify, control and properly consider the variability of actions and resistances in the structure. With the intention of discussing this topic considered complex and diffuse, this paper presents an introduction to the safety of concrete structures, a synthesis of the recommended procedures by Brazilian standards and another codes, associated with the topic, as well a realistic example of the safety assessment of an existing structure.

A patient safety objective structured clinical examination.

Science.gov (United States)

Singh, Ranjit; Singh, Ashok; Fish, Reva; McLean, Don; Anderson, Diana R; Singh, Gurdev

2009-06-01

There are international calls for improving education for health care workers around certain core competencies, of which patient safety and quality are integral and transcendent parts. Although relevant teaching programs have been developed, little is known about how best to assess their effectiveness. The objective of this work was to develop and implement an objective structured clinical examination (OSCE) to evaluate the impact of a patient safety curriculum. The curriculum was implemented in a family medicine residency program with 47 trainees. Two years after commencing the curriculum, a patient safety OSCE was developed and administered at this program and, for comparison purposes, to incoming residents at the same program and to residents at a neighboring residency program. All 47 residents exposed to the training, all 16 incoming residents, and 10 of 12 residents at the neighboring program participated in the OSCE. In a standardized patient case, error detection and error disclosure skills were better among trained residents. In a chart-based case, trained residents showed better performance in identifying deficiencies in care and described more appropriate means of addressing them. Third year residents exposed to a "Systems Approach" course performed better at system analysis and identifying system-based solutions after the course than before. Results suggest increased systems thinking and inculcation of a culture of safety among residents exposed to a patient safety curriculum. The main weaknesses of the study are its small size and suboptimal design. Much further investigation is needed into the effectiveness of patient safety curricula.

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.

Structural observation of long-span suspension bridges for safety assessment: implementation of an optical displacement measurement system

International Nuclear Information System (INIS)

Martins, L Lages; Ribeiro, A Silva; Rebordão, J M

2015-01-01

This paper addresses the implementation of an optical displacement measurement system in the observation scenario of a long-span suspension bridge and its contribution for structural safety assessment. The metrological background required for quality assurance of the measurements is described, namely, the system's intrinsic parameterization and integration in the SI dimensional traceability chain by calibration, including its measurement uncertainty assessment

System safety education focused on flight safety

Science.gov (United States)

Holt, E.

1971-01-01

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

System theory and safety models in Swedish, UK, Dutch and Australian road safety strategies.

Science.gov (United States)

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.

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)

Significance of FIZ Technik Databases in nuclear safety and environmental protection

International Nuclear Information System (INIS)

Das, N.K.

1993-01-01

The language of the abstracts of the FIZ Technik databases is primarly German (e.g. DOMA 80%, SDIM 70%). Furthermore FIZ Technik offers licence databases on engineering and technology, management, manufacturers, products, contacts, standards and specifications, geosciences and natural resources. The contents and structure of the databases are described in the FIZ Technik bluesheets and the database news. With some examples the significance of the FIZ Technik databases DOMA, ZDEE, SDIM, SILI and MEDI in nuclear safety and environmental protection is shown. (orig.)

Fire safety assessment of tunnel structures

DEFF Research Database (Denmark)

Gkoumas, Konstantinos; Giuliani, Luisa; Petrini, Francesco

2011-01-01

.g. structural and non structural, organizational, human behavior). This is even more truth for the fire safety design of such structures. Fire safety in tunnels is challenging because of the particular environment, bearing in mind also that a fire can occur in different phases of the tunnel’s lifecycle. Plans...... for upgrading fire safety provisions and tunnel management are also important for existing tunnels. In this study, following a brief introduction of issues regarding the above mentioned aspects, the structural performance of a steel rib for a tunnel infrastructure subject to fire is assessed by means...

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

Science.gov (United States)

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.

Resistance ability evaluation of safety-related structures for the simulated aircraft accident

Energy Technology Data Exchange (ETDEWEB)

Kim, Young Jin; Kim, Sung Woon; Choi, Jang Kyu [Daewoo E and C Co., Ltd., Suwon (Korea, Republic of)] (and others)

2003-03-15

Aircraft accidents on nuclear safety-related structures can cause severe damage to the safety of NPP(Nuclear Power Plant)s. To assess the safety of nuclear safety-related structures, the local damage and the dynamic response of global structures should be investigated together. This study have compared several local damage assessment formulas suggested for aircraft as an impactor, and have set the assessment system of local damage for impact-proof design of NPP containment buildings. And the local damage of nuclear safety-related structures in operation in Korea for commercial aircraft as impactor have been estimated. Impact load-time functions of the aircraft crash have been decided to assessment the safety of nuclear safety-related structures against the intentional colliding of commercial aircraft. Boeing 747 and Boeing 767 is selected as target aircraft based on the operation frequencies and weights. Comparison of the fire analysis methods showed that the method considering heat convection and radiation is adequate for the temperature analysis of the aircraft fuel fire. Finally, the study covered the analysis of the major structural drawings and design drawings with which three-dimensional finite element model analysis is expected to be performed.

System reliability developments in structural engineering

International Nuclear Information System (INIS)

Moses, F.

1982-01-01

Two major limitations occur in present structural design code developments utilizing reliability theory. The notional system reliabilities may differ significantly from calibrated component reliabilities. Secondly, actual failures are often due to gross errors not reflected in most present code formats. A review is presented of system reliability methods and further new concepts are developed. The incremental load approach for identifying and expressing collapse modes is expanded by employing a strategy to identify and enumerate the significant structural collapse modes. It further isolates the importance of critical components in the system performance. Ductile and brittle component behavior and strength correlation is reflected in the system model and illustrated in several examples. Modal combinations for the system reliability are also reviewed. From these developments a system factor can be addended to component safety checking equations. Values may be derived from system behavior by substituting in a damage model which accounts for the response range from component failure to collapse. Other strategies are discussed which emphasize quality assurance during design and in-service inspection for components whose behavior is critical to the system reliability. (Auth.)

Urban street structure and traffic safety.

Science.gov (United States)

Mohan, Dinesh; Bangdiwala, Shrikant I; Villaveces, Andres

2017-09-01

This paper reports the influence of road type and junction density on road traffic fatality rates in U.S. cities. The Fatality Analysis Reporting System (FARS) files were used to obtain fatality rates for all cities for the years 2005-2010. A stratified random sample of 16 U.S. cities was taken, and cities with high and low road traffic fatality rates were compared on their road layout details (TIGER maps were used). Statistical analysis was done to determine the effect of junction density and road type on road traffic fatality rates. The analysis of road network and road traffic crash fatality rates in these randomly selected U.S. cities shows that, (a) higher number of junctions per road length was significantly associated with a lower motor- vehicle crash and pedestrian mortality rates, and, (b) increased number of kilometers of roads of any kind was associated with higher fatality rates, but an additional kilometer of main arterial road was associated with a significantly higher increase in total fatalities. When compared to non-arterial roads, the higher the ratio of highways and main arterial roads, there was an association with higher fatality rates. These results have important implications for road safety professionals. They suggest that once the road and street structure is put in place, that will influence whether a city has low or high traffic fatality rates. A city with higher proportion of wider roads and large city blocks will tend to have higher traffic fatality rates, and therefore in turn require much more efforts in police enforcement and other road safety measures. Urban planners need to know that smaller block size with relatively less wide roads will result in lower traffic fatality rates and this needs to be incorporated at the planning stage. Copyright © 2017 National Safety Council and 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

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

International Nuclear Information System (INIS)

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

2016-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-10-15

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

Public safety risk management at socio-economic and / or historic-cultural significant dam sites

Energy Technology Data Exchange (ETDEWEB)

Earle, Gordon D.; Ryan, Katherine; Pyykonen, Nicole K.; Pitts, Lucas [Otonabee Region Conservation Authority, Peterborough, (Canada)

2010-07-01

The Lang Dam and adjoining gristmill, located near Peterborough are integral parts of the Lang Pioneer Village museum. Activities occurring within close proximity to the dam have led to safety issues. The owner (ORCA) has developed and implemented public safety management plans (PSMPs) for each of its water control structures, including the Lang Dam. ORCA gave special attention to the social, economic, aesthetic, historic and cultural dimensions associated the implementation of public safety management plans. These factors play a significant role in how well public safety measures (PSMs) are received by stakeholder groups and the general public. This paper reported the challenges of developing and implementing a PSMP for the Lang Dam, with the focus on property site-specific PSMS while preserving socio-economic and historic-cultural character and values. It was demonstrated that the dam owners, regulatory authorities, control agencies and preservationists need to come together to develop a holistic public safety management process.

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

The application of new mathematical structures to safety analysis

International Nuclear Information System (INIS)

Cooper, J.A.; Ross, T.J.

1997-10-01

Probabilistic safety analyses (PSAs) often depend on significant subjectivity. The recent successes of fuzzy logic and fuzzy and hybrid mathematics in portraying subjectivity is a reminder that a selection made from the most applicable mathematical tools is more important than forced adaptation of conventional tools. In this paper, the authors consider new approaches that enhance conventional and fuzzy PSA by improved handling of subjectivity. The most significant of the mathematical structures were have investigated (from a standpoint of safety analysis applications) will be described, and the general types of applications will be outlined

Time-space structure of nuclear safety

International Nuclear Information System (INIS)

Miya, Kenzo

2003-01-01

New idea to analyze the structure of nuclear safety and to investigate functioning property of hierarchical principle is applied to nuclear safety in this paper. The nuclear safety is expressed by three principles such as 1) the action and subject are partitioned and classified by time and space, 2) introduction of hierarchy with three strata to the closed object and hierarchy with many strata to the open object and 3) application of 'element, relation and abstraction' to the engineering system as a framework of intellectual activity. For example, prevention of core melt is the closed object and it is obtained by acting hierarchies with three strata (operation stop, cooling and closing radiation) as the safety functions. Prevention of increase of accident is open object, so that, space hierarchy with many strata of prevention is used for the safety security of reactor. The safety security method of reactor consists of three processes, that is 1) the basic process to make clear the continuous operating time on the basis of regular inspection, 2) the action process of operating ECCS to prevent core damage accident, when a large leakage happens and 3) many strata prevention process of stopping a leak in the environment. (S.Y.)

Software Safety Risk in Legacy Safety-Critical Computer Systems

Science.gov (United States)

Hill, Janice L.; Baggs, Rhoda

2007-01-01

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

Design characteristics of safety parameter display system for nuclear power plants

International Nuclear Information System (INIS)

Zhang Yuangfang

1992-02-01

The design features of safety parameter display system (SPDS) developed by Tsinghua University is introduced. Some new features have been added into the system functions and they are: (1) hierarchical display structure; (2) human factor in the display format design; (3)automatic diagnosis of safety status of nuclear power plant; (4) extension of SPDS use scope; (5) flexible hardware structure. The new approaches in the design are: (1)adopting the international design standards; (2) selecting safety parameters strictly; (3) developing software under multitask operating system; (4) using a nuclear power plant simulator to verify the SPDS design

Fire Safety Design of Wood Structures

DEFF Research Database (Denmark)

Hertz, Kristian Dahl

2006-01-01

Lecture Notes on Fire Safety Design of Wood Structures including charring of wood and load bearing capacity of beams, columns, and connections.......Lecture Notes on Fire Safety Design of Wood Structures including charring of wood and load bearing capacity of beams, columns, and connections....

Method to classify the safety class of Structure, System and Components in a Defueled Condition of Nuclear Power Plant

Energy Technology Data Exchange (ETDEWEB)

Kim, Dong-Hak; Jeon, Dang-Hee [KHNP CRI, Daejeon (Korea, Republic of)

2016-10-15

During pre-decommissioning phase, licensing and engineering work need to change the design basis of the plant such as safety analysis report, downgrade of systems, technical specifications and program and procedures to change of NPP condition from in an operation condition to in a defueled condition. The many systems to need to operate in an operational condition will not be operated during in a defueled condition and the function of systems will be changed from in an operation condition to in a defueled condition. So a downgrade of systems may be needed and reclassifying the safety class of structure, system and component (SSC) may be conducted. By the reclassification of SSC, activity related with quality assurance and maintenance of SSC is affected. In this paper, the method to reclassify SSC in a defueled condition is studied. The many systems to need to operate in an operational condition will not be operated during in a defueled condition and the function of systems will be changed from in an operation condition to in a defueled condition. The operation of NPP during a defueled condition need to conduct licensing and engineering work need to change the design basis of the plant optimize by downgrading systems and reclassifying the safety class of SSC. In this paper, the method to reclassify safety class for a defueled condition is studied.

Safety culture : a significant influence on safety in transportation

Science.gov (United States)

2017-08-01

An organizations safety culture can influence safety outcomes. Research and experience show that when safety culture is strong, accidents are less frequent and less severe. As a result, building and maintaining strong safety cultures should be a t...

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

NASA System Safety Handbook. Volume 2: System Safety Concepts, Guidelines, and Implementation Examples

Science.gov (United States)

Dezfuli, Homayoon; Benjamin, Allan; Everett, Christopher; Feather, Martin; Rutledge, Peter; Sen, Dev; Youngblood, Robert

2015-01-01

This is the second of two volumes that collectively comprise the NASA System Safety Handbook. Volume 1 (NASASP-210-580) was prepared for the purpose of presenting the overall framework for System Safety and for providing the general concepts needed to implement the framework. Volume 2 provides guidance for implementing these concepts as an integral part of systems engineering and risk management. This guidance addresses the following functional areas: 1.The development of objectives that collectively define adequate safety for a system, and the safety requirements derived from these objectives that are levied on the system. 2.The conduct of system safety activities, performed to meet the safety requirements, with specific emphasis on the conduct of integrated safety analysis (ISA) as a fundamental means by which systems engineering and risk management decisions are risk-informed. 3.The development of a risk-informed safety case (RISC) at major milestone reviews to argue that the systems safety objectives are satisfied (and therefore that the system is adequately safe). 4.The evaluation of the RISC (including supporting evidence) using a defined set of evaluation criteria, to assess the veracity of the claims made therein in order to support risk acceptance decisions.

Application of life-cycle information for advancement in safety of nuclear fuel cycle facilities. Application of safety information to advanced safety management support system

International Nuclear Information System (INIS)

Suzuki, Kazuhiko; Ishida, Michihiko

2005-08-01

Risk management is major concern to nuclear energy reprocessing plants to improve plant and process reliability and ensure their safety. This is because we are required to predict potential risks before any accident or disaster occurs. The advancement of safety design and safety systems technologies showed large amount of useful safety-related knowledge that can be of great importance to plant operation to reduce operation risks and ensure safety. This research proposes safety knowledge modeling framework on the basis of ontology technologies to systematically construct plant knowledge model, which includes plant structure, operation, and the associated behaviors. In such plant knowledge model safety related information is defined and linked to the different elements of plant knowledge model. Ontology editor is employed to define the basic concepts and their inter-relations, which are used to capture and construct plant safety knowledge. In order to provide detailed safety knowledgebase, HAZOP results are analyzed and structured so that safety-related knowledge are identified and structured within the plant knowledgebase. The target safety knowledgebase includes: failures, deviations, causes, consequences, and fault propagation as mapped to plant knowledge. The proposed ontology-based safety framework is applied on case study nuclear plant to structure failures, causes, consequences, and fault propagation, which are used to support plant operation. (author)

From Board to Bedside: How the Application of Financial Structures to Safety and Quality Can Drive Accountability in a Large Health Care System.

Science.gov (United States)

Austin, J Matthew; Demski, Renee; Callender, Tiffany; Lee, K H Ken; Hoffman, Ann; Allen, Lisa; Radke, Deborah A; Kim, Yungjin; Werthman, Ronald J; Peterson, Ronald R; Pronovost, Peter J

2017-04-01

As the health care system in the United States places greater emphasis on the public reporting of quality and safety data and its use to determine payment, provider organizations must implement structures that ensure discipline and rigor regarding these data. An academic health system, as part of a performance management system, applied four key components of a financial reporting structure to support the goal of top-to-bottom accountability for improving quality and safety. The four components implemented by Johns Hopkins Medicine were governance, accountability, reporting of consolidated quality performance statements, and auditing. Governance is provided by the health system's Patient Safety and Quality Board Committee, which reviews goals and strategy for patient safety and quality, reviews quarterly performance for each entity, and holds organizational leaders accountable for performance. An accountability plan includes escalating levels of review corresponding to the number of months an entity misses the defined performance target for a measure. A consolidated quality statement helps inform the Patient Safety and Quality Board Committee and leadership on key quality and safety issues. An audit evaluates the efficiency and effectiveness of processes for data collection, validation, and storage, as to ensure the accuracy and completeness of quality measure reporting. If hospitals and health systems truly want to prioritize improvements in safety and quality, they will need to create a performance management system that ensures data validity and supports performance accountability. Without valid data, it is difficult to know whether a performance gap is due to data quality or clinical quality. Copyright © 2017 The Joint Commission. Published by Elsevier Inc. All rights reserved.

Identifying behaviour patterns of construction safety using system archetypes.

Science.gov (United States)

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.

Using system dynamics simulation for assessment of hydropower system safety

Science.gov (United States)

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.

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

Integrated model of port oil piping transportation system safety including operating environment threats

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.

Integrated model of port oil piping transportation system safety including operating environment threats

OpenAIRE

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.

Reactor safety systems

International Nuclear Information System (INIS)

Kafka, P.

1975-01-01

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

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

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)

Procedure for getting safety classed concrete structures approved by Finnish Radiation and Nuclear Safety Authority

International Nuclear Information System (INIS)

Halme, Ville-Juhani

2015-01-01

Posiva is preparing geological final disposal in the Finnish bedrock in Olkiluoto, Eurajoki. The final disposal facility includes encapsulation plant and underground repository. Most of the main civil structures are concrete structures. STUK is the supervising authority in civil structures. The National Building Code of Finland and STUK's Regulatory Guide on nuclear safety (YVL) are the most important instructions when constructing concrete structures into nuclear installation. Posiva has classified concrete structures in two classes according STUK's YVL-guidance: EYT (non-nuclear) and Safety Class 3 (SC 3, nuclear safety significance). When building SC 3 concrete structures, specific protocol must be followed. Protocol includes planned routines for design, construction, supervision, quality control (QC) and quality assurance (QA) activities. Documents relating concrete structures must be approved by Posiva and STUK before construction work. After structures have been designed and actual building is ongoing, there are two main steps. Before concreting, readiness inspection for concreting must be arranged. Readiness inspection will be arranged according to a specific plan and the date must be informed to STUK. After establishing readiness for concreting, casting work can begin. Once concrete structures are done, inspected and approved, final documentation according to a quality control plan will be reviewed by Posiva. After Posiva's approval, final documentation will be sent for STUK's approval. In the end STUK will give the permission for commissioning of the concrete structures after approved commissioning inspection. The document is made up of an abstract and a poster

Structural safety features for superconducting magnets

International Nuclear Information System (INIS)

Lehner, J.; Reich, M.; Powell, J.; Bezler, P.; Gardner, D.; Yu, W.; Chang, T.Y.

1975-01-01

A survey has been carried out for various potential structural safety problems of superconducting fusion magnets. These areas include: (1) Stresses due to inhomogeneous temperature distributions in magnets where normal regions have been initiated. (2) Stress distributions and yield forces due to cracks and failed regions. (3) Superconducting magnet response due to seismic excitation. These analyses have been carried out using a variety of large capacity finite element computer codes that allow for the evaluation of stresses in elastic or elastic-plastic zones and around singularities in the magnet structure. Thus far, these analyses have been carried out on UWMAK-I type magnet systems

Structural empowerment and patient safety culture among registered nurses working in adult critical care units.

Science.gov (United States)

Armellino, Donna; Quinn Griffin, Mary T; Fitzpatrick, Joyce J

2010-10-01

The aim of the present study was to examine the relationship between structural empowerment and patient safety culture among staff level Registered Nurses (RNs) within adult critical care units (ACCU). There is literature to support the value of RNs' structurally empowered work environments and emerging literature towards patient safety culture; the link between empowerment and patient safety culture is being discovered. A sample of 257 RNs, working within adult critical care of a tertiary hospital in the United States, was surveyed. Instruments included a background data sheet, the Conditions of Workplace Effectiveness and the Hospital Survey on Patient Safety Culture. Structural empowerment and patient safety culture were significantly correlated. As structural empowerment increased so did the RNs' perception of patient safety culture. To foster patient safety culture, nurse leaders should consider providing structurally empowering work environments for RNs. This study contributes to the body of knowledge linking structural empowerment and patient safety culture. Results link structurally empowered RNs and increased patient safety culture, essential elements in delivering efficient, competent, quality care. They inform nursing management of key factors in the nurses' environment that promote safe patient care environments. © 2010 The Authors. Journal compilation © 2010 Blackwell Publishing Ltd.

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...

Performance scorecard for occupational safety and health management systems

Directory of Open Access Journals (Sweden)

Hernâni Veloso Neto

2012-06-01

Full Text Available The pro-active and systematic search for best performances should be the two assumptions of any management system, so safety and health management in organizations must also be guided by these same precepts. However, the scientific production evidences that the performance evaluation processes in safety and health continue to be guided, in their essence, by intermittency, reactivity and negativity, which are not consistent with the assumptions referenced above. Therefore, it is essential that health and safety at work management systems (HSW MS are structured from an active and positive viewpoint, focusing on continuous improvement. This implies considering performance evaluation processes that incorporate, on the one hand, monitoring, measuring and verification procedures, and on the other hand, structured matrixes of results that capture the key factors of success, by mobilizing both reactive and proactive indicators. One of the instruments that can fulfill these precepts of health and safety performance evaluation is the SafetyCard, a performance scorecard for HSW MS that we developed and will seek to outline and demonstrate over this paper.

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

Probabilistic safety analysis of earth retaining structures during earthquakes

Science.gov (United States)

Grivas, D. A.; Souflis, C.

1982-07-01

A procedure is presented for determining the probability of failure of Earth retaining structures under static or seismic conditions. Four possible modes of failure (overturning, base sliding, bearing capacity, and overall sliding) are examined and their combined effect is evaluated with the aid of combinatorial analysis. The probability of failure is shown to be a more adequate measure of safety than the customary factor of safety. As Earth retaining structures may fail in four distinct modes, a system analysis can provide a single estimate for the possibility of failure. A Bayesian formulation of the safety retaining walls is found to provide an improved measure for the predicted probability of failure under seismic loading. The presented Bayesian analysis can account for the damage incurred to a retaining wall during an earthquake to provide an improved estimate for its probability of failure during future seismic events.

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

Industrial safety: its structuring and content

International Nuclear Information System (INIS)

Munoz, A.; Rodriguez, J.; Martinez-Val, J.M.

1999-01-01

Industrial development has led to an on-going increase in productivity, but the concept of safety has also become highly relevant. In this article, the authors address the structuring and content of industrial safety which involves laying down essential safety requirements, both in manufacturing and processes and in products. (Author)

Integrated therapy safety management system.

Science.gov (United States)

Podtschaske, Beatrice; Fuchs, Daniela; Friesdorf, Wolfgang

2013-09-01

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

Integrated therapy safety management system

Science.gov (United States)

Podtschaske, Beatrice; Fuchs, Daniela; Friesdorf, Wolfgang

2013-01-01

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

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

Science.gov (United States)

Hill, Janice L.

2007-01-01

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

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

Significance of Alkali-Silica reaction in nuclear safety-related concrete structures

International Nuclear Information System (INIS)

Le Pape, Y.; Field, K.G.; Mattus, C.H.; Naus, D.J.; Busby, J.T.; Saouma, V.; Ma, Z.J.; Cabage, J.V.; Guimaraes, M.

2015-01-01

Nuclear Power Plant license renewal up to 60 years and possible life extension beyond has established a renewed focus on long-term aging of nuclear generating stations materials, and particularly, on concrete. Large irreplaceable sections of most nuclear generating stations include concrete components. The Expanded Materials Degradation Analysis, jointly performed by the Department of Energy, the U.S. Nuclear Regulatory Commission, the Academia and the Power Generation Industry, identified the need to develop a consistent knowledge base of alkali-silica reaction (ASR) within concrete as an urgent priority (Graves et al., 2014). ASR results in an expansion of Concrete produced by the reaction between alkali (generally from cement), reactive aggregate (like amorphous silica) and water absorption. ASR causes expansion, cracking and loss of mechanical properties. Considering that US commercial reactors in operation enter the age when ASR distress can be potentially observed and that numerous non-nuclear infrastructures (transportation, energy production) in a majority of the States have already experienced ASR-related concrete degradation, the susceptibility and significance of ASR for nuclear concrete structures must be addressed. This paper outlines an on-going research program including the investigation of the possibility of ASR in nuclear power plants, and the assessment of the residual shear bearing capacity of ASR-subjected nuclear structures. (authors)

Safety system function trends

International Nuclear Information System (INIS)

Johnson, C.

1989-01-01

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

Safety Metrics for Human-Computer Controlled Systems

Science.gov (United States)

Leveson, Nancy G; Hatanaka, Iwao

2000-01-01

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

Robustness of Structural Systems

DEFF Research Database (Denmark)

Canisius, T.D.G.; Sørensen, John Dalsgaard; Baker, J.W.

2007-01-01

The importance of robustness as a property of structural systems has been recognised following several structural failures, such as that at Ronan Point in 1968,where the consequenceswere deemed unacceptable relative to the initiating damage. A variety of research efforts in the past decades have...... attempted to quantify aspects of robustness such as redundancy and identify design principles that can improve robustness. This paper outlines the progress of recent work by the Joint Committee on Structural Safety (JCSS) to develop comprehensive guidance on assessing and providing robustness in structural...... systems. Guidance is provided regarding the assessment of robustness in a framework that considers potential hazards to the system, vulnerability of system components, and failure consequences. Several proposed methods for quantifying robustness are reviewed, and guidelines for robust design...

The Management System for Nuclear Installations. Safety Guide (Spanish Edition)

International Nuclear Information System (INIS)

2017-01-01

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

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)

Structural safety - Is the safety margin measurable

International Nuclear Information System (INIS)

Rintamaa, R.

1992-01-01

In ensuring the structural safety of the nuclear components one must be aware of the uncertainties related to the material deorientation, loadings and other operational conditions, geometrical dimensions as well as the service environment. Furthermore, the validation of the analysis tools and procedures is of great importance in overall safety assessment of a pressure retaining component. In order to identify and quantify the concerns and risks arising from the uncertainties in the safety related issue intensive research is being carried out all over the world, in particular, on the ageing, plant life extension and management of old nuclear power plants. The presentation includes a general survey of the factors relevant to the assessment of safe and reliable operation of a nuclear component throughout its planned service life. Certain aspects are outlined based on the research work being carried out at the Technical Research Centre of Finland (VTT)(orig.)

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)

Safety logic systems of PFBR

International Nuclear Information System (INIS)

Sambasivan, S. Ilango

2004-01-01

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

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

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)

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

Assessment of the factors with significant influence on safety culture

International Nuclear Information System (INIS)

Farcasiu, M.; Nitoi, M.

2013-01-01

In this paper, a qualitative and a quantitative evaluation of the factors with significant impact on safety culture were performed. These techniques were established and applied in accordance with IAEA standards. In order to show the applicability and opportunity of the methodology a specific case study was prepared: safety culture evaluation for INR Pitesti. The qualitative evaluation was performed using specific developed questionnaires. Through analysis of the completed questionnaires was established the development stage of safety culture at INR. The quantitative evaluation was performed using a guide to rate the influence factors. For each factor was identified the influence (negative or positive) and ranking score was estimated using scoring criteria. The results have emphasized safety culture stages. The paper demonstrates the fact that using both quantitative and qualitative assessment techniques, a practical value of the safety culture concept is given. (authors)

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

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)

Safety system upgrades to a research reactor: A regulatory perspective

International Nuclear Information System (INIS)

Lamarre, G.B.; Martin, W.G.

2003-01-01

The NRU (National Research Universal) reactor, located at the Chalk River Laboratories of Atomic Energy of Canada Limited (AECL), first achieved criticality November 3, 1957. AECL continues to operate NRU for research to support safety and reliability studies for CANDU reactors and as a major supplier of medical radioisotopes. Following a detailed systematic review and assessment of NRU's design and the condition of its primary systems, AECL formally notified the Canadian Nuclear Safety Commission's (CNSC) predecessor - the Atomic Energy Control Board - in 1992 of its intention to upgrade NRU's safety systems. AECL proposed seven major upgrades to provide improvements in shutdown capability, heat removal, confinement, and reactor monitoring, particularly during and after a seismic event. From a CNSC perspective, these upgrades were necessary to meet modern safety standards. From the start of the upgrades project, the CNSC provided regulatory oversight aimed at ensuring that AECL maintained a structured approach to the upgrades. The elements of the approach include, but are not limited to, the determination of project milestones and target dates; the formalization of the design process and project quality assurance requirements; the requirements for updated documentation, including safety reports, safety notes and commissioning reports; and the approval and authorization process. This paper details, from a regulatory perspective, the structured approach used in approving the design, construction, commissioning and subsequent operation of safety system upgrades for an existing and operating research reactor, including the many challenges faced when attempting to balance the requirements of the upgrades project with AECL's need to keep NRU operating to meet its important research and production objectives. (author)

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.

Identifying the most significant indicators of the total road safety performance index.

Science.gov (United States)

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

Study on criticality safety evaluation of a system where flood will never occur

International Nuclear Information System (INIS)

Naito, Yoshitaka; Yamamoto, Toshihiro; Komuro, Yuichi; Itahara, Kuniyuki.

1995-03-01

Criticality safety evaluation for a single unit containing nuclear fuel has usually been performed on the assumption that there is a fully thick water reflector around the unit. For a system where flood will never occur, however, the thick reflector assumption is usually not applied recently. In such cases, a method is proposed, which models surrounding structural material and branch pipes as 2.5cm thick water reflector. This report shows that reactivity worth of structural material and branch pipes is, in many cases, less than that of 2.5cm thick water reflector. Further, another method is shown to evaluate criticality safety for a multiple unit system, using computed results with surrounding structural material and branch pipes neglected. And it is shown with many sample calculations that the method with 2.5cm thick water reflector in place of structural material and pipes gives safety side results to similar systems to real reprocessing plants. (author)

A concept of safety indicator system for nuclear power plants

International Nuclear Information System (INIS)

Lehtinen, E.

1995-12-01

The fundamental principle in the safety technology of nuclear power is embodied in the strategy of defence in depth. The defence lines of the strategy, completed with a PSA logic model and structure, are considered to provide an appropriate framework for identification and structuring of the operational safety performance areas for nuclear power plants. Once these areas are identified the safety indicators can be defined. Based on this approach a concept of safety indicator system was outlined. About one hundred indicator specifications have been collected, refined and related to the performance areas. The specifications enable the utilities and authorities to check the coverage of their indicators set from the operational safety point of view and select or refine indicators for testing and routine use. Finally various statistical approaches and methods for using indicators in performance evaluation are presented. (orig.) (16 refs., 2 figs., 2 tabs.)

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

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.

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)

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)

Implementation of safety management systems in Hong Kong construction industry - A safety practitioner's perspective.

Science.gov (United States)

Yiu, Nicole S N; Sze, N N; Chan, Daniel W M

2018-02-01

In the 1980s, the safety management system (SMS) was introduced in the construction industry to mitigate against workplaces hazards, reduce the risk of injuries, and minimize property damage. Also, the Factories and Industrial Undertakings (Safety Management) Regulation was introduced on 24 November 1999 in Hong Kong to empower the mandatory implementation of a SMS in certain industries including building construction. Therefore, it is essential to evaluate the effectiveness of the SMS in improving construction safety and identify the factors that influence its implementation in Hong Kong. A review of the current state-of-the-practice helped to establish the critical success factors (CSFs), benefits, and difficulties of implementing the SMS in the construction industry, while structured interviews were used to establish the key factors of the SMS implementation. Results of the state-of-the-practice review and structured interviews indicated that visible senior commitment, in terms of manpower and cost allocation, and competency of safety manager as key drivers for the SMS implementation. More so, reduced accident rates and accident costs, improved organization framework, and increased safety audit ratings were identified as core benefits of implementing the SMS. Meanwhile, factors such as insufficient resources, tight working schedule, and high labor turnover rate were the key challenges to the effective SMS implementation in Hong Kong. The findings of the study were consistent and indicative of the future development of safety management practice and the sustainable safety improvement of Hong Kong construction industry in the long run. Copyright © 2018 National Safety Council and Elsevier Ltd. All rights reserved.

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)

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

Science.gov (United States)

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

2011-01-01

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

A safety equipment list for rotary mode core sampling systems operation in single shell flammable gas tanks

International Nuclear Information System (INIS)

SMALLEY, J.L.

1999-01-01

This document identifies all interim safety equipment to be used for rotary mode core sampling of single-shell flammable gas tanks utilizing Rotary Mode Core Sampling systems (RMCS). This document provides the safety equipment for RMCS trucks HO-68K-4600, HO-68K-4647, trucks three and four respectively, and associated equipment. It is not intended to replace or supersede WHC-SD-WM-SEL-023, (Kelly 1991), or WHC-SD-WM-SEL-032, (Corbett 1994), which classifies 80-68K-4344 and HO-68K-4345 respectively. The term ''safety equipment'' refers to safety class (SC) and safety significant (SS) equipment, where equipment refers to structures, systems and components (SSC's). The identification of safety equipment in this document is based on the credited design safety features and analysis contained in the Authorization Basis (AB) for rotary mode core sampling operations in single-shell flammable gas tanks. This is an interim safety classification since the AB is interim. This document will be updated to reflect the final RMCS equipment safety classification designations upon completion of a final AB which will be implemented with the release of the Final Safety Analysis Report (FSAR)

Computer-aided safety systems of industrial high energy objects

International Nuclear Information System (INIS)

Topolsky, N.G.; Gordeev, S.G.

1995-01-01

Modern objects of fuel and energy, chemical industries are characterized by high power consumption; by presence of large quantities of combustible and explosive substances used in technological processes; by advanced communications of submission systems of initial liquid and gasiform reagents, lubricants and coolants, the products of processing, and wastes of production; by advanced ventilation and pneumatic transport; and by complex control systems of energy, material and information flows. Such objects have advanced infrastructures, including a significant quantity of engineering buildings intended for storage, transportation, and processing of combustible liquids, gasiform fuels and materials, and firm materials. Examples of similar objects are nuclear and thermal power stations, chemical plants, machine-building factories, iron and steel industry enterprises, etc. Many tasks and functions characterizing the problem of fire safety of these objects can be accomplished only upon the development of special Computer-Aided Fire Safety Systems (CAFSS). The CAFSS for these objects are intended to reduce the hazard of disastrous accidents both causing fires and caused by them. The tasks of fire prevention and rescue work of large-scale industrial objects are analyzed within the bounds of the recommended conception. A functional structure of CAFSS with a list of the main subsystems forming a part of its composition has been proposed

Development of the safety evaluation system in the respects of organizational factors and workers' consciousness. Pt. 4. Application of the system for contract companies

International Nuclear Information System (INIS)

Hasegawa, Naoko; Hirose, Ayako; Hayase, Kenichi; Tsuge, Tadafumi; Sasou, Kunihide; Takano, Kenichi

2003-01-01

The purpose of our study is to develop a safety evaluation system which clarifies the safety level of an organization. As a basic method of evaluation using a questionnaire had been established, now that the generalization is needed for the system. Hence, this paper is intended to consider the applicability of the system for contract companies. Subjects were workers who belonged to contract companies engaging in the maintenance of power plants in regular inspections. The following results were obtained: 1) The Comprehensive Safety Index (CSI) taking into account individual and organizational factors was identified using the principal component analysis. 2) The validity of CSI was confirmed with significant correlations between the CSI score and the rate of accidents. 3) Careful consideration should be provided for individual factors especially when evaluating the safety level of subcontract companies. 4) It seemed necessary to take into account the influence of parent companies and occupational hazards level. 5) The comparison among different industries should be avoided because of the difference in organizational structures and subjects of attention for keeping safety. (author)

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 ?

NARCIS (Netherlands)

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

2017-01-01

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

Scale development of safety management system evaluation for the airline industry.

Science.gov (United States)

Chen, Ching-Fu; Chen, Shu-Chuan

2012-07-01

The airline industry relies on the implementation of Safety Management System (SMS) to integrate safety policies and augment safety performance at both organizational and individual levels. Although there are various degrees of SMS implementation in practice, a comprehensive scale measuring the essential dimensions of SMS is still lacking. This paper thus aims to develop an SMS measurement scale from the perspective of aviation experts and airline managers to evaluate the performance of company's safety management system, by adopting Schwab's (1980) three-stage scale development procedure. The results reveal a five-factor structure consisting of 23 items. The five factors include documentation and commands, safety promotion and training, executive management commitment, emergency preparedness and response plan and safety management policy. The implications of this SMS evaluation scale for practitioners and future research are discussed. Copyright © 2012 Elsevier Ltd. All rights reserved.

The Daresbury personnel safety system

International Nuclear Information System (INIS)

Poole, D.E.; Ring, T.

1989-01-01

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

Structure Optimization of Safety Investment of Petrochemical Port Enterprises

Directory of Open Access Journals (Sweden)

Zhiqiang Hou

2017-01-01

Full Text Available Safety investment is an essential guarantee to identify and manage potential security problems in petrochemical port enterprises. The reasonability of safety investment structure determines overall security risks in an enterprise. Based on the definition of risks, combining Cobb-Douglas production function with FTA probability model, and taking Gompertz curve model as the constraint condition, structure optimization model of safety investment is built in order to minimize risks and work out the safety investment structure of petrochemical port enterprises. According to the case study, the calculations indicate that safety investment in corporate management presents a larger growth rate than past years and that unsafe act of human being is the main factor accounting for the greatest probability of occurrence, which is consistent with previous accident investigation results as well as enterprise reality. This testifies that the model is effective and that the results can guide the allocation of safety investment of petrochemical port enterprises scientifically.

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

Application of the AHP method to analyze the significance of the factors affecting road traffic safety

Directory of Open Access Journals (Sweden)

Justyna SORDYL

2015-06-01

Full Text Available Over the past twenty years, the number of vehicles registered in Poland has grown rapidly. At the same time, a relatively small increase in the length of the road network has been observed. As a result of the limited capacity of available infrastructure, it leads to significant congestion and to increase of the probability of road accidents. The overall level of road safety depends on many factors - the behavior of road users, infrastructure solutions and the development of automotive technology. Thus the detailed assessment of the importance of individual elements determining road safety is difficult. The starting point is to organize the factors by grouping them into categories which are components of the DVE system (driver - vehicle - environment. In this work, to analyze the importance of individual factors affecting road safety, the use of analytic hierarchy process method (AHP was proposed. It is one of the multi-criteria methods which allows us to perform hierarchical analysis of the decision process, by means of experts’ opinions. Usage of AHP method enabled us to evaluate and rank the factors affecting road safety. This work attempts to link the statistical data and surveys in significance analysis of the elements determining road safety.

Major structural response methods used in the seismic safety margins research program

International Nuclear Information System (INIS)

Chou, C.K.; Lo, T.; Vagliente, V.

1979-01-01

In order to evaluate the conservatisms in present nuclear power plant seismic safety requirements, a probabilistic based systems model is being developed. This model will also be used to develop improved requirements. In Phase I of the Seismic Safety Margins Research Program (SSMRP), this methodology will be developed for a specific nuclear power plant and used to perform probabilistic sensitivity studies to gain engineering insights into seismic safety requirements. Random variables in the structural response analysis area, or parameters which cause uncertainty in the response, are discussed and classified into three categories; i.e., material properties, structural dynamic characteristics and related modeling techniques, and analytical methods. The sensitivity studies are grouped into two categories; deterministic and probabilistic. In a system analysis, transfer functions in simple form are needed since there are too many responses which have to be calculated in a Monte Carlo simulation to use the usual straightforward calculation approach. Therefore, the development of these simple transfer functions is one of the important tasks in SSMRP. Simplified as well as classical transfer functions are discussed

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

Structure soil structure interaction effects: Seismic analysis of safety related collocated concrete structures

International Nuclear Information System (INIS)

Joshi, J.R.

2000-01-01

The Process, Purification and Stack Buildings are collocated safety related concrete shear wall structures with plan dimensions in excess of 100 feet. An important aspect of their seismic analysis was the determination of structure soil structure interaction (SSSI) effects, if any. The SSSI analysis of the Process Building, with one other building at a time, was performed with the SASSI computer code for up to 50 frequencies. Each combined model had about 1500 interaction nodes. Results of the SSSI analysis were compared with those from soil structure interaction (SSI) analysis of the individual buildings, done with ABAQUS and SASSI codes, for three parameters: peak accelerations, seismic forces and the in-structure floor response spectra (FRS). The results may be of wider interest due to the model size and the potential applicability to other deep soil layered sites. Results obtained from the ABAQUS analysis were consistently higher, as expected, than those from the SSI and SSSI analyses using the SASSI. The SSSI effect between the Process and Purification Buildings was not significant. The Process and Stack Building results demonstrated that under certain conditions a massive structure can have an observable effect on the seismic response of a smaller and less stiff structure

The roles of the seismic safety and monitoring systems in the PEC fast reactor

International Nuclear Information System (INIS)

Masoni, P.; Di Tullio, E.M.; Massa, B.; Martelli, A.; Sano, T.

1988-01-01

Two different seismic systems are foreseen in the case of PEC: the seismic safety system, that provides the automatic scram, and the seismic monitoring system. During earthquake, three triaxial seismic switches are triggered if a threshold value of the ground acceleration is exceeded. In this case, the signals from the seismic switches are processed by the safety system (with a 2/3 logic) and the shutdown system is triggered. Peak acceleration is the parameter used by the safety system to quantify the seismic event. This way, however, no information is obtained with regard to earthquake frequency content. Thus, reactor safety is guaranteed by adopting a threshold considerably lower than the Z.P.A. of the Design Basis Earthquake. Furthermore, in the case of significant earthquakes, the seismic motion is measured by about 20 triaxial accelerometers, located both in the free field and on the plant's structures. Data are digitazed and recordered by the seismic monitoring system. This system also elaborates the recordered time-histories providing floor response spectra and compares such spectra to the design values. The above-mentioned elaborations and comparisons are performed in short time for two triaxial measuring positions, thus allowing the Operator to immediately get a more complete information on the seismic event. The complete set of data recorded by the seismic monitoring system also allows the actual dynamic response of the plant to be determined and compared to the design values. On the basis of this comparison the necessary safety analysis can be carried out to verify whether the design limits of the plant were respected: in the positive case the reactor can be restarted. (author)

Comprehensive Lifecycle for Assuring System Safety

Science.gov (United States)

Knight, John C.; Rowanhill, Jonathan C.

2017-01-01

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

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

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

A practical monitoring system for the structural safety of mega-trusses using wireless vibrating wire strain gauges.

Science.gov (United States)

Park, Hyo Seon; Lee, Hwan Young; Choi, Se Woon; Kim, Yousok

2013-12-16

Sensor technologies have been actively employed in structural health monitoring (SHM) to evaluate structural safety. To provide stable and real-time monitoring, a practical wireless sensor network system (WSNS) based on vibrating wire strain gauges (VWSGs) is proposed and applied to a building under construction. In this WSNS, the data measured from each VWSG are transmitted to the sensor node via a signal line and then transmitted to the master node through a short-range wireless communication module (operating on the Industrial, Scientific, and Medical (ISM) band). The master node also employs a long-range wireless communication module (Code Division Multiple Access-CDMA) to transmit the received data from the sensor node to a server located in a remote area, which enables a manager to examine the measured data in real time without any time or location restrictions. In this study, a total of 48 VWSGs, 14 sensor nodes, and seven master nodes were implemented to measure long-term strain variations of mega-trusses in an irregular large-scale building under construction. Based on strain data collected over a 16-month period, a quantitative evaluation of the construction process was performed to determine the aspects that exhibit the greatest influence on member behavior and to conduct a comparison with numerical simulation results. The effect of temperature stress on the structural elements was also analyzed. From these observations, the feasibility of a long-term WSNS based on VWSGs to evaluate the structural safety of an irregular building under construction was confirmed.

A Practical Monitoring System for the Structural Safety of Mega-Trusses Using Wireless Vibrating Wire Strain Gauges

Directory of Open Access Journals (Sweden)

Hyo Seon Park

2013-12-01

Full Text Available Sensor technologies have been actively employed in structural health monitoring (SHM to evaluate structural safety. To provide stable and real-time monitoring, a practical wireless sensor network system (WSNS based on vibrating wire strain gauges (VWSGs is proposed and applied to a building under construction. In this WSNS, the data measured from each VWSG are transmitted to the sensor node via a signal line and then transmitted to the master node through a short-range wireless communication module (operating on the Industrial, Scientific, and Medical (ISM band. The master node also employs a long-range wireless communication module (Code Division Multiple Access—CDMA to transmit the received data from the sensor node to a server located in a remote area, which enables a manager to examine the measured data in real time without any time or location restrictions. In this study, a total of 48 VWSGs, 14 sensor nodes, and seven master nodes were implemented to measure long-term strain variations of mega-trusses in an irregular large-scale building under construction. Based on strain data collected over a 16-month period, a quantitative evaluation of the construction process was performed to determine the aspects that exhibit the greatest influence on member behavior and to conduct a comparison with numerical simulation results. The effect of temperature stress on the structural elements was also analyzed. From these observations, the feasibility of a long-term WSNS based on VWSGs to evaluate the structural safety of an irregular building under construction was confirmed.

Advanced analysis and design for fire safety of steel structures

CERN Document Server

Li, Guoqiang

2013-01-01

Advanced Analysis and Design for Fire Safety of Steel Structures systematically presents the latest findings on behaviours of steel structural components in a fire, such as the catenary actions of restrained steel beams, the design methods for restrained steel columns, and the membrane actions of concrete floor slabs with steel decks. Using a systematic description of structural fire safety engineering principles, the authors illustrate the important difference between behaviours of an isolated structural element and the restrained component in a complete structure under fire conditions. The book will be an essential resource for structural engineers who wish to improve their understanding of steel buildings exposed to fires. It is also an ideal textbook for introductory courses in fire safety for master’s degree programs in structural engineering, and is excellent reading material for final-year undergraduate students in civil engineering and fire safety engineering. Furthermore, it successfully bridges th...

Operating experience feedback from safety significant events at research reactors

Energy Technology Data Exchange (ETDEWEB)

Shokr, A.M. [Atomic Energy Authority, Abouzabal (Egypt). Egypt Second Research Reactor; Rao, D. [Bhabha Atomic Research Centre, Mumbai (India)

2015-05-15

Operating experience feedback is an effective mechanism to provide lessons learned from the events and the associated corrective actions to prevent recurrence of events, resulting in improving safety in the nuclear installations. This paper analyzes the events of safety significance that have been occurred at research reactors and discusses the root causes and lessons learned from these events. Insights from literature on events at research reactors and feedback from events at nuclear power plants that are relevant to research reactors are also presented along with discussions. The results of the analysis showed the importance of communication of safety information and exchange of operating experience are vital to prevent reoccurrences of events. The analysis showed also the need for continued attention to human factors and training of operating personnel, and the need for establishing systematic ageing management programmes of reactor facilities, and programmes for safety management of handling of nuclear fuel, core components, and experimental devices.

Operating experience feedback from safety significant events at research reactors

International Nuclear Information System (INIS)

Shokr, A.M.

2015-01-01

Operating experience feedback is an effective mechanism to provide lessons learned from the events and the associated corrective actions to prevent recurrence of events, resulting in improving safety in the nuclear installations. This paper analyzes the events of safety significance that have been occurred at research reactors and discusses the root causes and lessons learned from these events. Insights from literature on events at research reactors and feedback from events at nuclear power plants that are relevant to research reactors are also presented along with discussions. The results of the analysis showed the importance of communication of safety information and exchange of operating experience are vital to prevent reoccurrences of events. The analysis showed also the need for continued attention to human factors and training of operating personnel, and the need for establishing systematic ageing management programmes of reactor facilities, and programmes for safety management of handling of nuclear fuel, core components, and experimental devices.

A safety equipment list for rotary mode core sampling systems operation in single shell flammable gas tanks; TOPICAL

International Nuclear Information System (INIS)

SMALLEY, J.L.

1999-01-01

This document identifies all interim safety equipment to be used for rotary mode core sampling of single-shell flammable gas tanks utilizing Rotary Mode Core Sampling systems (RMCS). This document provides the safety equipment for RMCS trucks HO-68K-4600, HO-68K-4647, trucks three and four respectively, and associated equipment. It is not intended to replace or supersede WHC-SD-WM-SEL-023, (Kelly 1991), or WHC-SD-WM-SEL-032, (Corbett 1994), which classifies 80-68K-4344 and HO-68K-4345 respectively. The term ''safety equipment'' refers to safety class (SC) and safety significant (SS) equipment, where equipment refers to structures, systems and components (SSC's). The identification of safety equipment in this document is based on the credited design safety features and analysis contained in the Authorization Basis (AB) for rotary mode core sampling operations in single-shell flammable gas tanks. This is an interim safety classification since the AB is interim. This document will be updated to reflect the final RMCS equipment safety classification designations upon completion of a final AB which will be implemented with the release of the Final Safety Analysis Report (FSAR)

Breaking Dense Structures: Proving Stability of Densely Structured Hybrid Systems

Directory of Open Access Journals (Sweden)

Eike Möhlmann

2015-06-01

Full Text Available Abstraction and refinement is widely used in software development. Such techniques are valuable since they allow to handle even more complex systems. One key point is the ability to decompose a large system into subsystems, analyze those subsystems and deduce properties of the larger system. As cyber-physical systems tend to become more and more complex, such techniques become more appealing. In 2009, Oehlerking and Theel presented a (de-composition technique for hybrid systems. This technique is graph-based and constructs a Lyapunov function for hybrid systems having a complex discrete state space. The technique consists of (1 decomposing the underlying graph of the hybrid system into subgraphs, (2 computing multiple local Lyapunov functions for the subgraphs, and finally (3 composing the local Lyapunov functions into a piecewise Lyapunov function. A Lyapunov function can serve multiple purposes, e.g., it certifies stability or termination of a system or allows to construct invariant sets, which in turn may be used to certify safety and security. In this paper, we propose an improvement to the decomposing technique, which relaxes the graph structure before applying the decomposition technique. Our relaxation significantly reduces the connectivity of the graph by exploiting super-dense switching. The relaxation makes the decomposition technique more efficient on one hand and on the other allows to decompose a wider range of graph structures.

Improved Management of Part Safety Classification System for Nuclear Power Plant

Energy Technology Data Exchange (ETDEWEB)

Park, Jin Young; Park, Youn Won; Park, Heung Gyu; Park, Hyo Chan [BEES Inc., Daejeon (Korea, Republic of)

2016-10-15

As, in recent years, many quality assurance (QA) related incidents, such as falsely-certified parts and forged documentation, etc., were reported in association with the supply of structures, systems, components and parts to nuclear power plants, a need for a better management of safety classification system was addressed so that it would be based more on the level of parts . Presently, the Korean nuclear power plants do not develop and apply relevant procedures for safety classifications, but rather the safety classes of parts are determined solely based on the experience of equipment designers. So proposed in this paper is a better management plan for safety equipment classification system with an aim to strengthen the quality management for parts. The plan was developed through the analysis of newly introduced technical criteria to be applied to parts of nuclear power plant.

The complexity of patient safety reporting systems in UK dentistry.

Science.gov (United States)

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.

A Methodological Framework for Software Safety in Safety Critical Computer Systems

OpenAIRE

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 ...

Seismic safety of building structures of NPP Kozloduy III

International Nuclear Information System (INIS)

Varbanov, G.I.; Kostov, M.K.; Stefanov, D.D.; Kaneva, A.D.

2005-01-01

In the proposed paper is presented a general summary of the analyses carried out to evaluate the dynamic behavior and to assess the seismic safety of some safety related building structures of NPP Kozloduy. The design seismic loads for the site of Kozloduy NPP has been reevaluated and increased during and after the construction of investigated Units 5 and 6. Deterministic and probabilistic approaches are applied to assess the seismic vulnerability of the investigated structures, taking into account the newly defined seismic excitations. The presented results show sufficient seismic safety for the studied critical structures and good efficiency of the seismic upgrading. The applicability of the investigated structures at sites with some higher seismic activities is discussed. The presented study is dealing mainly with the civil structures of the Reactor building, Turbine hall, Diesel Generator Station and Water Intake Structure. (authors)

Monitoring system of arch bridge for safety network management

Science.gov (United States)

Joo, Bong Chul; Yoo, Young Jun; Lee, Chin Hyung; Park, Ki Tae; Hwang, Yoon Koog

2010-03-01

Korea has constructed the safety management network monitoring test systems for the civil infrastructure since 2006 which includes airport structure, irrigation structure, railroad structure, road structure, and underground structure. Bridges among the road structure include the various superstructure types which are Steel box girder bridge, suspension bridge, PSC-box-girder bridge, and arch bridge. This paper shows the process of constructing the real-time monitoring system for the arch bridge and the measured result by the system. The arch type among various superstructure types has not only the structural efficiency but the visual beauty, because the arch type superstructure makes full use of the feature of curve. The main measuring points of arch bridges composited by curved members make a difference to compare with the system of girder bridges composited by straight members. This paper also shows the method to construct the monitoring system that considers the characteristic of the arch bridge. The system now includes strain gauges and thermometers, and it will include various sensor types such as CCTV, accelerometers and so on additionally. For the long term and accuracy monitoring, the latest optical sensors and equipments are applied to the system.

Risk assessment of computer-controlled safety systems for fusion reactors

International Nuclear Information System (INIS)

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

1983-01-01

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

Study of system safety evaluation on LTO of national project. NISA safety research project on system safety of nuclear power plants

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)

Safety assessment, safety performance indicators at the Paks Nuclear Power Plant

International Nuclear Information System (INIS)

Baji, C.; Vamos, G.; Toth, J.

2001-01-01

The Paks Nuclear Power Plant has been using different methods of safety assessment (event analysis, self-assessment, probabilistic safety analysis), including performance indicators characterizing both operational and safety performance since the early years of operation of the plant. Regarding the safety performance, the indicators include safety system performance, number of scrams, release of radioactive materials, number of safety significant events, industrial safety indicator, etc. The Paks NPP also reports a set of ten indicators to WANO Performance Indicator Programme which, among others, include safety related indicators as well. However, a more systematic approach to structuring and trending safety indicators is needed so that they can contribute to the enhancement of the operational safety. A more comprehensive set of indicators and a systematic evaluation process was introduced in 1996. The performance indicators framework proposed by the IAEA was adapted to Paks in this year to further improve the process. Safety culture assessment and characterizing safety culture is part of the assessment process. (author)

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

Assessing nuclear power plant safety and recovery from earthquakes using a system-of-systems approach

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

Survey and evaluation of inherent safety characteristics and passive safety systems for use in probabilistic safety analyses

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

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)

Operation and safety decision-making support expert system in NPP

International Nuclear Information System (INIS)

Wei Yanhui; Su Desong; Chen Weihua; Zhang Jianbo

2014-01-01

The article first reviewed three operation support systems currently used in NPP: real-time information surveillance system, important equipment surveillance system and plant process control and monitoring system, then presents the structure and function of three expert support sub-systems (intelligent alarm monitoring system, computer-based operating procedure support system, safety information expert decision support system). Finally the article discussed the meaning of a kind of operation decision making support system. (authors)

Reliability of redundant structures of nuclear reactor protection systems

International Nuclear Information System (INIS)

Vojnovic, B.

1983-01-01

In this paper, reliability of various redundant structures of PWR protection systems has been analysed. Structures of reactor tip systems as well as the systems for activation of safety devices have been presented. In all those systems redundancy is achieved by means of so called majority voting logic ('r out of n' structures). Different redundant devices have been compared, concerning probability of occurrence of safe as well as unsafe failures. (author)

SAFE-KBS, Substantiating the safety of systems containing knowledge-based components

International Nuclear Information System (INIS)

Mesa, E.; Jimenez, A.

1998-01-01

The overall objective of the Safe-KBS project is to develop generic development and certification methodologies that allow the introduction of knowledge-based components in safety-related applications. The expert system technology presents a set of features, such as the capability to provide the rationale for its conclusions, that may significantly contribute to the new operation support systems. Nevertheless, the use of this technology in safety-related applications is limited by the lack of recognised methodologies and standards that allow a formal demonstration of the quality and reliability of these systems, as required for obtaining the approval for their use at nuclear power plants. The development methodology is structured in three hierarchical levels: life cycle model, i.e., processes and activities constituting the life cycle, life cycle plans, i.e., tasks, and support packages, i.e., set of techniques and methods to perform certain activities or tasks. The certification methodology consists of a set of certification requirements and a certification scheme for demonstrating the compliance with these requirements. This project was developed within the European framework ESPRIT, with the collaboration of Sextant, Cise, Qualience, Ilog, Computes, DNV and Uninfo. (Author)

Systems Analysis of NASA Aviation Safety Program: Final Report

Science.gov (United States)

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.

Replacement cross-site transfer system project W-058 safety class upgrade summary report

International Nuclear Information System (INIS)

Schlosser, R.L.

1998-01-01

This report evaluates the design of the replacement cross-site transfer system structures, systems, and components for safety related applications as defined in the Tank Waste Remediation Systems Basis for Interim Operations

Comparison, with regard to safety, between a hard-wired reactor protection system and a computerized protection system. Pt. 1

International Nuclear Information System (INIS)

Buettner, W.E.

1976-07-01

The study compares a conventional hard-wired dynamic reactor protection system with a computerized protection system. In the comparison, only the unequivocally safety-oriented protection actions are considered. In the first part, the different structures of both systems and the method of verification for their functional safety will be described. In the second part, the mean unavailability in case of demand for both systems under defined conditions will be determined. (orig.) [de

Reliability assessment of nuclear structural systems

International Nuclear Information System (INIS)

Reich, M.; Hwang, H.

1983-01-01

Reliability assessment of nuclear structural systems has been receiving more emphasis over the last few years. This paper deals with the recent progress made by the Structural Analysis Division of Brookhaven National Laboratory (BNL), in the development of a probability-based reliability analysis methodology for safety evaluation of reactor containments and other seismic category I structures. An important feature of this methodology is the incorporation of finite element analysis and random vibration theory. By utilizing this method, it is possible to evaluate the safety of nuclear structures under various static and dynamic loads in terms of limit state probability. Progress in other related areas, such as the establishment of probabilistic characteristics for various loads and structural resistance, are also described. Results of an application of the methodology to a realistic reinforced concrete containment subjected to dead and live loads, accidental internal pressures and earthquake ground accelerations are presented

Structural mode significance using INCA. [Interactive Controls Analysis computer program

Science.gov (United States)

Bauer, Frank H.; Downing, John P.; Thorpe, Christopher J.

1990-01-01

Structural finite element models are often too large to be used in the design and analysis of control systems. Model reduction techniques must be applied to reduce the structural model to manageable size. In the past, engineers either performed the model order reduction by hand or used distinct computer programs to retrieve the data, to perform the significance analysis and to reduce the order of the model. To expedite this process, the latest version of INCA has been expanded to include an interactive graphical structural mode significance and model order reduction capability.

Does the concept of safety culture help or hinder systems thinking in safety?

Science.gov (United States)

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.

The aviation safety reporting system

Science.gov (United States)

Reynard, W. D.

1984-01-01

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

Multi-dimensional database design and implementation of dam safety monitoring system

Directory of Open Access Journals (Sweden)

Zhao Erfeng

2008-09-01

Full Text Available To improve the effectiveness of dam safety monitoring database systems, the development process of a multi-dimensional conceptual data model was analyzed and a logic design was achieved in multi-dimensional database mode. The optimal data model was confirmed by identifying data objects, defining relations and reviewing entities. The conversion of relations among entities to external keys and entities and physical attributes to tables and fields was interpreted completely. On this basis, a multi-dimensional database that reflects the management and analysis of a dam safety monitoring system on monitoring data information has been established, for which factual tables and dimensional tables have been designed. Finally, based on service design and user interface design, the dam safety monitoring system has been developed with Delphi as the development tool. This development project shows that the multi-dimensional database can simplify the development process and minimize hidden dangers in the database structure design. It is superior to other dam safety monitoring system development models and can provide a new research direction for system developers.

NASA Aviation Safety Reporting System (ASRS)

Science.gov (United States)

Connell, Linda J.

2017-01-01

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

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

Continuous restraint control systems: safety improvement for various occupants

NARCIS (Netherlands)

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

2009-01-01

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

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.

Proposed and existing passive and inherent safety-related structures, systems, and components (building blocks) for advanced light-water reactors

International Nuclear Information System (INIS)

Forsberg, C.W.; Moses, D.L.; Lewis, E.B.; Gibson, R.; Pearson, R.; Reich, W.J.; Murphy, G.A.; Staunton, R.H.; Kohn, W.E.

1989-10-01

A nuclear power plant is composed of many structures, systems, and components (SSCs). Examples include emergency core cooling systems, feedwater systems, and electrical systems. The design of a reactor consists of combining various SSCs (building blocks) into an integrated plant design. A new reactor design is the result of combining old SSCs in new ways or use of new SSCs. This report identifies, describes, and characterizes SSCs with passive and inherent features that can be used to assure safety in light-water reactors. Existing, proposed, and speculative technologies are described. The following approaches were used to identify the technologies: world technical literature searches, world patent searches, and discussions with universities, national laboratories and industrial vendors. 214 refs., 105 figs., 26 tabs

Proposed and existing passive and inherent safety-related structures, systems, and components (building blocks) for advanced light-water reactors

Energy Technology Data Exchange (ETDEWEB)

Forsberg, C.W.; Moses, D.L.; Lewis, E.B.; Gibson, R.; Pearson, R.; Reich, W.J.; Murphy, G.A.; Staunton, R.H.; Kohn, W.E.

1989-10-01

A nuclear power plant is composed of many structures, systems, and components (SSCs). Examples include emergency core cooling systems, feedwater systems, and electrical systems. The design of a reactor consists of combining various SSCs (building blocks) into an integrated plant design. A new reactor design is the result of combining old SSCs in new ways or use of new SSCs. This report identifies, describes, and characterizes SSCs with passive and inherent features that can be used to assure safety in light-water reactors. Existing, proposed, and speculative technologies are described. The following approaches were used to identify the technologies: world technical literature searches, world patent searches, and discussions with universities, national laboratories and industrial vendors. 214 refs., 105 figs., 26 tabs.

Integrating system safety into the basic systems engineering process

Science.gov (United States)

Griswold, J. W.

1971-01-01

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

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

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

Compartmentalized safety coolant injection system

International Nuclear Information System (INIS)

Johnson, F.T.

1983-01-01

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

Safety Design Strategy for the Advanced Test Reactor Emergency Firewater Injection System Replacement Project

International Nuclear Information System (INIS)

Duckwitz, Noel

2011-01-01

In accordance with the requirements of U.S. Department of Energy (DOE) Order 413.3B, 'Program and Project Management for the Acquisition of Capital Assets,' safety must be integrated into the design process for new or major modifications to DOE Hazard Category 1, 2, and 3 nuclear facilities. The intended purpose of this requirement involves the handling of hazardous materials, both radiological and chemical, in a way that provides adequate protection to the public, workers, and the environment. Requirements provided in DOE Order 413.3B and DOE Order 420.1B, 'Facility Safety,' and the expectations of DOE-STD-1189-2008, 'Integration of Safety into the Design Process,' provide for identification of hazards early in the project and use of an integrated team approach to design safety into the facility. This safety design strategy provides the basic safety-in-design principles and concepts that will be used for the Advanced Test Reactor Reliability Sustainment Project. While this project does not introduce new hazards to the ATR, it has the potential for significant impacts to safety-related systems, structures, and components that are credited in the ATR safety basis and are being replaced. Thus the project has been determined to meet the definition of a major modification and is being managed accordingly.

Advanced Range Safety System for High Energy Vehicles

Science.gov (United States)

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.

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)

Assessment of the nuclear installation's safety significant events

International Nuclear Information System (INIS)

Vidican, D.

2005-01-01

This document tries to establish, based on the available documentation, the main steps in development of Assessment of the Events in Nuclear Installations. It takes into account: selection of the safety significant occurrences, establishing the direct cause and contributors as well as the root cause and contributors. Also, the document presents the necessary corrective actions and generic lessons to be learned from the event. The document is based especially on IAEA - ASSET guidelines and DOE root cause analysis Guidance. (author)

Patient Safety Learning Systems: A Systematic Review and Qualitative Synthesis.

Science.gov (United States)

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

System safety engineering analysis handbook

Science.gov (United States)

Ijams, T. E.

1972-01-01

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

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.

78 FR 29392 - Embedded Digital Devices in Safety-Related Systems, Systems Important to Safety, and Items Relied...

Science.gov (United States)

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...

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)

77 FR 70409 - System Safety Program

Science.gov (United States)

2012-11-26

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

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.

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

System safety education focused on industrial engineering

Science.gov (United States)

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

1971-01-01

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

Aspects of safety and reliability for fusion magnet systems first annual report

International Nuclear Information System (INIS)

Powell, J.

1976-01-01

General systems aspects of fusion magnet safety are examined first, followed by specific detailed analyses covering structural, thermal, electrical, and other aspects of fusion magnet safety. The design examples chosen for analysis are illustrative and are not intended to be definitive, since fusion magnet designs are rapidly evolving. Included is a comprehensive collection of design and operating data relating to the safety of existing superconducting magnet systems. The remainder of the overview lists the main conclusions developed from the work to date. These should be regarded as initial steps. Since this study has concentrated on examining potential safety concerns, it may tend to overemphasize the problems of fusion magnets. In fact, many aspects of fusion magnets are well developed and are consistent with good safety practice. A short summary of the findings of this study is given

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

New reactor technology: safety improvements in nuclear power systems.

Science.gov (United States)

Corradini, M L

2007-11-01

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

Ullage Compatible Optical Sensor for Monitoring Safety Significant Malfunctions, Phase II

Data.gov (United States)

National Aeronautics and Space Administration — Significant emphasis has been placed on aircraft fuel tank safety following the TWA Flight 800 accident in July 1996. Upon investigation, the National Transportation...

Structural design concept and static analysis of CANDU spent fuel compact dry storage system

International Nuclear Information System (INIS)

Choi, K. S.; Yang, K. H.; Paek, C. R.; Jung, J. S.; Lee, H. Y.

2003-01-01

In this study, an structural design concept on CANDU spent fuel compact dry storage system MACSTOR/KN-400 module has been established with a view to optimally design the structural members of the system. Design loads, loading combination and structural safety criteria of the module were reviewed assuming W olsung Site. The static analysis of the module showed that compressive stress concentration due to dead load and live load occurred around the center of roof slab. Maximum stress resulted from dead load is about twice as much as the stress from live load, and structural behavior of module caused by wind load was not significant. The static analysis results will have influence on the reinforcement bar design of structural members with other structural analyses

Review guidelines for software languages for use in nuclear power plant safety systems: Final report. Revision 1

Energy Technology Data Exchange (ETDEWEB)

Hecht, M.; Decker, D.; Graff, S.; Green, W.; Lin, D.; Dinsmore, G.; Koch, S. [SoHaR, Inc., Beverly Hills, CA (United States)

1997-10-01

Guidelines for the programming and auditing of software written in high level languages for safety systems are presented. The guidelines are derived from a framework of issues significant to software safety which was gathered from relevant standards and research literature. Language-specific adaptations of these guidelines are provided for the following high level languages: Ada83 and Ada95; C and C++; International Electrochemical Commission (IEC) Standard 1131-3 Ladder Logic, Sequential Function Charts, Structured Text, and Function Block Diagrams; Pascal; and PL/M. Appendices to the report include a tabular summary of the guidelines and additional information on selected languages.

Review guidelines for software languages for use in nuclear power plant safety systems: Final report. Revision 1

International Nuclear Information System (INIS)

Hecht, M.; Decker, D.; Graff, S.; Green, W.; Lin, D.; Dinsmore, G.; Koch, S.

1997-10-01

Guidelines for the programming and auditing of software written in high level languages for safety systems are presented. The guidelines are derived from a framework of issues significant to software safety which was gathered from relevant standards and research literature. Language-specific adaptations of these guidelines are provided for the following high level languages: Ada83 and Ada95; C and C++; International Electrochemical Commission (IEC) Standard 1131-3 Ladder Logic, Sequential Function Charts, Structured Text, and Function Block Diagrams; Pascal; and PL/M. Appendices to the report include a tabular summary of the guidelines and additional information on selected languages

Design of marine structures with improved safety for environment

International Nuclear Information System (INIS)

Klanac, Alan; Varsta, Petri

2011-01-01

The paper describes a method for design of marine structures with increased safety for environment, considering also the required investment costs as well as the aspects of risk distribution onto the maritime stakeholders. Practically, the paper seeks to answer what is the optimal amount that should be invested into certain safety measure for any given vessel. Due to the uneven distribution of risk, as well as the differing impact of costs emerging from safety improvements, stakeholders experience conflicting ranking of alternatives. To solve this multi-stakeholder decision-making problem, in which each stakeholder is a decision-maker, the method applies concepts of group decision-making theory, namely the Game Theory. The method fosters axiomatic definition of the optimum solution, arguing that the solution, or the final selected design, should satisfy the non-dominance, efficiency, and fairness. These three are thoroughly discussed in terms of structural design, especially the latter. Considering the coupling of environmental risk and structural design, the method also builds on the preference structure of four maritime stakeholders: yards, owners, oil receivers and the public, who either share the risks or directly influence structural design. Method is presented on a practical study of structural design of a tanker with a crashworthy side structure that is capable of reducing the risk of collision. The outcome of this study outlines a number of possibilities for successful improvement of tanker safety that can benefit, concurrently, all maritime stakeholders.

Importance of Decision Support Systems About Food Safety in Raw Milk Production

Directory of Open Access Journals (Sweden)

Ecem Akan

2015-12-01

Full Text Available In raw milk production decision support systems for control of food safety hazards has not been developed but main points of this system are available. The decision support systems’ elements include data identification at critical points in the milk supply chain, an information management system and data exchange. Decision supports systems has been developed on the basis of these elements. In dairy sector decision support systems are significant for controlling of food safety hazards and preferred by producers. When these systems are implemented in the milk supply chain, it can be prevented unnecessary sampling and analysis. In this article it will be underlined effects of decision support system elements on food safety of raw milk.

Editorial: emerging issues in sociotechnical systems thinking and workplace safety.

Science.gov (United States)

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.

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

International Nuclear Information System (INIS)

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

1982-01-01

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

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

Different design approaches to structural fire safety

DEFF Research Database (Denmark)

Giuliani, Luisa; Budny, I.

2013-01-01

-priori evaluate which design is the safest or the most economical one: a punctual analysis of the different aspects and a comparison of the resulting designs is therefore of interest and is presented in this paper with reference to the case study considered.The third approach refers instead to a performance......-based fire design of the structure(PBFD), where safety goals are explicitly defined and a deeper knowledge of the structural response to fire effects can be achieved, for example with the avail of finite element analyses (FEA). On the other hand, designers can’t follow established procedures when undertaking...... such advanced investigations, which are generally quite complex ones, due to the presence of material degradation and large displacements induced by fire, as well as the possible triggering of local mechanism in the system. An example of advanced investigations for fire design is given in the paper...

Systems Safety and Engineering Division

Data.gov (United States)

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

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

Directory of Open Access Journals (Sweden)

Ai Lin Evelyn Teo

2016-12-01

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

Linking empowering leadership to safety participation in nuclear power plants: a structural equation model.

Science.gov (United States)

Martínez-Córcoles, Mario; Schöbel, Markus; Gracia, Francisco J; Tomás, Inés; Peiró, José M

2012-07-01

Safety participation is of paramount importance in guaranteeing the safe running of nuclear power plants. The present study examined the effects of empowering leadership on safety participation. Based on a sample of 495 employees from two Spanish nuclear power plants, structural equation modeling showed that empowering leadership has a significant relationship with safety participation, which is mediated by collaborative team learning. In addition, the results revealed that the relationship between empowering leadership and collaborative learning is partially mediated by the promotion of dialogue and open communication. The implications of these findings for safety research and their practical applications are outlined. An empowering leadership style enhances workers' safety performance, particularly safety participation behaviors. Safety participation is recommended to detect possible rule inconsistencies or misunderstood procedures and make workers aware of critical safety information and issues. Crown Copyright © 2012. Published by Elsevier Ltd. All rights reserved.

Preliminary study for unified management of CANDU safety codes and construction of database system

International Nuclear Information System (INIS)

Min, Byung Joo; Kim, Hyoung Tae

2003-03-01

It is needed to develop the Graphical User Interface(GUI) for the unified management of CANDU safety codes and to construct database system for the validation of safety codes, for which the preliminary study is done in the first stage of the present work. The input and output structures and data flow of CATHENA and PRESCON2 are investigated and the interaction of the variables between CATHENA and PRESCON2 are identified. Furthermore, PC versions of CATHENA and PRESCON2 codes are developed for the interaction of these codes and GUI(Graphic User Interface). The PC versions are assessed by comparing the calculation results with those by HP workstation or from FSAR(Final Safety Analysis Report). Preliminary study on the GUI for the safety codes in the unified management system are done. The sample of GUI programming is demonstrated preliminarily. Visual C++ is selected as the programming language for the development of GUI system. The data for Wolsong plants, reactor core, and thermal-hydraulic experiments executed in the inside and outside of the country, are collected and classified following the structure of the database system, of which two types are considered for the final web-based database system. The preliminary GUI programming for database system is demonstrated, which is updated in the future work

Product Engineering Class in the Software Safety Risk Taxonomy for Building Safety-Critical Systems

Science.gov (United States)

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

Software system safety

Science.gov (United States)

Uber, James G.

1988-01-01

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

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

Probabilistic frequency variations of structure-soil systems

International Nuclear Information System (INIS)

Hamilton, C.W.; Hadjian, A.H.

1976-01-01

During earthquakes, structure-soil systems act as filters greatly amplifying the response of equipment whose frequencies are at or near their natural frequencies. Thus, the estimation of these structure-soil system frequencies assumes significant importance both for safety and cost. Actual in-situ frequencies of structures differ from calculated frequencies due both to variations in mathematical modelling techniques and to variations of material properties. This paper studies the second source only. This variability is usually gauged by the 'worst case' analyses technique in which extreme high- and low- parameter values are assumed and the associated frequencies are used as upper and lower bounds. This approach is not entirely satisfactory because it does not provide any indication of the probability of these limits being exceeded, of the distribution between these limits, or of the level of conservation introduced into the design process. The present approach provides this additional information. The emphasis in this part is both on developing the methodology and on the results obtained. It covers both the fixed-base structure and the effects of soil-structure interaction. Empirical data on concrete proerties were obtained from previously published results. Much less is known about variability of soil properties, so the soil structure interaction coefficients are assumed to be normally distributed. As data on the variation of soil properties become available, they can be readily incorporated via the methodology developed here. (Auth.)

[Road map for health and safety management systems in healthcare facilities, according to the OHSAS 18001:2007 standard].

Science.gov (United States)

Pugliese, F; Albini, E; Serio, O; Apostoli, P

2011-01-01

The 81/2008 Act has defined a model of a health and safety management system that can contribute to prevent the occupational health and safety risks. We have developed the structure of a health and safety management system model and the necessary tools for its implementation in health care facilities. The realization of a model is structured in various phases: initial review, safety policy, planning, implementation, monitoring, management review and continuous improvement. Such a model, in continuous evolution, is based on the responsibilities of the different corporate characters and on an accurate analysis of risks and involved norms.

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

RATU - Nuclear power plant structural safety research programme

International Nuclear Information System (INIS)

Rintamaa, R.

1992-07-01

Studies on the structural materials in nuclear power plants create the experimental data and background information necessary for the structural integrity assessments of mechanical components. The research is carried out by developing experimental fracture mechanics methods including statistical analysis methods of materials property data, and by studying material ageing and, in particular, mechanisms of material deterioration due to neutron irradiation, corrosion and water chemistry. Besides material studies, new testing methods and sensors for measurement of loading and water chemistry parameters have been developed. The monitoring data obtained in real power plants has been used to simulate more precisely the real environment during laboratory tests. The research on structural analysis has focused on extending and verifying the analysis capabilities for structural assessments of nuclear power plants. A widely applicable system including various computational fracture assessment methods has been created with which different structural problems can be solved reliably and effectively. Research on reliability assessment of maintenance in nuclear power plants is directed to practical case studies on components and structures of safety importance, and to the development of models for maintenance related decision support. A systematic analysis of motor-operated valve has been performed

Evaluation of implementation an Integrated Safety and Preventive Maintenance System for Improving of Safety Indexes

Directory of Open Access Journals (Sweden)

I mohammadfam

2014-03-01

Full Text Available Accident analysis shows that one of the main reasons for accidents is non-integration of maintenance units with safety. Merging these two processes through an integrated system can reduce and or eliminate accidents, diseases, and environmental pollution. These issues lead to improvement in organizational performance, as well. The aim of this study is to design and establish an integrated system for obtaining the aforementioned goal. Integration was carried out at Nirou Moharreke Machine Tools Company via Structured System Analysis & Design Method (SSADM. In order to measure the effectiveness of the system, selected indexes were compared using statistical methods prior and after system establishment. Results show that the accident severity index reduced from 135.46 in 2010, to 43.85 in 2012. Moreover, system effectiveness improved equipment reliability and availability (e.g. reliability of the Pfeiffer Milling machine (P (t>50 increased from 0.89 in 2010, to 0.9 in 2012. This system by forecasting various failures, and planning and designing the required operations for preventing occurrence of these failures, plays an important role in improving safety conditions of equipment, and increasing organizational performance, and is capable of presenting an excellent accident prevention program.

The effects of organizational commitment and structural empowerment on patient safety culture.

Science.gov (United States)

Horwitz, Sujin K; Horwitz, Irwin B

2017-03-20

Purpose The purpose of this paper is to investigate the relationship between patient safety culture and two attitudinal constructs: affective organizational commitment and structural empowerment. In doing so, the main and interaction effects of the two constructs on the perception of patient safety culture were assessed using a cohort of physicians. Design/methodology/approach Affective commitment was measured with the Organizational Commitment Questionnaire, whereas structural empowerment was assessed with the Conditions of Work Effectiveness Questionnaire-II. The abbreviated versions of these surveys were administered to a cohort of 71 post-doctoral medical residents. For the data analysis, hierarchical regression analyses were performed for the main and interaction effects of affective commitment and structural empowerment on the perception of patient safety culture. Findings A total of 63 surveys were analyzed. The results revealed that both affective commitment and structural empowerment were positively related to patient safety culture. A potential interaction effect of the two attitudinal constructs on patient safety culture was tested but no such effect was detected. Research limitations/implications This study suggests that there are potential benefits of promoting affective commitment and structural empowerment for patient safety culture in health care organizations. By identifying the positive associations between the two constructs and patient safety culture, this study provides additional empirical support for Kanter's theoretical tenet that structural and organizational support together helps to shape the perceptions of patient safety culture. Originality/value Despite the wide recognition of employee empowerment and commitment in organizational research, there has still been a paucity of empirical studies specifically assessing their effects on patient safety culture in health care organizations. To the authors' knowledge, this study is the first

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...

ERC Safety and Hygiene Programs functional organization structure and mission statement

International Nuclear Information System (INIS)

Coleman, S.R.

2000-01-01

This document provides a description of the functions, structure, commitments, and goals of the Environmental Restoration Contractor Safety and Hygiene Program. The current structure of the ERC Safety and Hygiene organization is described herein

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

The problem of maintenance of strength, lifetime and safety of the structural components operational NPP from items of a system approach

International Nuclear Information System (INIS)

Getman, A.F.

2005-01-01

The strength of the structural components and pipelines NPP determines largely their lifetime, radiation and nuclear safety. As shows world experience of operation NPP, in a series of cases during operation the local damages, breaks and destructions of elements of the equipment and pipelines emerge. These cases force maintaining organizations to execute on NPP large volumes of the control, repair and modernizing of the equipment and pipelines. These activities require large material inputs (net less than 50% of the cost of operation NPP), however not always are effective and in a series of cases do not allow to reach required of level reliability and safety. The reason of such condition of a problem is that the measures executed on NPP, developed on stretch of long time and under influence of the diverse factors. Besides until now there is no unified scientific methodology of a case study of maintenance of strength, lifetime and safety of the operational equipment and pipelines NPP. The application of a system approach to a problem of maintenance of strength, lifetime and safety of the equipment and pipelines operational NPP allows to consider a problem as a whole, from unified items, to define the most fast and effective paths it of the solution and to decide problems in as much as possible short times at minimum material inputs. The experience of practical application of a system approach at the solution of separate problems of operation NPP has allowed to develop a lot of effective new techniques, technologies and means, which application allows essentially to increase reliability and gamma percentage lifetime of elements NPP, to which they were applied. As an example in the report the outcomes of application of a system approach for maintenance of SG tubes integrity of reactors of a type WWER are adduced. The application of a system approach for all elements of the equipment and pipelines NPP will allow to increase it reliability and safety not less, than on 10

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

Modeling patient safety incidents knowledge with the Categorial Structure method.

Science.gov (United States)

Souvignet, Julien; Bousquet, Cédric; Lewalle, Pierre; Trombert-Paviot, Béatrice; Rodrigues, Jean Marie

2011-01-01

Following the WHO initiative named World Alliance for Patient Safety (PS) launched in 2004 a conceptual framework developed by PS national reporting experts has summarized the knowledge available. As a second step, the Department of Public Health of the University of Saint Etienne team elaborated a Categorial Structure (a semi formal structure not related to an upper level ontology) identifying the elements of the semantic structure underpinning the broad concepts contained in the framework for patient safety. This knowledge engineering method has been developed to enable modeling patient safety information as a prerequisite for subsequent full ontology development. The present article describes the semantic dissection of the concepts, the elicitation of the ontology requirements and the domain constraints of the conceptual framework. This ontology includes 134 concepts and 25 distinct relations and will serve as basis for an Information Model for Patient Safety.

Safety Concepts in Structural Glass Engineering : Towards an Integrated Approach

NARCIS (Netherlands)

Bos, F.P.

2009-01-01

This dissertation proposes the Integrated Approach to Structural Glass Safety, based on four clearly defined element safety properties, damage sensitivity, relative resistance, redundancy, and fracture mode. The Element Safety Diagram (ESD) is introduced to provide an easy-to-read graphical

Significance assessment of small-medium sized reactors

Energy Technology Data Exchange (ETDEWEB)

Kanno, Minoru [Japan Atomic Power Co., Research and Development Dept., Tokyo (Japan)

2002-12-01

Preliminary assessment for deployment of small-medium sized reactor (S and M reactor) as a future option has been conducted at the JAPCO (Japan Atomic Power Company) under the cooperation with the CRIERI (Central Research Institute of Electric Power Industry). Significance of the S and M reactor introduction is listed as follows; lower investment cost, possible siting near demand side, enlarged freedom of siting, shorter transmission line, good compatibility with slow increase of demand and plain explanation of safety using simpler system such as integral type vessel without piping, natural convection core cooling and passive safety system. The deployment of simpler plant system, modular shop fabrication, ship-shell structured building and longer operation period can assure economics comparable with that of a large sized reactor, coping with scale-demerit. Also the S and M reactor is preferable in size for the nuclear heat utilization such as hydrogen production. (T. Tanaka)

Topology optimization for simplified structural fire safety

DEFF Research Database (Denmark)

Madsen, Søren; Lange, Nis P.; Giuliani, Luisa

2016-01-01

Topology optimization is applied in an idealized structural fire safety model, where the minimum compliance problem is constrained by temperature-controlled structural degradation. The constraint ensures a certain structural stiffness after a prescribed time. As this time period is extended......, resulting optimized topologies tend to become thicker or introduce redundant members that can take over when structural parts near the origin of the fire lose their load carrying capability. Hence, the structural degradation model acts as an erosion operator on the topology and indirectly enforces a minimum...

Reliability of containment and safety-related structures

International Nuclear Information System (INIS)

Nessim, M.A.

1995-09-01

A research program on Reliability of Containment and Safety-related Structures has been developed and is described in this document. This program is designed to support AECB's regulatory activities aimed at ensuring the safety of these structures. These activities include evaluating submissions by operators and requesting special assessments when necessary. The results of the proposed research will also be useful in revising and enhancing the CSA design standards for containment and safety-related structures. The process of developing the research program started with an information collection and review phase. The sources of information included C-FER's previous work in the area, various recent research publications, regulatory documents and relevant design standards, and a detailed discussion with AECB staff. The second step was to outline the process of reliability evaluation, and identify the required models and parameters. Comparison between the required and available information was used to identify gaps in the state-of-the-art, and the research program was designed to fill these gaps. The program is organized in four major topics, namely: development of an approach for reliability analysis; compilation and development of the required analysis tools; application to specific problems related to design, assessment, maintenance and testing of structures; and testing and validation. It is suggested that the program should be supported by an on-going process of communication and consultation between AECB staff and industry experts. This will lend credibility to the results and facilitate their future application. (author). 1 fig

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)

Structural equation model to investigate the dimensions influencing safety culture improvement in construction sector: A case in Indonesia

Science.gov (United States)

Machfudiyanto, Rossy Armyn; Latief, Yusuf; Yogiswara, Yoko; Setiawan, R. Mahendra Fitra

2017-06-01

In facing the ASEAN Economic Community, the level of prevailing working accidents becomes one of the competitiveness factors among the companies. A construction industry is one of the industries prone to high level of accidents. Improving the safety record will not be completely effective unless the occupational safety and healthy culture is enhanced. The aim of this research was to develop a model and to conduct empirical investigation on the relationships among the dimensions of construction occupational safety culture. This research used the structural equation model as a means to examine the hypothesis of positive relationships between dimensions and objectives. The method used in this research was questionnaire survey which was distributed to the respondents from construction companies in a state-owned enterprise in Indonesia. Moreover, there were dimensions of occupational safety culture that was established, such as leadership, behavior, value, strategy, policy, process, employee, safety cost, and contract system. The results of this study indicated that all dimensions were significant and inter-related in forming the safety culture. The result of R2 yielded the safety performance was 54%, which means it was in low category and evaluation of policies on construction companies was required in addressing the issue of working accidents.

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

'Shelter' object safety. Structural aspects

International Nuclear Information System (INIS)

Krivosheev, P.I.; Nemchinov, Yu.I.; Bambura, A.N.; Sokolov, A.P.; Shenderovich, V.Ya.; Vasyagin, R.V.; Klyuchnikov, A.A.; Shcherbin, V.N.; Rud'ko, V.M.; Tokarevskij, V.V.; Belousov, E.L.; Khejger, D.; Gorodetskij, L.

2001-01-01

In 1997 the group of the international experts including Ukrainian organizations and 'Shelter' Object ChNPP are developed the plan of SO transformation into ecological safe status (plan SIP). The realization of the plan was carried out on basis of international tender. The results of structural aspects of SO safety and it transformation into ecological safe state (ESS) are resented in this report

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

Science.gov (United States)

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.

Validation study on reliability analysis of main safety system in Nuclear Power Plant

Energy Technology Data Exchange (ETDEWEB)

Cho, Nam Jin; Cho, Chang Keun; Kim, Yong Hui; Kim, Tae Hyeong; Hong, Seo Kee; Park, Keon Woo; Park, Chang Jea [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of); Cheong, Woo Sik [Institute for Advanced Engineering, Yongin (Korea, Republic of); Park, Moon Kyu [KEPRI, Taejon (Korea, Republic of)

1993-12-15

The scope and contents of this validation study are to review the design changes of the four main safety systems in Wolsong 2/3/4 Nuclear Power Plants, to review the consideration of the above design changes in the AECL reports, the structure of fault trees, and the data base used in the quantification of the fault trees, to quantify the unavailabilities of main safety systems and check them if they meet the requirements, and to recommend desirable design changes in the emergency core cooling system to reduce the unavailability.

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.)

Integrated piping structural analysis system

International Nuclear Information System (INIS)

Motoi, Toshio; Yamadera, Masao; Horino, Satoshi; Idehata, Takamasa

1979-01-01

Structural analysis of the piping system for nuclear power plants has become larger in scale and in quantity. In addition, higher quality analysis is regarded as of major importance nowadays from the point of view of nuclear plant safety. In order to fulfill to the above requirements, an integrated piping structural analysis system (ISAP-II) has been developed. Basic philosophy of this system is as follows: 1. To apply the date base system. All information is concentrated. 2. To minimize the manual process in analysis, evaluation and documentation. Especially to apply the graphic system as much as possible. On the basis of the above philosophy four subsystems were made. 1. Data control subsystem. 2. Analysis subsystem. 3. Plotting subsystem. 4. Report subsystem. Function of the data control subsystem is to control all information of the data base. Piping structural analysis can be performed by using the analysis subsystem. Isometric piping drawing and mode shape, etc. can be plotted by using the plotting subsystem. Total analysis report can be made without the manual process through the reporting subsystem. (author)

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.

77 FR 11120 - Patient Safety Organizations: Voluntary Relinquishment From UAB Health System Patient Safety...

Science.gov (United States)

2012-02-24

... Organizations: Voluntary Relinquishment From UAB Health System Patient Safety Organization AGENCY: Agency for... notification of voluntary relinquishment from the UAB Health System Patient Safety Organization of its status as a Patient Safety Organization (PSO). The Patient Safety and Quality Improvement Act of 2005...

Significance of coast down time on safety and availability of a pool type fast breeder reactor

International Nuclear Information System (INIS)

Natesan, K.; Velusamy, K.; Selvaraj, P.; Chellapandi, P.

2015-01-01

Highlights: • Plant dynamics studies for quantifying the benefits of flow coast down time. • Establishment of minimum flow coast down time required for safety. • Assessment of influence of flow coast down on enhancing plant availability. • Synthesis of thermo mechanical benefits of flow coast down time on component design. - Abstract: Plant dynamic investigation towards establishing the influence of flow coast down time of primary and secondary sodium systems on safety and availability of plant has been carried out based on one dimensional analysis. From safety considerations, a minimum flow coast down time for primary sodium circuit is essential to be provided to limit the consequences of loss of flow event within allowable limits. Apart from safety benefits, large primary coast down time also improves plant availability by the elimination of reactor SCRAM during short term power failure events. Threshold values of SCRAM parameters also need optimization. By suitably selecting the threshold values for SCRAM parameters, significant reduction in the inertia of pumping systems can be derived to obtain desirable results on plant availability. With the optimization of threshold values and primary flow coast down behaviour equivalent to a halving time of 8 s, there is a possibility to eliminate reactor SCRAM during short term power failure events extending up to 0.75 s duration. Benefits of secondary flow halving on reducing transient thermal loading on components have also been investigated and mixed effects have been observed

Parametric study for the fire safety design of steel structures

DEFF Research Database (Denmark)

Aiuti, Riccardo; Giuliani, Luisa

2013-01-01

the considered time of fire exposure. A deeper knowledge on the failure mode of steel structure is however important in order to ensure the safety of the people and properties outside the building. Aim of this paper is to analyze the behaviour of single elements, sub-assemblies and frames exposed to fire...... or hindered thermal expansion induced on the element by the rest of the structure. Nevertheless, restrained thermal expansion is known to significantly affect the behaviour of steel structures in fire, and the compliance with a prescribed resistance class doesn’t ensure the integrity of the building after...... and find out the basic collapse mechanisms of structural elements in fire conditions, considering the rest of the construction with appropriate constraints. The analysis is carried out taking into account material and geometrical nonlinearities as well as the degradation of steel properties at high...

Assessment and management of aging of nuclear power plant safety-related structures

International Nuclear Information System (INIS)

Naus, D.J.; Graves, H.L. III; Ellingwood, B.R.

2003-01-01

Background information and data have been developed for improving existing and developing new methods to assist in quantifying the effects of age-related degradation on the performance of nuclear power plant (NPP) safety-related structures. Factors that can lead to age-related degradation of safety-related structures are identified and their manifestations described. Current regulatory testing and inspection requirements are reviewed and a summary of degradation experience presented. Techniques commonly used to inspect NPP concrete structures to assess and quantify age-related degradation are summarized. An approach for conduct of condition assessments of structures in NPPs is presented. Criteria, based primarily on visual indications, are provided for use in classification and assessment of concrete degradation. Materials and techniques for repair of degraded structures are noted and guidance provided on repair options available for various forms of degradation. A probabilistic methodology for condition assessment and reliability-based life prediction has been developed and applied to structures subject to combinations of structural load processes and to structural systems. The methodology has also been used to investigate optimization of in-service inspection and maintenance strategies to maintain failure probability below a specified target value as well as to minimize costs. Fragility assessments involving analytical solutions and finite-element methods have been utilized to predict the effect of aging degradation on structural component performance. (author)

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.

Safety margins associated with containment structures under dynamic loading

International Nuclear Information System (INIS)

Lu, S.C.

1978-01-01

A technical basis for assessing the true safety margins of containment structures involved with MARK I boiling water reactor reevaluation activities is presented. It is based on the results of a plane-strain, large displacement, elasto-plastic, finite-element analysis of a thin cylindrical shell subjected to external and internal pressure pulses. An analytical procedure is presented for estimating the ultimate load capacity of the thin shell structure, and subsequently, for quantifying the design margins of safety for the type of loads under consideration. For defining failure of structures, a finite strain failure criterion is derived that accounts for multiaxiality effects

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.

Safety-Critical Java for Embedded Systems

DEFF Research Database (Denmark)

Rios Rivas, Juan Ricardo

for Java aims at providing a reduced set of the Java programming language that can be used for systems that need to be certified at the highest levels of criticality. Safety-critical Java (SCJ) restricts how a developer can structure an application by providing a specific programming model...... and by restricting the set of methods and libraries that can be used. Furthermore, its memory model do not use a garbage-collected heap but scoped memories. In this thesis we examine the use of the SCJ specification through an implementation in a time-predictable, FPGA-based Java processor. The specification is now...

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．

Evaluation of food safety management systems in Serbian dairy industry

Directory of Open Access Journals (Sweden)

Igor Tomašević

2016-01-01

implementation was high, either as standalone food safety system or incorporated in the ISO 22000 and its benefits to the dairy industry in Serbia were significant.

On the classification of structures, systems and components of nuclear research and test reactors

International Nuclear Information System (INIS)

Mattar Neto, Miguel

2009-01-01

The classification of structures, systems and components of nuclear reactors is a relevant issue related to their design because it is directly associated with their safety functions. There is an important statement regarding quality standards and records that says Structures, systems, and components important to safety shall be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions to be performed. The definition of the codes, standards and technical requirements applied to the nuclear reactor design, fabrication, inspection and tests may be seen as the main result from this statement. There are well established guides to classify structures, systems and components for nuclear power reactors such as the Pressurized Water Reactors but one can not say the same for nuclear research and test reactors. The nuclear reactors safety functions are those required to the safe reactor operation, the safe reactor shutdown and continued safe conditions, the response to anticipated transients, the response to potential accidents and the control of radioactive material. So, it is proposed in this paper an approach to develop the classification of structures, systems and components of these reactors based on their intended safety functions in order to define the applicable set of codes, standards and technical requirements. (author)

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

Nuclear Power Safety Reporting System. Final evaluation results

International Nuclear Information System (INIS)

Finlayson, F.C.; Newton, R.D.

1986-02-01

This document presents the results of a study conducted by the US Nuclear Regulatory Commission of an unobtrusive, voluntary, anonymous third-party managed, nonpunitive human factors data gathering system (the Nuclear power Safety Reporting System - NPSRS) for the nuclear electric power production industry. The data to be gathered by the NPSRS are intended for use in identifying and quantifying the factors that contribute to the occurrence of significant safety incidents involving humans in nuclear power plants. The NPSRS has been designed to encourage participation in the System through guarantees of reporter anonymity provided by a third-party organization that would be responsible for NPSRS management. As additional motivation to reporters for contributing data to the NPSRS, conditional waivers of NRC disciplinary action would be provided to individuals. These conditional waivers of immunity would apply to potential violations of NRC regulations that might be disclosed through reports submitted to the System about inadvertent, noncriminal incidents in nuclear plants. This document summarizes the overall results of the study of the NPSRS concept. In it, a functional description of the NPSRS is presented together with a review and assessment of potential problem areas that might be met if the System were implemented. Conclusions and recommendations resulting from the study are also presented. A companion volume (NUREG/CR-4133, Nuclear Power Safety Reporting System: Implementation and Operational Specifications'') presented in detail the elements, requirements, forms, and procedures for implementing and operating the System. 13 refs

System safety education focused on system management

Science.gov (United States)

Grose, V. L.

1971-01-01

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

Maintenance of civil engineering structures important to safety of Nuclear Power Plants

International Nuclear Information System (INIS)

2002-03-01

Civil engineering structures in nuclear installations form an important feature having implications to safety performance of these installations. This safety standard is written to specify the objectives and minimum requirements for the design of civil engineering buildings/structures that are to be fulfilled to provide adequate assurance for safety of nuclear installations in India

Safety Management System in Croatia Control Ltd.

OpenAIRE

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

2009-01-01

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

Light water reactor sequence timing: its significance to probabilistic safety assessment modeling

International Nuclear Information System (INIS)

Bley, D.C.; Buttemer, D.R.; Stetkar, J.W.

1988-01-01

This paper examines event sequence timing in light water reactor plants from the viewpoint of probabilistic safety assessment (PSA). The analytical basis for the ideas presented here come primarily from the authors' work in support of more than 20 PSA studies over the past several years. Timing effects are important for establishing success criteria for support and safety system response and for identifying the time available for operator recovery actions. The principal results of this paper are as follows: 1. Analysis of event sequence timing is necessary for meaningful probabilistic safety assessment - both the success criteria for systems performance and the probability of recovery are tightly linked to sequence timing. 2. Simple engineering analyses based on first principles are often sufficient to provide adequate resolution of the time available for recovery of PSA scenarios. Only those parameters that influence sequence timing and its variability and uncertainty need be examined. 3. Time available for recovery is the basic criterion for evaluation of human performance, whether time is an explicit parameter of the operator actions analysis or not. (author)

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.

Application of Structure Monitoring Systems to the Assessment of the Behaviour of Bridges in Mining Areas

Science.gov (United States)

Parkasiewicz, Beata; Kadela, Marta; Bętkowski, Piotr; Sieńko, Rafał; Bednarski, Łukasz

2017-10-01

Structure monitoring systems are increasingly used to assess the technical condition and improve the safety of structures. Monitoring the structural behaviour becomes necessary in the case of structures located in areas with complicated ground conditions. Due to the risk of failures and the resulting economic and non-material costs, monitoring should be in particular applied to linear structures, including railways, tramlines, motorways and expressways, as well as related facilities (e.g. bridges). Monitoring shall consist in regular observations, measurements and documenting all significant data during construction, after its completion and during usage, and in analysing and evaluating the results. This paper presents the application of structure monitoring systems to the assessment of the behaviour of bridges exposed to the impact of mining operations.

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

A Case Study of Dynamic Response Analysis and Safety Assessment for a Suspended Monorail System.

Science.gov (United States)

Bao, Yulong; Li, Yongle; Ding, Jiajie

2016-11-10

A suspended monorail transit system is a category of urban rail transit, which is effective in alleviating traffic pressure and injury prevention. Meanwhile, with the advantages of low cost and short construction time, suspended monorail transit systems show vast potential for future development. However, the suspended monorail has not been systematically studied in China, and there is a lack of relevant knowledge and analytical methods. To ensure the health and reliability of a suspended monorail transit system, the driving safety of vehicles and structure dynamic behaviors when vehicles are running on the bridge should be analyzed and evaluated. Based on the method of vehicle-bridge coupling vibration theory, the finite element method (FEM) software ANSYS and multi-body dynamics software SIMPACK are adopted respectively to establish the finite element model for bridge and the multi-body vehicle. A co-simulation method is employed to investigate the vehicle-bridge coupling vibration for the transit system. The traffic operation factors, including train formation, track irregularity and tire stiffness, are incorporated into the models separately to analyze the bridge and vehicle responses. The results show that the coupling of dynamic effects of the suspended monorail system between vehicle and bridge are significant in the case studied, and it is strongly suggested to take necessary measures for vibration suppression. The simulation of track irregularity is a critical factor for its vibration safety, and the track irregularity of A-level road roughness negatively influences the system vibration safety.

A Case Study of Dynamic Response Analysis and Safety Assessment for a Suspended Monorail System

Directory of Open Access Journals (Sweden)

Yulong Bao

2016-11-01

Full Text Available A suspended monorail transit system is a category of urban rail transit, which is effective in alleviating traffic pressure and injury prevention. Meanwhile, with the advantages of low cost and short construction time, suspended monorail transit systems show vast potential for future development. However, the suspended monorail has not been systematically studied in China, and there is a lack of relevant knowledge and analytical methods. To ensure the health and reliability of a suspended monorail transit system, the driving safety of vehicles and structure dynamic behaviors when vehicles are running on the bridge should be analyzed and evaluated. Based on the method of vehicle-bridge coupling vibration theory, the finite element method (FEM software ANSYS and multi-body dynamics software SIMPACK are adopted respectively to establish the finite element model for bridge and the multi-body vehicle. A co-simulation method is employed to investigate the vehicle-bridge coupling vibration for the transit system. The traffic operation factors, including train formation, track irregularity and tire stiffness, are incorporated into the models separately to analyze the bridge and vehicle responses. The results show that the coupling of dynamic effects of the suspended monorail system between vehicle and bridge are significant in the case studied, and it is strongly suggested to take necessary measures for vibration suppression. The simulation of track irregularity is a critical factor for its vibration safety, and the track irregularity of A-level road roughness negatively influences the system vibration safety.

Institutional failure: are safety management systems the answer?

Energy Technology Data Exchange (ETDEWEB)

Waddington, J.G.; Lafortune, J.F. [International Safety Research, Ottawa, Ontario (Canada); Duffey, R.B. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

2009-07-01

In spite of an overwhelming number of safety management programs, incidents and accidents that could seemingly, in hindsight, have been prevented, still occur. Institutional failure is seen as a major contributor in almost all cases. With the anticipated significant increase in the number of nuclear plants around the world, a drastic step in the way we manage safety is deemed essential to further reduce the currently already very low rate of accidents to levels that will not cause undue public concern and threaten the success of the nuclear 'renaissance'. To achieve this, many industries have already started implementing a Safety Management System (SMS) approach, aimed at harmonizing, rationalizing and integrating management processes, safety culture and operational risk assessment. This paper discusses the origins and the nature of SMS based in part on the experience of the aviation industry, and shows how SMS is poised to be the next generation in the way the nuclear industry manages safety. It also discusses the need for better direct measures of risk to demonstrate the success of SMS implementation. (author)

Institutional failure: are safety management systems the answer?

International Nuclear Information System (INIS)

Waddington, J.G.; Lafortune, J.F.; Duffey, R.B.

2009-01-01

In spite of an overwhelming number of safety management programs, incidents and accidents that could seemingly, in hindsight, have been prevented, still occur. Institutional failure is seen as a major contributor in almost all cases. With the anticipated significant increase in the number of nuclear plants around the world, a drastic step in the way we manage safety is deemed essential to further reduce the currently already very low rate of accidents to levels that will not cause undue public concern and threaten the success of the nuclear 'renaissance'. To achieve this, many industries have already started implementing a Safety Management System (SMS) approach, aimed at harmonizing, rationalizing and integrating management processes, safety culture and operational risk assessment. This paper discusses the origins and the nature of SMS based in part on the experience of the aviation industry, and shows how SMS is poised to be the next generation in the way the nuclear industry manages safety. It also discusses the need for better direct measures of risk to demonstrate the success of SMS implementation. (author)

Use of probabilistic safety assessment in structuring conceptual design of accident mitigation systems

Energy Technology Data Exchange (ETDEWEB)

Nishiura, Hiroshi; Urata, Shigeru; Tsujikura, Yonezo [Kansai Electric Power Co., Inc., Osaka (Japan); Kuroiwa, Katsuya; Fujimoto, Haruo

2000-07-01

When there is an opportunity to develop a new safety design, it should be a rational design that serves its intended purpose while giving due consideration to factors such as reliability, economic efficiency, and others. Therefore, we have aimed to establish a methodical conceptual design process for accident mitigation systems as part of the core cooling system. In this consideration, we have proposed a process made up of 4 steps and have confirmed that the PSA method can be used as a tool in this process. (author)

Use of probabilistic safety assessment in structuring conceptual design of accident mitigation systems

International Nuclear Information System (INIS)

Nishiura, Hiroshi; Urata, Shigeru; Tsujikura, Yonezo; Kuroiwa, Katsuya; Fujimoto, Haruo

2000-01-01

When there is an opportunity to develop a new safety design, it should be a rational design that serves its intended purpose while giving due consideration to factors such as reliability, economic efficiency, and others. Therefore, we have aimed to establish a methodical conceptual design process for accident mitigation systems as part of the core cooling system. In this consideration, we have proposed a process made up of 4 steps and have confirmed that the PSA method can be used as a tool in this process. (author)

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.

Science.gov (United States)

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

2017-07-01

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

Track 2: business outlook and the significance of safety - safety of nuclear power reactor installations in a deregulated environment. Panel Discussion

International Nuclear Information System (INIS)

Schultz, Stephen P.; Floyd, Stephen; Berkow, Herbert N.; Quinn, Edward L.; Hagen, Ronald E.; Esselman, Thomas C.

2001-01-01

Full text of publication follows: Deregulation in the nuclear industry sets the stage for a new production environment with cost and profit-based competition. This not only affects the commercial reactor sector but also is a key element of the U.S. Department of Energy's (DoE's) strategy for facility deployment and commercialization. This panel is designed to explore how these transitions are being made with the assurance of the principle of safety in operations. Panelists have been assembled from industry, the U.S. Nuclear Regulatory Commission (NRC), and DOE to explore these challenges and opportunities in the deregulated operational environment from a variety of perspectives. Presentations by each panelist will be followed by a panel session that will explore both technical and process elements of these major transitions. Success in a deregulated environment requires both safety and economic operation of facilities. With respect to each of these attributes, competition may be expected to result in substantially higher levels of performance. This panel will examine how these expectations are likely to be set, how they will be measured, and what will constitute success. The baseline for evaluation begins with assessment of past performance. The panelists have examined this for both the commercial nuclear reactor sector, as well as for the DOE complex. The Nuclear Energy Institute and the NRC have been working over the past few years on a program to establish new performance indicators for monitoring safety of reactor operation. The session participants will first examine the effectiveness of this system as it is designed to measure performance in today's environment. Second, the panelists will project the capabilities and attributes of this system in monitoring safety in transition to a deregulated, consolidated market. Will these systems have the capability to identify trends that will be significant in this new environment? Will today's system and metrics be

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

System safety engineering in the development of advanced surface transportation vehicles

Science.gov (United States)

Arnzen, H. E.

1971-01-01

Applications of system safety engineering to the development of advanced surface transportation vehicles are described. As a pertinent example, the paper describes a safety engineering efforts tailored to the particular design and test requirements of the Tracked Air Cushion Research Vehicle (TACRV). The test results obtained from this unique research vehicle provide significant design data directly applicable to the development of future tracked air cushion vehicles that will carry passengers in comfort and safety at speeds up to 300 miles per hour.

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

Long-Term Structural Health Monitoring System for a High-Speed Railway Bridge Structure

Directory of Open Access Journals (Sweden)

You-Liang Ding

2015-01-01

Full Text Available Nanjing Dashengguan Bridge, which serves as the shared corridor crossing Yangtze River for both Beijing-Shanghai high-speed railway and Shanghai-Wuhan-Chengdu railway, is the first 6-track high-speed railway bridge with the longest span throughout the world. In order to ensure safety and detect the performance deterioration during the long-time service of the bridge, a Structural Health Monitoring (SHM system has been implemented on this bridge by the application of modern techniques in sensing, testing, computing, and network communication. The SHM system includes various sensors as well as corresponding data acquisition and transmission equipment for automatic data collection. Furthermore, an evaluation system of structural safety has been developed for the real-time condition assessment of this bridge. The mathematical correlation models describing the overall structural behavior of the bridge can be obtained with the support of the health monitoring system, which includes cross-correlation models for accelerations, correlation models between temperature and static strains of steel truss arch, and correlation models between temperature and longitudinal displacements of piers. Some evaluation results using the mean value control chart based on mathematical correlation models are presented in this paper to show the effectiveness of this SHM system in detecting the bridge’s abnormal behaviors under the varying environmental conditions such as high-speed trains and environmental temperature.

Long-Term Structural Health Monitoring System for a High-Speed Railway Bridge Structure

Science.gov (United States)

Wu, Lai-Yi

2015-01-01

Nanjing Dashengguan Bridge, which serves as the shared corridor crossing Yangtze River for both Beijing-Shanghai high-speed railway and Shanghai-Wuhan-Chengdu railway, is the first 6-track high-speed railway bridge with the longest span throughout the world. In order to ensure safety and detect the performance deterioration during the long-time service of the bridge, a Structural Health Monitoring (SHM) system has been implemented on this bridge by the application of modern techniques in sensing, testing, computing, and network communication. The SHM system includes various sensors as well as corresponding data acquisition and transmission equipment for automatic data collection. Furthermore, an evaluation system of structural safety has been developed for the real-time condition assessment of this bridge. The mathematical correlation models describing the overall structural behavior of the bridge can be obtained with the support of the health monitoring system, which includes cross-correlation models for accelerations, correlation models between temperature and static strains of steel truss arch, and correlation models between temperature and longitudinal displacements of piers. Some evaluation results using the mean value control chart based on mathematical correlation models are presented in this paper to show the effectiveness of this SHM system in detecting the bridge's abnormal behaviors under the varying environmental conditions such as high-speed trains and environmental temperature. PMID:26451387

Overall System Description and Safety Characteristics of Prototype Gen IV Sodium Cooled Fast Reactor in Korea

Directory of Open Access Journals (Sweden)

Jaewoon Yoo

2016-10-01

Full Text Available The Prototype Gen IV sodium cooled fast reactor (PGSFR has been developed for the last 4 years, fulfilling the technology demonstration of the burning capability of transuranic elements included in light water reactor spent nuclear fuel. The PGSFR design has been focused on the robustness of safety systems by enhancing inherent safety characteristics of metal fuel and strengthening passive safety features using natural circulation and thermal expansion. The preliminary safety information document as a major outcome of the first design phase of PGSFR development was issued at the end of 2015. The project entered the second design phase at the beginning of 2016. This paper summarizes the overall structures, systems, and components of nuclear steam supply system and safety characteristics of the PGSFR. The research and development activities to demonstrate the safety performance are also briefly introduced in the paper.

Overall system description and safety characteristics of Prototype Gen IV Sodium Cooled Fast Reactor in Korea

International Nuclear Information System (INIS)

Yoo, Jae Woon; Chang, Jin Wook; Lim, Jae Yong; Cheon, Jin Sik; Lee, Tae Ho; Kim, Sung Kyun; Lee, Kwi Lim; Joo, Hyung Kook

2016-01-01

The Prototype Gen IV sodium cooled fast reactor (PGSFR) has been developed for the last 4 years, fulfilling the technology demonstration of the burning capability of transuranic elements included in light water reactor spent nuclear fuel. The PGSFR design has been focused on the robustness of safety systems by enhancing inherent safety characteristics of metal fuel and strengthening passive safety features using natural circulation and thermal expansion. The preliminary safety information document as a major outcome of the first design phase of PGSFR development was issued at the end of 2015. The project entered the second design phase at the beginning of 2016. This paper summarizes the overall structures, systems, and components of nuclear steam supply system and safety characteristics of the PGSFR. The research and development activities to demonstrate the safety performance are also briefly introduced in the paper

Construction of Traceability System for Quality Safety of Cereal and Oil Products

Science.gov (United States)

Zheng, Huoguo; Liu, Shihong; Meng, Hong; Hu, Haiyan

After several significant food safety incident, global food industry and governments in many countries are putting increasing emphasis on establishment of food traceability systems. Food traceability has become an effective way in food quality and safety management. The traceability system for quality safety of cereal and oil products was designed and implemented with HACCP and FMECA method, encoding, information processing, and hardware R&D technology etc, according to the whole supply chain of cereal and oil products. Results indicated that the system provide not only the management in origin, processing, circulating and consuming for enterprise, but also tracing service for customers and supervisor by means of telephone, internet, SMS, touch machine and mobile terminal.

Ageing of significant to safety structure elements of nuclear power plants

International Nuclear Information System (INIS)

Maksimovas, G.; Ramanauskiene, A.; Ziliukas, A.

1999-01-01

The paper analyzes the ageing problems of structure elements in nuclear power plants. The standard documents and principal parts of the ageing evaluation program are presented. The ageing evaluation model is being worked out and degradation mechanisms of different atomic reactor materials are being compared. (author)

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

Science.gov (United States)

Guo, Jingjing

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

Is Model-Based Development a Favorable Approach for Complex and Safety-Critical Computer Systems on Commercial Aircraft?

Science.gov (United States)

Torres-Pomales, Wilfredo

2014-01-01

A system is safety-critical if its failure can endanger human life or cause significant damage to property or the environment. State-of-the-art computer systems on commercial aircraft are highly complex, software-intensive, functionally integrated, and network-centric systems of systems. Ensuring that such systems are safe and comply with existing safety regulations is costly and time-consuming as the level of rigor in the development process, especially the validation and verification activities, is determined by considerations of system complexity and safety criticality. A significant degree of care and deep insight into the operational principles of these systems is required to ensure adequate coverage of all design implications relevant to system safety. Model-based development methodologies, methods, tools, and techniques facilitate collaboration and enable the use of common design artifacts among groups dealing with different aspects of the development of a system. This paper examines the application of model-based development to complex and safety-critical aircraft computer systems. Benefits and detriments are identified and an overall assessment of the approach is given.

An aspect-oriented approach for designing safety-critical systems

Science.gov (United States)

Petrov, Z.; Zaykov, P. G.; Cardoso, J. P.; Coutinho, J. G. F.; Diniz, P. C.; Luk, W.

The development of avionics systems is typically a tedious and cumbersome process. In addition to the required functions, developers must consider various and often conflicting non-functional requirements such as safety, performance, and energy efficiency. Certainly, an integrated approach with a seamless design flow that is capable of requirements modelling and supporting refinement down to an actual implementation in a traceable way, may lead to a significant acceleration of development cycles. This paper presents an aspect-oriented approach supported by a tool chain that deals with functional and non-functional requirements in an integrated manner. It also discusses how the approach can be applied to development of safety-critical systems and provides experimental results.

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)

Operating experience and systems analysis at Trillo NPP: A program intended for systematic review of plant safety systems to assess design basis requirements compliance

International Nuclear Information System (INIS)

Vega, R. de la

1996-01-01

The program was defined to apply to all plant safety systems and/or systems included in plant Technical Specifications. The goal of the program was to ensure, by systematic design, construction, and commissioning review, the adequacy of safety systems, structures and components to fulfill their safety functions. Also, as a result of the program, it was established that a complete, unambiguous, systematic, design basis definition shall take place. And finally, a complete documental review of the plant design shall result from the program execution

Safety equipment list for the 241-SY-101 RAPID mitigation project

Energy Technology Data Exchange (ETDEWEB)

MORRIS, K.L.

1999-06-29

This document provides the safety classification for the safety (safety class and safety RAPID Mitigation Project. This document is being issued as the project SEL until the supporting authorization basis documentation, this document will be superseded by the TWRS SEL (LMHC 1999), documentation istlralized. Upon implementation of the authorization basis significant) structures, systems, and components (SSCS) associated with the 241-SY-1O1 which will be updated to include the information contained herein.

Safety equipment list for the 241-SY-101 RAPID mitigation project

International Nuclear Information System (INIS)

Morris, K.L.

1999-01-01

This document provides the safety classification for the safety (safety class and safety RAPID Mitigation Project. This document is being issued as the project SEL until the supporting authorization basis documentation, this document will be superseded by the TWRS SEL (LMHC 1999), documentation istlralized. Upon implementation of the authorization basis significant) structures, systems, and components (SSCS) associated with the 241-SY-1O1 which will be updated to include the information contained herein

Safety status system for operating room devices.

Science.gov (United States)

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

2014-01-01

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

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

Renovated Korean nuclear safety and security system: A review and suggestions to successful settlement

Energy Technology Data Exchange (ETDEWEB)

Chung, W. S.; Yun, S. W.; Lee, D. S. [Korea Atomic Energy Research Inst., Dukjin-dong 150, Yusung-gu, Daejon, R.O., 305-353 (Korea, Republic of); Go, D. Y. [Kyung Hee Univ., Kyung Hee daero 26, Dongdaemoon-gu, Seoul, R.O., 130-701 (Korea, Republic of)

2012-07-01

Questions of whether past nuclear regulatory body of Korea is not a proper system to monitor and check the country's nuclear energy policy and utilization have been raised. Moreover, a feeling of insecurity regarding nuclear safety after the nuclear accident in Japan has spread across the public. This has stimulated a renovation of the nuclear safety regime in Korea. The Nuclear Safety and Security Commission (NSSC) was launched on October 26, 2011 as a regulatory body directly under the President in charge of strengthening independence and nuclear safety. This was a meaningful event as the NSSC it is a much more independent regulatory system for Korea. However, the NSSC itself does not guarantee an enhanced public acceptance of the nuclear policy and stable use nuclear energy. This study introduces the new NSSC system and its details in terms of organization structure, appropriateness of specialty, budget stability, and management system. (authors)

Determining the in situ concrete strength of existing structures for assessing their structural safety

NARCIS (Netherlands)

Steenbergen, R.D.J.M.; Vervuurt, A.H.J.M.

2012-01-01

EN 13791 applies when assessing the in situ compressive strength of structures and precast concrete components. According to the code itself, it may be adopted when doubt arises about the compressive strength of a concrete. For assessing the structural safety of existing structures, however, the

Medication Safety Systems and the Important Role of Pharmacists.

Science.gov (United States)

Mansur, Jeannell M

2016-03-01

Preventable medication-related adverse events continue to occur in the healthcare setting. While the Institute of Medicine's To Err is Human, published in 2000, highlighted the prevalence of medical and medication-related errors in patient morbidity and mortality, there has not been significant documented progress in addressing system contributors to medication errors. The lack of progress may be related to the myriad of pharmaceutical options now available and the nuances of optimizing drug therapy to achieve desired outcomes and prevent undesirable outcomes. However, on a broader scale, there may be opportunities to focus on the design and performance of the many processes that are part of the medication system. Errors may occur in the storage, prescribing, transcription, preparation and dispensing, or administration and monitoring of medications. Each of these nodes of the medication system, with its many components, is prone to failure, resulting in harm to patients. The pharmacist is uniquely trained to be able to impact medication safety at the individual patient level through medication management skills that are part of the clinical pharmacist's role, but also to analyze the performance of medication processes and to lead redesign efforts to mitigate drug-related outcomes that may cause harm. One population that can benefit from a focus on medication safety through clinical pharmacy services and medication safety programs is the elderly, who are at risk for adverse drug events due to their many co-morbidities and the number of medications often used. This article describes the medication safety systems and provides a blueprint for creating a foundation for medication safety programs within healthcare organizations. The specific role of pharmacists and clinical pharmacy services in medication safety is also discussed here and in other articles in this Theme Issue.

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

Safety assessment and life time management of nuclear power plants: from reasonable design to reliable structural health monitoring

International Nuclear Information System (INIS)

Savov, K.

2005-01-01

Nowadays the safety of Nuclear Power Plants is becoming more and more significant. Therefore consideration of severe accidents shall be included in both design and operating process of Nuclear Power Plants. In particular ground motion forms one of the important natural hazards. For structural analysis both linear-elastic and non-linear methods are specified by the engineering codes for earthquake resistance design. However, time history analysis is required for investigation of non-linear structural behaviour. Moreover, non-linearities are often caused by the presence of damage. This can be detected by means of structural health monitoring and subsequently system identification. In this paper the advantages of both dynamic time history analysis and damage detection by means of wavelet analysis are discussed. First, the non-linear behaviour of a frame structure due to an artificial earthquake motion is analyzed. A comparison to non-time history techniques is performed as well. Next, failure sources are simulated in the system and then detected by means of a novel wavelet approach. (author)

The CEA/DRN innovative R and D programme: significant studies on passive systems

International Nuclear Information System (INIS)

Fiorini, G.L.; Magistris, F. de; Dumaz, P.; Gautier, G.M.; Pignatel, J.F.; Richard, P.

1999-01-01

The work on passive systems is an essential item of the R and D programme for future reactors; it is structured following four main guidelines: Research, and validation of innovative solutions for the safety functions achievement; An enlarged assessment of the performances of passive systems; Extension of the data base and of the tools qualification range; Assessment of new plant operation modes. After a recalling on the general framework, the paper describes, following these guidelines, the status of the art of the main corresponding programmes within CEA/DRN. (author)

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.

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