WorldWideScience

Sample records for safety analyses developments

  1. SCALE Graphical Developments for Improved Criticality Safety Analyses

    International Nuclear Information System (INIS)

    Barnett, D.L.; Bowman, S.M.; Horwedel, J.E.; Petrie, L.M.

    1999-01-01

    New computer graphic developments at Oak Ridge National Ridge National Laboratory (ORNL) are being used to provide visualization of criticality safety models and calculational results as well as tools for criticality safety analysis input preparation. The purpose of this paper is to present the status of current development efforts to continue to enhance the SCALE (Standardized Computer Analyses for Licensing Evaluations) computer software system. Applications for criticality safety analysis in the areas of 3-D model visualization, input preparation and execution via a graphical user interface (GUI), and two-dimensional (2-D) plotting of results are discussed

  2. Additional methodology development for statistical evaluation of reactor safety analyses

    International Nuclear Information System (INIS)

    Marshall, J.A.; Shore, R.W.; Chay, S.C.; Mazumdar, M.

    1977-03-01

    The project described is motivated by the desire for methods to quantify uncertainties and to identify conservatisms in nuclear power plant safety analysis. The report examines statistical methods useful for assessing the probability distribution of output response from complex nuclear computer codes, considers sensitivity analysis and several other topics, and also sets the path for using the developed methods for realistic assessment of the design basis accident

  3. Distinguishing traffic modes in analysing road safety development.

    NARCIS (Netherlands)

    Stipdonk, H. & Berends, E.

    2008-01-01

    Changes in mobility influence road safety. The effects of safety measures may even be overshadowed by the effects of temporary mobility fluctuations. Usually mobility is corrected for by defining risk as the ratio between fatalities and mobility. Due to lack of sufficient data, mobility is often

  4. Periodic safety analyses

    International Nuclear Information System (INIS)

    Gouffon, A.; Zermizoglou, R.

    1990-12-01

    The IAEA Safety Guide 50-SG-S8 devoted to 'Safety Aspects of Foundations of Nuclear Power Plants' indicates that operator of a NPP should establish a program for inspection of safe operation during construction, start-up and service life of the plant for obtaining data needed for estimating the life time of structures and components. At the same time the program should ensure that the safety margins are appropriate. Periodic safety analysis are an important part of the safety inspection program. Periodic safety reports is a method for testing the whole system or a part of the safety system following the precise criteria. Periodic safety analyses are not meant for qualification of the plant components. Separate analyses are devoted to: start-up, qualification of components and materials, and aging. All these analyses are described in this presentation. The last chapter describes the experience obtained for PWR-900 and PWR-1300 units from 1986-1989

  5. Code development and analyses within the area of transmutation and safety

    International Nuclear Information System (INIS)

    Maschek, W.

    2002-01-01

    A strong code development is going on to meet various demands resulting from the development of dedicated reactors for transmutation and incineration. Code development is concerned with safety codes and general codes needed for assessing scenarios and transmutation strategies. Analyses concentrate on various ADS systems with solid and liquid molten salt fuels. Analyses deal with ADS Demo Plant (5th FP EU) and transmuters with advanced fuels

  6. Probabilistic safety analyses (PSA)

    International Nuclear Information System (INIS)

    1997-01-01

    The guide shows how the probabilistic safety analyses (PSA) are used in the design, construction and operation of light water reactor plants in order for their part to ensure that the safety of the plant is good enough in all plant operational states

  7. Safety analyses of surface facilities

    International Nuclear Information System (INIS)

    Anspach, W.; Baran, A.; Dorst, H.J.; Eifert, B.; Gruen, M.; Behrendt, V.; Berkhan, W.; Dincklage, R.D. v.; Doehler, J.; Bruecher, H.

    1981-01-01

    The investigations were carried out using the example of the Gorleben waste disposal center and the planning documents established for this center. The safety analyses refer to the transport of spent fuel elements, the water-cooled interin storage and the reprocessing stage. Regarding the risk analysis of the technical systems the dynamics of the courses of incidents can be better taken into account by doing a methodical development. (DG) [de

  8. Current regulatory developments concerning the implementation of probabilistic safety analyses for external hazards in Germany

    International Nuclear Information System (INIS)

    Krauss, Matias; Berg, Heinz-Peter

    2014-01-01

    The Federal Ministry for the Environment, Nature Conservation and Nuclear Safety (BMU) initiated in September 2003 a comprehensive program for the revision of the national nuclear safety regulations which has been successfully completed in November 2012. These nuclear regulations take into account the current recommendations of the International Atomic Energy Agency (IAEA) and Western European Nuclear Regulators Association (WENRA). In this context, the recommendations and guidelines of the Nuclear Safety Standards Commission (KTA) and the technical documents elaborated by the respective expert group on Probabilistic Safety Analysis for Nuclear Power Plants (FAK PSA) are being updated or in the final process of completion. A main topic of the revision was the issue external hazards. As part of this process and in the light of the accident at Fukushima and the findings of the related actions resulting in safety reviews of nuclear power plants at national level in Germany and on European level, a revision of all relevant standards and documents has been made, especially the recommendations of KTA and FAK PSA. In that context, not only design issues with respect to events such as earthquakes and floods have been discussed, but also methodological issues regarding the implementation of improved probabilistic safety analyses on this topic. As a result of the revision of the KTA 2201 series 'Design of Nuclear Power Plants against Seismic Events' with their parts 1 to 6, part 1 'Principles' was published as the first standard in November 2011, followed by the revised versions of KTA 2201.2 (soil) and 2201.4 (systems and components) in 2012. The modified the standard KTA 2201.3 (structures) is expected to be issued before the end of 2013. In case of part 5 (seismic instrumentation) and part 6 (post>seismic actions) draft amendments are expected in 2013. The expert group 'Probabilistic Safety Assessments for Nuclear Power Plants' (FAK PSA) is an advisory body of the Federal

  9. Development of advanced methods and related software for human reliability evaluation within probabilistic safety analyses

    International Nuclear Information System (INIS)

    Kosmowski, K.T.; Mertens, J.; Degen, G.; Reer, B.

    1994-06-01

    Human Reliability Analysis (HRA) is an important part of Probabilistic Safety Analysis (PSA). The first part of this report consists of an overview of types of human behaviour and human error including the effect of significant performance shaping factors on human reliability. Particularly with regard to safety assessments for nuclear power plants a lot of HRA methods have been developed. The most important of these methods are presented and discussed in the report, together with techniques for incorporating HRA into PSA and with models of operator cognitive behaviour. Based on existing HRA methods the concept of a software system is described. For the development of this system the utilization of modern programming tools is proposed; the essential goal is the effective application of HRA methods. A possible integration of computeraided HRA within PSA is discussed. The features of Expert System Technology and examples of applications (PSA, HRA) are presented in four appendices. (orig.) [de

  10. Response surface use in safety analyses

    International Nuclear Information System (INIS)

    Prosek, A.

    1999-01-01

    When thousands of complex computer code runs related to nuclear safety are needed for statistical analysis, the response surface is used to replace the computer code. The main purpose of the study was to develop and demonstrate a tool called optimal statistical estimator (OSE) intended for response surface generation of complex and non-linear phenomena. The performance of optimal statistical estimator was tested by the results of 59 different RELAP5/MOD3.2 code calculations of the small-break loss-of-coolant accident in a two loop pressurized water reactor. The results showed that OSE adequately predicted the response surface for the peak cladding temperature. Some good characteristic of the OSE like monotonic function between two neighbor points and independence on the number of output parameters suggest that OSE can be used for response surface generation of any safety or system parameter in the thermal-hydraulic safety analyses.(author)

  11. 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.

  12. Thermal hydraulic reactor safety analyses and experiments

    International Nuclear Information System (INIS)

    Holmstroem, H.; Eerikaeinen, L.; Kervinen, T.; Kilpi, K.; Mattila, L.; Miettinen, J.; Yrjoelae, V.

    1989-04-01

    The report introduces the results of the thermal hydraulic reactor safety research performed in the Nuclear Engineering Laboratory of the Technical Research Centre of Finland (VTT) during the years 1972-1987. Also practical applications i.e. analyses for the safety authorities and power companies are presented. The emphasis is on description of the state-of-the-art know how. The report describes VTT's most important computer codes, both those of foreign origin and those developed at VTT, and their assessment work, VTT's own experimental research, as well as international experimental projects and other forms of cooperation VTT has participated in. Appendix 8 contains a comprehensive list of the most important publications and technical reports produced. They present the content and results of the research in detail.(orig.)

  13. Implementing partnerships in nonreactor facility safety analyses

    International Nuclear Information System (INIS)

    Courtney, J.C.; Perry, W.H.; Phipps, R.D.

    1996-01-01

    Faculty and students from LSU have been participating in nuclear safety analyses and radiation protection projects at ANL-W at INEL since 1973. A mutually beneficial relationship has evolved that has resulted in generation of safety-related studies acceptable to Argonne and DOE, NRC, and state regulatory groups. Most of the safety projects have involved the Hot Fuel Examination Facility or the Fuel Conditioning Facility; both are hot cells that receive spent fuel from EBR-II. A table shows some of the major projects at ANL-W that involved LSU students and faculty

  14. ORNL analyses of AVR performance and safety

    Energy Technology Data Exchange (ETDEWEB)

    Cleveland, J.C.

    1985-01-01

    Because of the high interest in modular High Temperature Reactor performance and safety, a cooperative project has been established involving the Oak Ridge National Laboratory (ORNL), Arbeitsgemeinschaft Versuchs Reaktor GmbH (AVR), and Kernforschungsanlage Juelich GmbH (KFA) in reactor physics, performance and safety. This paper presents initial results of ORNL's examination of a hypothetical depressurized core heatup accident and consideration of how a depressurized core heatup test might be conducted by AVR staff. Also presented are initial analyses of a test involving a reduction in core flow and of a test involving reactivity insertion via control rod withdrawal.

  15. ORNL analyses of AVR performance and safety

    International Nuclear Information System (INIS)

    Cleveland, J.C.

    1985-01-01

    Because of the high interest in modular High Temperature Reactor performance and safety, a cooperative project has been established involving the Oak Ridge National Laboratory (ORNL), Arbeitsgemeinschaft Versuchs Reaktor GmbH (AVR), and Kernforschungsanlage Juelich GmbH (KFA) in reactor physics, performance and safety. This paper presents initial results of ORNL's examination of a hypothetical depressurized core heatup accident and consideration of how a depressurized core heatup test might be conducted by AVR staff. Also presented are initial analyses of a test involving a reduction in core flow and of a test involving reactivity insertion via control rod withdrawal

  16. Reliability and safety analyses under fuzziness

    International Nuclear Information System (INIS)

    Onisawa, T.; Kacprzyk, J.

    1995-01-01

    Fuzzy theory, for example possibility theory, is compatible with probability theory. What is shown so far is that probability theory needs not be replaced by fuzzy theory, but rather that the former works much better in applications if it is combined with the latter. In fact, it is said that there are two essential uncertainties in the field of reliability and safety analyses: One is a probabilistic uncertainty which is more relevant for mechanical systems and the natural environment, and the other is fuzziness (imprecision) caused by the existence of human beings in systems. The classical probability theory alone is therefore not sufficient to deal with uncertainties in humanistic system. In such a context this collection of works will put a milestone in the arguments of probability theory and fuzzy theory. This volume covers fault analysis, life time analysis, reliability, quality control, safety analysis and risk analysis. (orig./DG). 106 figs

  17. Ares I-X Range Safety Analyses Overview

    Science.gov (United States)

    Starr, Brett R.; Gowan, John W., Jr.; Thompson, Brian G.; Tarpley, Ashley W.

    2011-01-01

    Ares I-X was the first test flight of NASA's Constellation Program's Ares I Crew Launch Vehicle designed to provide manned access to low Earth orbit. As a one-time test flight, the Air Force's 45th Space Wing required a series of Range Safety analysis data products to be developed for the specified launch date and mission trajectory prior to granting flight approval on the Eastern Range. The range safety data package is required to ensure that the public, launch area, and launch complex personnel and resources are provided with an acceptable level of safety and that all aspects of prelaunch and launch operations adhere to applicable public laws. The analysis data products, defined in the Air Force Space Command Manual 91-710, Volume 2, consisted of a nominal trajectory, three sigma trajectory envelopes, stage impact footprints, acoustic intensity contours, trajectory turn angles resulting from potential vehicle malfunctions (including flight software failures), characterization of potential debris, and debris impact footprints. These data products were developed under the auspices of the Constellation's Program Launch Constellation Range Safety Panel and its Range Safety Trajectory Working Group with the intent of beginning the framework for the operational vehicle data products and providing programmatic review and oversight. A multi-center NASA team in conjunction with the 45th Space Wing, collaborated within the Trajectory Working Group forum to define the data product development processes, performed the analyses necessary to generate the data products, and performed independent verification and validation of the data products. This paper outlines the Range Safety data requirements and provides an overview of the processes established to develop both the data products and the individual analyses used to develop the data products, and it summarizes the results of the analyses required for the Ares I-X launch.

  18. The role of CFD computer analyses in hydrogen safety management

    International Nuclear Information System (INIS)

    Komen, E.M.J; Visser, D.C; Roelofs, F.; Te Lintelo, J.G.T

    2014-01-01

    The risks of hydrogen release and combustion during a severe accident in a light water reactor have attracted considerable attention after the Fukushima accident in Japan. Reliable computer analyses are needed for the optimal design of hydrogen mitigation systems, like e.g. passive autocatalytic recombiners (PARs), and for the assessment of the associated residual risk of hydrogen combustion. Traditionally, so-called Lumped Parameter (LP) computer codes are being used for these purposes. In the last decade, significant progress has been made in the development, validation, and application of more detailed, three-dimensional Computational Fluid Dynamics (CFD) simulations for hydrogen safety analyses. The objective of the current paper is to address the following questions: - When are CFD computer analyses needed complementary to the traditional LP code analyses for hydrogen safety management? - What is the validation status of the CFD computer code for hydrogen distribution, mitigation, and combustion analyses? - Can CFD computer analyses nowadays be executed in practical and reliable way for full scale containments? The validation status and reliability of CFD code simulations will be illustrated by validation analyses performed for experiments executed in the PANDA, THAI, and ENACCEF facilities. (authors)

  19. NPP Krsko periodic safety review. Safety assessment and analyses

    International Nuclear Information System (INIS)

    Basic, I.; Spiler, J.; Thaulez, F.

    2002-01-01

    Definition of a PSR (Periodic Safety Review) project is a comprehensive safety review of a plant after ten years of operation. The objective is a verification by means of a comprehensive review using current methods that the plant remains safe when judged against current safety objectives and practices and that adequate arrangements are in place to maintain plant safety. The overall goals of the NEK PSR Program are defined in compliance with the basic role of a PSR and the current practice typical for most of the countries in EU. This practice is described in the related guides and good practice documents issued by international organizations. The overall goals of the NEK PSR are formulated as follows: to demonstrate that the plant is as safe as originally intended; to evaluate the actual plant status with respect to aging and wear-out identifying any structures, systems or components that could limit the life of the plant in the foreseeable future, and to identify appropriate corrective actions, where needed; to compare current level of safety in the light of modern standards and knowledge, and to identify where improvements would be beneficial for minimizing deviations at justifiable costs. The Krsko PSR will address the following safety factors: Operational Experience, Safety Assessment, EQ and Aging Management, Safety Culture, Emergency Planning, Environmental Impact and Radioactive Waste.(author)

  20. Plasma-safety assessment model and safety analyses of ITER

    International Nuclear Information System (INIS)

    Honda, T.; Okazaki, T.; Bartels, H.-H.; Uckan, N.A.; Sugihara, M.; Seki, Y.

    2001-01-01

    A plasma-safety assessment model has been provided on the basis of the plasma physics database of the International Thermonuclear Experimental Reactor (ITER) to analyze events including plasma behavior. The model was implemented in a safety analysis code (SAFALY), which consists of a 0-D dynamic plasma model and a 1-D thermal behavior model of the in-vessel components. Unusual plasma events of ITER, e.g., overfueling, were calculated using the code and plasma burning is found to be self-bounded by operation limits or passively shut down due to impurity ingress from overheated divertor targets. Sudden transition of divertor plasma might lead to failure of the divertor target because of a sharp increase of the heat flux. However, the effects of the aggravating failure can be safely handled by the confinement boundaries. (author)

  1. Supporting Fernald Site Closure with Integrated Health and Safety Plans as Documented Safety Analyses

    International Nuclear Information System (INIS)

    Kohler, S.; Brown, T.; Fisk, P.; Krach, F.; Klein, B.

    2004-01-01

    At the Fernald Closure Project (FCP) near Cincinnati, Ohio, environmental restoration activities are supported by Documented Safety Analyses (DSAs) that combine the required project-specific Health and Safety Plans, Safety Basis Requirements (SBRs), and Process Requirements (PRs) into single Integrated Health and Safety Plans (I-HASPs). These integrated DSAs employ Integrated Safety Management methodology in support of simplified restoration and remediation activities that, so far, have resulted in the decontamination and demolition (D and D) of over 200 structures, including eight major nuclear production plants. There is one of twelve nuclear facilities still remaining (Silos containing uranium ore residues) with its own safety basis documentation. This paper presents the status of the FCP's safety basis documentation program, illustrating that all of the former nuclear facilities and activities have now replaced. Basis of Interim Operations (BIOs) with I-HASPs as their safety basis during the closure process

  2. Development of safety analysis technology for LMR

    Energy Technology Data Exchange (ETDEWEB)

    Hahn, Do Hee; Kwon, Y. M.; Kim, K. D. [and others

    2000-05-01

    The analysis methodologies as well as the analysis computer code system for the transient, HCDA, and containment performance analyses, which are required for KALIMER safety analyses, have been developed. The SSC-K code has been developed based on SSC-L which is an analysis code for loop type LMR, by improving models necessary for the KALIMER system analysis, and additional models have been added to the code. In addition, HCDA analysis model has been developed and the containment performance analysis code has been also improved. The preliminary basis for the safety analysis has been established, and the preliminary safety analyses for the key design features have been performed. In addition, a state-of-art analysis for LMR PSA and overseas safety and licensing requirements have been reviewed. The design database for the systematic management of the design documents as well as design processes has been established as well.

  3. Planning and core analyses for periodic aggregate safety data reviews.

    Science.gov (United States)

    Xia, H Amy; Crowe, Brenda J; Schriver, Robert C; Oster, Manfred; Hall, David B

    2011-04-01

    In 2009, the Safety Planning, Evaluation and Reporting Team gave detailed recommendations for a well-planned and systematic approach for safety data collection and analysis. Important aspects of this approach included regular reviews of aggregate data by a multidisciplinary team focusing on safety. This article provides information to facilitate the planning and implementation of aggregate data reviews. Our recommendations are based on experience of the authors and review of relevant literature. We present information regarding the planning of aggregate data reviews as well as examples of data displays that are useful for many different compounds. A subset of these data displays could form a set of 'core' analyses to be generated for aggregate data reviews.

  4. The development of safety requirements

    International Nuclear Information System (INIS)

    Jorel, M.

    2009-01-01

    This document describes the safety approach followed in France for the design of nuclear reactors. This safety approach is based on safety principles from which stem safety requirements that set limiting values for specific parameters. The improvements in computerized simulation, the use of more adequate new materials, a better knowledge of the concerned physical processes, the changes in the reactor operations (higher discharge burnups for instance) have to be taken into account for the definition of safety criteria and the setting of limiting values. The developments of the safety criteria linked to the risks of cladding failure and loss of primary coolant are presented. (A.C.)

  5. Safety analyses of the electrical systems on VVER NPP

    International Nuclear Information System (INIS)

    Andel, J.

    2004-01-01

    Energoprojekt Praha has been the main entity responsible for the section on 'Electrical Systems' in the safety reports of the Temelin, Dukovany and Mochovce nuclear power plants. The section comprises 2 main chapters, viz. Offsite Power System (issues of electrical energy production in main generators and the link to the offsite transmission grid) and Onsite Power Systems (AC and DC auxiliary system, both normal and safety related). In the chapter on the off-site system, attention is paid to the analysis of transmission capacity of the 400 kV lines, analysis of transient stability, multiple fault analyses, and probabilistic analyses of the grid and NPP power system reliability. In the chapter on the on-site system, attention is paid to the power balances of the electrical sources and switchboards set for various operational and accident modes, checks of loading and function of service and backup sources, short circuit current calculations, analyses of electrical protections, and analyses of the function and sizing of emergency sources (DG sets and UPS systems). (P.A.)

  6. Road safety in developing countries.

    NARCIS (Netherlands)

    Schreuder, D.A.

    1991-01-01

    This paper presents a classification of countries (developing and developed alike), divided into two main categories: an economical and historical entry. When road safety problems are placed into the economical context, it then appears that, among other things: (1) The road safety problem in the

  7. Method of accounting for code safety valve setpoint drift in safety analyses

    International Nuclear Information System (INIS)

    Rousseau, K.R.; Bergeron, P.A.

    1989-01-01

    In performing the safety analyses for transients that result in a challenge to the reactor coolant system (RCS) pressure boundary, the general acceptance criterion is that the peak RCS pressure not exceed the American Society of Mechanical Engineers limit of 110% of the design pressure. Without crediting non-safety-grade pressure mitigating systems, protection from this limit is mainly provided by the primary and secondary code safety valves. In theory, the combination of relief capacity and setpoints for these valves is designed to provide this protection. Generally, banks of valves are set at varying setpoints staggered by 15- to 20-psid increments to minimize the number of valves that would open by an overpressure challenge. In practice, however, when these valves are removed and tested (typically during a refueling outage), setpoints are sometimes found to have drifted by >50 psid. This drift should be accounted for during the performance of the safety analysis. This paper describes analyses performed by Yankee Atomic Electric Company (YAEC) to account for setpoint drift in safety valves from testing. The results of these analyses are used to define safety valve operability or acceptance criteria

  8. Criticality safety analyses in SKODA JS a.s

    International Nuclear Information System (INIS)

    Mikolas, P.; Svarny, J.

    1999-01-01

    This paper describes criticality safety analyses of spent fuel systems for storage and transport of spent fuel performed in SKODA JS s.r.o.. Analyses were performed for different systems both at NPP site including originally designed spent fuel pool with a large pitch between assemblies without any special absorbing material, high density spent fuel pool with an additional absorption by boron steel, depository rack for fresh fuel assemblies with a very large pitch between fuel assemblies, a container for transport of fresh fuel into the reactor pool and a cask for transport and storage of spent fuel and container for final storage depository. required subcriticality has been proven taking into account all possible unfavourable conditions, uncertainties etc. In two cases, burnup credit methodology is expected to be used. (Authors)

  9. Requirements on the provisional safety analyses and technical comparison of safety measures

    International Nuclear Information System (INIS)

    2010-04-01

    The concept of a Geological Underground Repository (SGT) was adopted by the Swiss Federal Council on April 2 nd , 2008. It fixes the goals and the safety technical criteria as well as the procedures for the choice of the site for an underground repository. Those responsible for waste management evaluate possible site regions according to the present status of geological knowledge and based on the safety criteria defined in SGT as well as on technical feasibility. In a first step, they propose geological repository sites for high level (HAA) and for low and intermediate level (SMA) radioactive wastes and justify their choice in a report delivered to the Swiss Federal Office of Energy. The Swiss Federal Council reviews the choices presented and, in the case of positive evaluation, approves them and considers them as an initial orientation. In a second step, based on the possible sites according to step 1, the waste management institution responsible has to reduce the repositories chosen for HAA and SMA by taking into account safety aspects, technical feasibility as well as space planning and socio-economical aspects. In making this choice, safety aspects have the highest priority. The criteria used for the evaluation in the first step have to be defined using provisional quantitative safety analyses. On the basis of the whole appraisal, including space planning and socio-economical aspects, those responsible for waste management propose at least two repository sites for HAA- and SMA-waste. Their selection is then reviewed by the authorities and, in the case of a positive assesment, the selection is taken as an intermediate result. The remaining sites are further studied to examine site choice and the delivery of a request for a design license. If necessary, the requested geological knowledge has to be confirmed by new investigations. Based on the results of the choosing process and a positive evaluation by the safety authorities, the Swiss Federal Council has to

  10. DEVELOPING CITIZEN SAFETY

    Directory of Open Access Journals (Sweden)

    VRABIE Catalin

    2013-12-01

    Full Text Available Is it possible to involve citizens in the process of increasing public safety? Police used, even from its beginnings, the help of citizens, otherwise they would encounter problems in performing its duty - many of its successes were due to the unification of Police forces with the citizens. How citizens get involved? (1 They may be directly asked by the Police officers (a time consuming method because many police officers needs to go on the field to speak with the potential witnesses or (2 by using the mass-media channels (television can address to a large number of potential witnesses in a very short time. We still can see on TV portraits of missing persons, or some other kind of images with which the Police is trying to solve some of its cases (thieves, robbers or burglars surprised by surveillance cameras – why not Internet software application?!

  11. A tool for safety officers when analysing the basic causes of simple accidents

    DEFF Research Database (Denmark)

    Jørgensen, Kirsten

    situation no one could do anything about. In 1999 a new practical tool for use by safety officers was developed and used by training systems. The tool involves three steps: one for observing facts, one for analysing the event, and one for preventive solutions. The tool has been used by various companies...

  12. Sustainable Development of Food Safety

    DEFF Research Database (Denmark)

    Fabech, B.; Georgsson, F.; Gry, Jørn

    The Nordic countries, Denmark, Finland, Iceland, Norway and Sweden published a strategy for sustainable developments for 2001-2004. The strategy for 2005-2008 has 10 goals and some selected indicators. The 10 goals set for 2005-2008 are : - Improved efforts for animal health and welfare in relation...... - Strengthen scientific knowledge of food safety - Strengthen consumer knowledge The goals for sustainable development of food safety are listed from farm to fork". All of the steps and areas are important for food safety and consumer protection. Initiatives are needed in all areas. Many of the goals...... in other areas. It should be emphasized that an indicator will be an excellent tool to assess the efficacy of initiatives started to achieve a goal. Conclusions from the project are: - Sustainable development in food safety is important for humanity - Focus on the crucial goals would optimize the efforts...

  13. Concepts and examples of safety analyses for radioactive waste repositories in continental geological formations

    International Nuclear Information System (INIS)

    1983-01-01

    This document is addressed to authorities and specialists responsible for or involved in planning, performing and/or reviewing safety assessments of underground radioactive waste repositories. It is a companion to a general introductory document on the subject ''Safety Assessment for the Underground Disposal of Radioactive Wastes'', IAEA Safety Series No. 56, 1981, and reference to this earlier document will facilitate the reader's understanding of the present report. Since examples of safety analyses are summarized here, it is hoped that this document will contribute to providing a basis for a common understanding among authorities and specialists concerned with the numerous studies involving a variety of scientific disciplines. While providing technical information, this document is also intended to stimulate further international discussion. The purposes of this report are: a) to identify the factors to be taken into account in radiological safety analyses of deep geological repositories, indicating as far as possible their relative importance during the various phases of system development; b) to show how these factors have been analysed in various safety assessment studies; and c) to comment on the merits of the selected and alternative approaches

  14. Radiation Safety for Sustainable Development

    International Nuclear Information System (INIS)

    2015-10-01

    The objective of radiation safety is Assessments of Natural Radioactivity and its Radiological. The following topics were discussed during the conference: AFROSAFE Championing Radiation Safety in Africa, Radiation Calibration, and Development and Validation of a Laser Induced Breakdown Spectrometry Method for Cancer Detection and Characterization. Young Generation in NUCLEAR Initiative to Promote Nuclear Science and Technology, Radiation Protection Safety Culture and Application of Nuclear Techniques in Industry and the Environment were discuss. Rapid Chemometric X-Ray Fluorescence approaches for spectral Diagnostics of Cancer utilizing Tissue Trace Metals and Speciation profiles. Fundamental role of medical physics in Radiation Therapy

  15. Accelerated safety analyses - structural analyses Phase I - structural sensitivity evaluation of single- and double-shell waste storage tanks

    International Nuclear Information System (INIS)

    Becker, D.L.

    1994-11-01

    Accelerated Safety Analyses - Phase I (ASA-Phase I) have been conducted to assess the appropriateness of existing tank farm operational controls and/or limits as now stipulated in the Operational Safety Requirements (OSRs) and Operating Specification Documents, and to establish a technical basis for the waste tank operating safety envelope. Structural sensitivity analyses were performed to assess the response of the different waste tank configurations to variations in loading conditions, uncertainties in loading parameters, and uncertainties in material characteristics. Extensive documentation of the sensitivity analyses conducted and results obtained are provided in the detailed ASA-Phase I report, Structural Sensitivity Evaluation of Single- and Double-Shell Waste Tanks for Accelerated Safety Analysis - Phase I. This document provides a summary of the accelerated safety analyses sensitivity evaluations and the resulting findings

  16. Preliminary Results of Ancillary Safety Analyses Supporting TREAT LEU Conversion Activities

    Energy Technology Data Exchange (ETDEWEB)

    Brunett, A. J. [Argonne National Lab. (ANL), Argonne, IL (United States); Fei, T. [Argonne National Lab. (ANL), Argonne, IL (United States); Strons, P. S. [Argonne National Lab. (ANL), Argonne, IL (United States); Papadias, D. D. [Argonne National Lab. (ANL), Argonne, IL (United States); Hoffman, E. A. [Argonne National Lab. (ANL), Argonne, IL (United States); Kontogeorgakos, D. C. [Argonne National Lab. (ANL), Argonne, IL (United States); Connaway, H. M. [Argonne National Lab. (ANL), Argonne, IL (United States); Wright, A. E. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2015-10-01

    The Transient Reactor Test Facility (TREAT), located at Idaho National Laboratory (INL), is a test facility designed to evaluate the performance of reactor fuels and materials under transient accident conditions. The facility, an air-cooled, graphite-moderated reactor designed to utilize fuel containing high-enriched uranium (HEU), has been in non-operational standby status since 1994. Currently, in support of the missions of the Department of Energy (DOE) National Nuclear Security Administration (NNSA) Material Management and Minimization (M3) Reactor Conversion Program, a new core design is being developed for TREAT that will utilize low-enriched uranium (LEU). The primary objective of this conversion effort is to design an LEU core that is capable of meeting the performance characteristics of the existing HEU core. Minimal, if any, changes are anticipated for the supporting systems (e.g. reactor trip system, filtration/cooling system, etc.); therefore, the LEU core must also be able to function with the existing supporting systems, and must also satisfy acceptable safety limits. In support of the LEU conversion effort, a range of ancillary safety analyses are required to evaluate the LEU core operation relative to that of the existing facility. These analyses cover neutronics, shielding, and thermal hydraulic topics that have been identified as having the potential to have reduced safety margins due to conversion to LEU fuel, or are required to support the required safety analyses documentation. The majority of these ancillary tasks have been identified in [1] and [2]. The purpose of this report is to document the ancillary safety analyses that have been performed at Argonne National Laboratory during the early stages of the LEU design effort, and to describe ongoing and anticipated analyses. For all analyses presented in this report, methodologies are utilized that are consistent with, or improved from, those used in analyses for the HEU Final Safety Analysis

  17. Safety and environmental analyses for space nuclear programs

    Science.gov (United States)

    Mcconnell, D. J.

    1990-01-01

    The tools and procedures for analyzing environmental quality and safety are reviewed. The process of preparing an environmental impact statement is outlined and the data sources for a safety analysis are discussed. The environmental safety analysis process is demonstrated, using examples from the Galileo, Ulysses, and Venus-earth-earth-gravity-assist programs.

  18. Development of safety analysis technology for integral reactor

    International Nuclear Information System (INIS)

    Sim, Suk K.; Song, J. H.; Chung, Y. J. and others

    1999-03-01

    Inherent safety features and safety system characteristics of the SMART integral reactor are investigated in this study. Performance and safety of the SMART conceptual design have been evaluated and confirmed through the performance and safety analyses using safety analysis system codes as well as a preliminary performance and safety analysis methodology. SMART design base events and their acceptance criteria are identified to develop a preliminary PIRT for the SMART integral reactor. Using the preliminary PIRT, a set of experimental program for the thermal hydraulic separate effect tests and the integral effect tests was developed for the thermal hydraulic model development and the system code validation. Safety characteristics as well as the safety issues of the integral reactor has been identified during the study, which will be used to resolve the safety issues and guide the regulatory criteria for the integral reactor. The results of the performance and safety analyses performed during the study were used to feedback for the SMART conceptual design. The performance and safety analysis code systems as well as the preliminary safety analysis methodology developed in this study will be validated as the SMART design evolves. The performance and safety analysis technology developed during the study will be utilized for the SMART basic design development. (author)

  19. Review of methodologies used in safety analyses of nuclear waste disposal

    International Nuclear Information System (INIS)

    Robinson, P.C.

    1984-06-01

    The methodologies used in safety analyses of nuclear waste disposal are reviewed. Particular reference is made to the KBS-3 proposals and to the WP-Cave concept. These methodologies are aimed mainly at reducing uncertainties in the predictions of health consequences from nuclear waste repositories, and quantifying the uncertainties that remain. The steps of a safety analysis are described in turn and the currently used techniques are discussed. Recommendations are made as to the use and development of the techniques and the need for new techniques. (author)

  20. Swiss-Slovak cooperation program: a training strategy for safety analyses

    International Nuclear Information System (INIS)

    Husarcek, J.

    2000-01-01

    During the 1996-1999 period, a new training strategy for safety analyses was implemented at the Slovak Nuclear Regulatory Authority (UJD) within the Swiss-Slovak cooperation programme in nuclear safety (SWISSLOVAK). The SWISSLOVAK project involved the recruitment, training, and integration of the newly established team into UJD's organizational structure. The training strategy consisted primarily of the following two elements: a) Probabilistic Safety Analysis (PSA) applications (regulatory review and technical evaluation of Level-1/Level-2 PSAs; PSA-based operational events analysis, PSA applications to assessment of Technical Specifications; and PSA-based hardware and/or procedure modifications) and b) Deterministic accident analyses (analysis of accidents and regulatory review of licensee Safety Analysis Reports; analysis of severe accidents/radiological releases and the potential impact of the containment and engineered safety systems, including the development of technical bases for emergency response planning; and application of deterministic methods for evaluation of accident management strategies/procedure modifications). The paper discusses the specific aspects of the training strategy performed at UJD in both the probabilistic and deterministic areas. The integration of team into UJD's organizational structure is described and examples of contributions of the team to UJD's statutory responsibilities are provided. (author)

  1. Statistical modelling of traffic safety development

    DEFF Research Database (Denmark)

    Christens, Peter

    2004-01-01

    Road safety is a major concern for society and individuals. Although road safety has improved in recent years, the number of road fatalities is still unacceptably high. In 2000, road accidents killed over 40,000 people in the European Union and injured more than 1.7 million. In 2001 in Denmark...... there were 6861 injury trafficc accidents reported by the police, resulting in 4519 minor injuries, 3946 serious injuries, and 431 fatalities. The general purpose of the research was to improve the insight into aggregated road safety methodology in Denmark. The aim was to analyse advanced statistical methods......, that were designed to study developments over time, including effects of interventions. This aim has been achieved by investigating variations in aggregated Danish traffic accident series and by applying state of the art methodologies to specific case studies. The thesis comprises an introduction...

  2. C4P cross-section libraries for safety analyses with SIMMER and related studies

    International Nuclear Information System (INIS)

    Rineiski, A.; Sinitsa, V.; Gabrielli, F.; Maschek, W.

    2011-01-01

    A code and data system, C 4 P, is under development at KIT. It includes fine-group master libraries and tools for generating problem-oriented cross-section libraries, primarily for safety studies with the SIMMER code and related analyses. In the paper, the 560-group master library and problem oriented 40-group and 72-group cross-section libraries, for thermal and fast systems, respectively, are described and their performances are investigated. (author)

  3. Process hazards analysis (PrHA) program, bridging accident analyses and operational safety

    International Nuclear Information System (INIS)

    Richardson, J.A.; McKernan, S.A.; Vigil, M.J.

    2003-01-01

    Recently the Final Safety Analysis Report (FSAR) for the Plutonium Facility at Los Alamos National Laboratory, Technical Area 55 (TA-55) was revised and submitted to the US. Department of Energy (DOE). As a part of this effort, over seventy Process Hazards Analyses (PrHAs) were written and/or revised over the six years prior to the FSAR revision. TA-55 is a research, development, and production nuclear facility that primarily supports US. defense and space programs. Nuclear fuels and material research; material recovery, refining and analyses; and the casting, machining and fabrication of plutonium components are some of the activities conducted at TA-35. These operations involve a wide variety of industrial, chemical and nuclear hazards. Operational personnel along with safety analysts work as a team to prepare the PrHA. PrHAs describe the process; identi fy the hazards; and analyze hazards including determining hazard scenarios, their likelihood, and consequences. In addition, the interaction of the process to facility systems, structures and operational specific protective features are part of the PrHA. This information is rolled-up to determine bounding accidents and mitigating systems and structures. Further detailed accident analysis is performed for the bounding accidents and included in the FSAR. The FSAR is part of the Documented Safety Analysis (DSA) that defines the safety envelope for all facility operations in order to protect the worker, the public, and the environment. The DSA is in compliance with the US. Code of Federal Regulations, 10 CFR 830, Nuclear Safety Management and is approved by DOE. The DSA sets forth the bounding conditions necessary for the safe operation for the facility and is essentially a 'license to operate.' Safely of day-to-day operations is based on Hazard Control Plans (HCPs). Hazards are initially identified in the PrI-IA for the specific operation and act as input to the HCP. Specific protective features important to worker

  4. MOX LTA Fuel Cycle Analyses: Nuclear and Radiation Safety

    Energy Technology Data Exchange (ETDEWEB)

    Pavlovitchev, A.M.

    2001-09-28

    Tasks of nuclear safety assurance for storage and transport of fresh mixed uranium-plutonium fuel of the VVER-1000 reactor are considered in the view of 3 MOX LTAs introduction into the core. The precise code MCU that realizes the Monte Carlo method is used for calculations.

  5. Methodology for Determination of the Upper Safety Limit for Criticality Calculations for Criticality Safety Analyses

    Energy Technology Data Exchange (ETDEWEB)

    Smith, R.H.; Keener, H.J.; DeClue, J.F.; Krass, A.W.

    2001-04-01

    This report considers the methods for determination of an upper safety limit, and incorporating uncertainty and margin into the safety limit, provides comparisons, and recommends a preferred method for determining the Upper Safety Limit (USL). A USL is developed for CSAS25 from SCALE4.4a. The USL is applicable for the CSAS25 control module from the SCALE 4.4a computer code system for use in evaluating nuclear criticality safety of enriched uranium systems. The benchmark calculation results used for this report are documented in Y/DD-896. The statistical evaluation is documented in CCG-380. The 27-group ENDF/B-IV, 44-group ENDF/B-V, and 238-group ENDF/B-V cross-section libraries were used. Numerical methods for applying margins are described, but the determination of appropriate correlating parameters and values for additional margin, applicable to a particular analysis, must be determined as part of a process analysis. As such, this document does not specify final upper subcritical limits as has been done in the past. No correlation between calculation results and neutron energy causing fission was found for the critical experiment results. Analysts using these results are responsible for exercising sound engineering judgment using strong technical arguments to develop ''a margin in k{sub eff} or other correlating parameter that is sufficiently large to ensure that conditions (calculated by this method to be subcritical by this margin) will actually be subcritical.'' Documentation of area of applicability and determination and justification of the appropriate margin in the analyst's evaluation, in conjunction with this report, will constitute the complete Validation Report in accordance with ANSI/ANS-8.1-1998, Section 4.3.6(4).

  6. Nuclear criticality safety experiments, calculations, and analyses: 1958 to 1982. Volume 1. Lookup tables

    International Nuclear Information System (INIS)

    Koponen, B.L.; Hampel, V.E.

    1982-01-01

    This compilation contains 688 complete summaries of papers on nuclear criticality safety as presented at meetings of the American Nuclear Society (ANS). The selected papers contain criticality parameters for fissile materials derived from experiments and calculations, as well as criticality safety analyses for fissile material processing, transport, and storage. The compilation was developed as a component of the Nuclear Criticality Information System (NCIS) now under development at the Lawrence Livermore National Laboratory. The compilation is presented in two volumes: Volume 1 contains a directory to the ANS Transaction volume and page number where each summary was originally published, the author concordance, and the subject concordance derived from the keyphrases in titles. Volume 2 contains - in chronological order - the full-text summaries, reproduced here by permission of the American Nuclear Society from their Transactions, volumes 1-41

  7. Cardiac safety of citalopram: prospective trials and retrospective analyses

    DEFF Research Database (Denmark)

    Rasmussen, Søren Poul Lind; Overø, K F; Tanghøj, P

    1999-01-01

    The selective serotonin reuptake inhibitors (SSRIs) are believed to have a more benign cardiovascular safety profile than do the tricyclic antidepressants. The effects of the SSRI citalopram on cardiac conduction and repolarization have been extensively evaluated, both in prospective studies...... variability of the QTc interval, as well as possible changes during treatment with placebo or citalopram, and its correlation to plasma drug levels. To document any dose-related changes, ECGs were performed at baseline and at the end of study in three randomized, double-blind, placebo- or active...... no effect on cardiac conduction and repolarization during short- or long-term treatment....

  8. Analysing Scenarios of Cell Population System Development

    Directory of Open Access Journals (Sweden)

    M. S. Vinogradova

    2014-01-01

    Full Text Available The article considers an isolated population system consisting of two types of human stem cells, namely normal cells and cells with chromosomal abnormalities (abnormal ones. The system develops in the laboratory (in vitro. The article analyses possible scenarios of the population system development, which are implemented for different values of its parameters. An investigated model of the cell population system takes into account the limited resources. It is represented as a system of two nonlinear differential equations with continuous right-hand part. The model is considered with non-negative values of the variables; the domain is divided into four sets. The model feature is that in each set the right part of the system of differential equations has a different form.The article analyses a quality of the rest points of the system in each of four sets. The analytical conditions for determination of the number of rest points and the quality of rest points, with, at least, one zero coordinate, are obtained.It is shown that the population system under study cannot have more than two points of rest, both coordinates of which are positive (non-zero. It is difficult to determine quality of such rest points depending on the model parameters due to the complexity of the expressions, which define the systems of the first approximation, recorded in a neighborhood of these points of rest. Numerical research results of the stability of these points of rest are obtained, and phase portraits with the specified specific values of the system parameters are demonstrated. The main scenarios for the cell population development are adduced. Analysis of mathematical model shows that a cell population system may remain the system consisting of populations of normal and abnormal cells; it can degenerate into a population of abnormal cells or perish. The scenario, in which there is only the population of normal cells, is not implemented. The numerical simulation

  9. Development of safety analysis technology for LMR

    International Nuclear Information System (INIS)

    Hahn, Do Hee; Kwon, Y. M.; Suk, S. D.

    2002-05-01

    In the present study, the KALIMER safety analysis has been made for the transients considered in the design concept, hypothetical core disruptive accident (HCDA), and containment performance with the establishment of the design basis. Such analyses have not been possible without the computer code improvement, and the experience attained during this research period must have greatly contributed to the achievement of the self reliance in the domestic technology establishment on the safety analysis areas of the conceptual design. The safety analysis codes have been improved to extend their applicable ranges for detailed conceptual design, and a basic computer code system has been established for HCDA analysis. A code-to-code comparison analysis has been performed as a part of code verification attempt, and the leading edge technology of JNC also has been brought for the technology upgrade. In addition, the research and development on the area of the database establishment has been made for the efficient and systematic project implementation of the conceptual design, through performances on the development of a project scheduling management, integration of the individually developed technology, establishment of the product database, and so on, taking into account coupling of the activities conducted in each specific area

  10. Safety culture and learning from incidents: the role of incident reporting and causal analyses

    International Nuclear Information System (INIS)

    Wilpert, B.

    1994-01-01

    Nuclear industry more than any other industrial branch has developed and used predictive risk analysis as a method of feedforward control of safety and reliability. Systematic evaluation of operating experience, statistical documentation of component failures, systematic documentation and analysis of incidents are important complementary elements of feedback control: we are dealing here with adjustment and learning from experience, in particular from past incidents. Using preliminary findings from ongoing research at the Research Center Systems Safety at the Berlin University of Technology the contribution discusses preconditions for an effective use of lessons to be learnt from closely matched incident reporting and in depth analyses of causal chains leading to incidents. Such conditions are especially standardized documentation, reporting and analyzing methods of incidents; structured information flows and feedback loops; abstaining from culpability search; mutual trust of employees and management; willingness of all concerned to continually evaluate and optimize the established learning system. Thus, incident related reporting and causal analyses contribute to safety culture, which is seen to emerge from tightly coupled organizational measures and respective change in attitudes and behaviour. (author) 2 figs., 7 refs

  11. Development of safety analysis technology for LMR

    International Nuclear Information System (INIS)

    Lee, Y. B.; Kwon, Y. M.; Suk, S. D.

    2005-03-01

    The MATRA-LMR-FB has been developed internally for the damage prevention as well as the safety assessment during a channel blockage accident and, as a the result, the quality of the code becomes comparable to that developed in the leading countries. For a code-to-code comparison, KAERI could have access to the SASSYS-1 through a bilateral collaboration between KAERI and ANL. The study could bring into the reliability improvements both on the reactivity models in the SSC-K and on the SSC-K prediction capability. It finally leads to the completion of the SSC-K version 1.3 resulting from the qualitative and quantitative code-to-code comparison. The preliminary analysis for a metal fueled LMR could also become possible with the MELT-III and the VENUS-II, which had originally been developed for the HCDA analysis with an oxidized fuel, by developing the relevant models For the development of the safety evaluation technology, the safety limits have been set up, and the analyses of the internal and external channel blockages in an assembly have also been performed. Besides, the more reliable analysis results on the key design concepts could be obtained by way of the methodology improvement resulting from the qualitative and quantitative comparison study. For an efficient and systematic control of the main project, the integration of the developed technologies and the establishment of their data base have been pursued. It has gone through the development of the process control with taking account of interfaces among the sub-projects, the overall coordination of the developed technologies, the data base for the design products, and so on

  12. Further development and data basis for safety and accident analyses of nuclear front end and back end facilities and actualization and revision of calculation methods for nuclear safety analyses. Final report; Weiterentwicklung von Methoden und Datengrundlagen zu Sicherheits- und Stoerfallanalysen fuer Anlagen der nuklearen Ver- und Entsorgung sowie Aktualisierung und Ueberpruefung von Rechenmethoden zu nuklearen Sicherheitsanalysen. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Kilger, Robert; Peters, Elisabeth; Sommer, Fabian; Moser, Eberhard-Franz; Kessen, Sven; Stuke, Maik

    2016-07-15

    This report briefly describes the activities carried out under the project 3613R03350 on the GRS ''Handbook on Accident Analysis for Nuclear Front and Back End Facilities'', and in detail the continuing work on the revision and updating of the GRS ''Handbook on Criticality'', which here focused on fissile systems with plutonium and {sup 233}U. The in previous projects started and ongoing literature study on innovative fuel concepts is continued. Also described are the review and qualification of computational methods by research and active benchmark participation, and the results of tracking the state of science and technology in the field of computational methods for criticality safety analysis. Special in-depth analyzes of selected criticality-relevant occurrences in the past are also documented.

  13. Safety analyses of the ARIES tokamak reactor designs

    International Nuclear Information System (INIS)

    Herring, J.S.; McCarthy, K.A.; Dolan, T.J.

    1994-01-01

    The ARIES design has sought to maximize environmental and safety advantages of fusion through careful selection of materials and design. The ARIES-I tokamak reactor design consists of an SiC composite structure for the first wall and blanket, cooled by 10MPa helium. The breeder is Li 2 ZrO 3 . The divertor consists of SiC composite tubes coated with 2mm tungsten. Loss-of-cooling accident (LOCA) calculations indicate maximum temperatures will not cause damage if the plasma is promptly extinguished. The ARIES-II design includes liquid lithium and vanadium, both of which have low activation, multiple barriers between the lithium and air and an inert cover gas to prevent lithium-air reactions. The ARIES-II reactor is passively safe with a total 1km early dose of about 88rem (0.88Sv). ARIES-III was an extensive examination of the viability of a D- 3 He fueled tokamak power reactor. Because neutrons are produced only through side reactions (D+D→ 3 He+n, and D+D→T+p followed by D+T→ 4 He+n), the reactor has a reduced activation of the first wall and shield, low afterheat and class A or C low level waste disposal. Since no tritium is required for operation, no lithium-containing breeding blanket is necessary. We modeled a LOCA in which the organic coolant was burning in order to estimate the amount of radionuclides released from the first wall. Because the maximum temperature is low, below 600 C, release fractions are small. We analyzed the disposition of the 20g per day of tritium that is produced by D-D reactions and removed by vacuum pumps. The ARIES-IV coolant is helium and the breeder is lithium oxide. The structure is silicon carbide. Since the neutron multiplier, beryllium metal, is combustible, releasing about 60MJkg -1 , beryllium is the chief source of chemical energy. Less than 10% of the 24 Na inventory is likely to diffuse out of the SiC during a fire in which the beryllium is consumed. Therefore, the offsite dose would be less than 200rem. ((orig.))

  14. Recognising safety critical events: can automatic video processing improve naturalistic data analyses?

    Science.gov (United States)

    Dozza, Marco; González, Nieves Pañeda

    2013-11-01

    applications for NDS video processing. As new NDS such as SHRP2 are now providing the equivalent of five years of one vehicle data each day, the development of new methods, such as the one proposed in this paper, seems necessary to guarantee that these data can actually be analysed. Copyright © 2013 Elsevier Ltd. All rights reserved.

  15. Uncertainties and credibility building of safety analyses. Natural analogues

    International Nuclear Information System (INIS)

    Laciok, A.

    2001-07-01

    The substance of natural analogues and their studies is defined as a complementary method to laboratory and in-situ experiments and modelling. The role of natural analogues in the processes of development of repositories is defined, mainly in performance assessment of repository system and communication with public. The criteria for identification of natural analogues which should be evaluated in the phase of initiation of new studies are specified. Review part of this report is divided to study of natural analogues and study of anthropogenic and industrial analogues. The main natural analogue studies performed in various countries, in different geological setting, with various aims are characterized. New results acquired in recently finished studies are included: Palmottu (2nd phase of project financed by European Commission), Oklo (results of research financed also by European Commission), Maqarin (3rd phase) and other information obtained from last meetings and workshops of NAWG. In view of the fact that programmes of development of deep repositories in Czech and Slovak Republics are interconnected, the natural analogues studies carried out in the Czech republic are incorporated in separate chapter - study of uranium accumulation in Tertiary clays at Ruprechtov site and study of degradation of natural glasses. In final part the areas of natural analogue studies as an integral part of development of deep geological repository are proposed along with characterization of broader context and aspects of realization of these studies (international cooperation, preparation and evaluation of procedures, communication with public). (author)

  16. Developing patient safety in dentistry.

    Science.gov (United States)

    Pemberton, M N

    2014-10-01

    Patient safety has always been important and is a source of public concern. Recent high profile scandals and subsequent reports, such as the Francis report into the failings at Mid Staffordshire, have raised those concerns even higher. Mortality and significant morbidity associated with the practice of medicine has led to many strategies to help improve patient safety, however, with its lack of associated mortality and lower associated morbidity, dentistry has been slower at systematically considering how patient safety can be improved. Recently, several organisations, researchers and clinicians have discussed the need for a patient safety culture in dentistry. Strategies are available to help improve patient safety in healthcare and deserve further consideration in dentistry.

  17. Model-Driven Development of Safety Architectures

    Science.gov (United States)

    Denney, Ewen; Pai, Ganesh; Whiteside, Iain

    2017-01-01

    We describe the use of model-driven development for safety assurance of a pioneering NASA flight operation involving a fleet of small unmanned aircraft systems (sUAS) flying beyond visual line of sight. The central idea is to develop a safety architecture that provides the basis for risk assessment and visualization within a safety case, the formal justification of acceptable safety required by the aviation regulatory authority. A safety architecture is composed from a collection of bow tie diagrams (BTDs), a practical approach to manage safety risk by linking the identified hazards to the appropriate mitigation measures. The safety justification for a given unmanned aircraft system (UAS) operation can have many related BTDs. In practice, however, each BTD is independently developed, which poses challenges with respect to incremental development, maintaining consistency across different safety artifacts when changes occur, and in extracting and presenting stakeholder specific information relevant for decision making. We show how a safety architecture reconciles the various BTDs of a system, and, collectively, provide an overarching picture of system safety, by considering them as views of a unified model. We also show how it enables model-driven development of BTDs, replete with validations, transformations, and a range of views. Our approach, which we have implemented in our toolset, AdvoCATE, is illustrated with a running example drawn from a real UAS safety case. The models and some of the innovations described here were instrumental in successfully obtaining regulatory flight approval.

  18. Development of core design and analyses technology for integral reactor

    International Nuclear Information System (INIS)

    Zee, Sung Quun; Lee, C. C.; Kim, K. Y.

    2002-03-01

    In general, small and medium-sized integral reactors adopt new technology such as passive and inherent safety concepts to minimize the necessity of power source and operator actions, and to provide the automatic measures to cope with any accidents. Specifically, such reactors are often designed with a lower core power density and with soluble boron free concept for system simplification. Those reactors require ultra long cycle operation for higher economical efficiency. This cycle length requirement is one of the important factors in the design of burnable absorbers as well as assurance of shutdown margin. Hence, both computer code system and design methodology based on the today's design technology for the current commercial reactor cores require intensive improvement for the small and medium-sized soluble boron free reactors. New database is also required for the development of this type of reactor core. Under these technical requirements, conceptual design of small integral reactor SMART has been performed since July 1997, and recently completed under the long term nuclear R and D program. Thus, the final objectives of this work is design and development of an integral reactor core and development of necessary indigenous design technology. To reach the goal of the 2nd stage R and D program for basic design of SMART, design bases and requirements adequate for ultra long cycle and soluble boron free concept are established. These bases and requirements are satisfied by the core loading pattern. Based on the core loading pattern, nuclear, and thermal and hydraulic characteristics are analyzed. Also included are fuel performance analysis and development of a core protection and monitoring system that is adequate for the soluble boron free core of an integral reactor. Core shielding design analysis is accomplished, too. Moreover, full scope interface data are produced for reactor safety and performance analyses and other design activities. Nuclear, thermal and

  19. Advanced Messaging Concept Development Basic Safety Message

    Data.gov (United States)

    Department of Transportation — Contains all Basic Safety Messages (BSMs) collected during the Advanced Messaging Concept Development (AMCD) field testing program. For this project, all of the Part...

  20. Multi-person and multi-attribute design evaluations using evidential reasoning based on subjective safety and cost analyses

    International Nuclear Information System (INIS)

    Wang, J.; Yang, J.B.; Sen, P.

    1996-01-01

    This paper presents an approach for ranking proposed design options based on subjective safety and cost analyses. Hierarchical system safety analysis is carried out using fuzzy sets and evidential reasoning. This involves safety modelling by fuzzy sets at the bottom level of a hierarchy and safety synthesis by evidential reasoning at higher levels. Fuzzy sets are also used to model the cost incurred for each design option. An evidential reasoning approach is then employed to synthesise the estimates of safety and cost, which are made by multiple designers. The developed approach is capable of dealing with problems of multiple designers, multiple attributes and multiple design options to select the best design. Finally, a practical engineering example is presented to demonstrate the proposed multi-person and multi-attribute design selection approach

  1. Integration of safety culture in transient analyses for nuclear power plants

    International Nuclear Information System (INIS)

    Stosic, Zoran V.; Stoll, Uwe

    2009-01-01

    In the nuclear field Safety Culture is the arrangement of attitudes and characteristics in individuals and organisations which determines first and foremost that nuclear power plant safety issues receive adequate attention due to their outstanding significance. It differs from general Corporate Culture via its concept of core hazards and the potentially large effects associated with the release of radioactivity. One can talk about positive and negative Safety Cultures. A positive Safety Culture assumes that the whole is more than the sum of the parts. The different parts interact to increase the overall effectiveness. In a negative Safety Culture the opposite is the case, with the action of some individuals restricted by the cynicism of others. Some examples of issues that contribute to a negative safety culture are: non-adherence to the established instructions and procedures, unclear definition of responsibilities, disinterest and inattentiveness, overestimation of own capabilities and arrogance, unclear rules, and mistrust between involved organisations. In addition to differentiation and importance of Safety Culture, necessary commitment levels, safety management framework, the paper discusses integration of Safety Culture in transient analyses of nuclear power plants. In this course the commitment to Safety Culture is defined as: a good Safety Culture depends on the continuous commitment and fulfilment of all involved organizations, persons and processes without any exception. (author)

  2. Safety guides development process in Spain

    International Nuclear Information System (INIS)

    Butragueno, J.L.; Perello, M.

    1979-01-01

    Safety guides have become a major factor in the licensing process of nuclear power plants and related nuclear facilities of the fuel cycle. As far as the experience corroborates better and better engineering methodologies and procedures, the results of these are settled down in form of standards, guides, and similar issues. This paper presents the actual Spanish experience in nuclear standards and safety guides development. The process to develop a standard or safety guide is shown. Up to date list of issued and on development nuclear safety guides is included and comments on the future role of nuclear standards in the licensing process are made. (author)

  3. Development of safety analysis technology for integral reactor; evaluation on safety concerns of integral reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hee Chul; Kim, Woong Sik; Lee, J. H. [Korea Institute of Nuclear Safety, Taejeon (Korea)

    2002-03-01

    The Nuclear Desalination Plant (NDP) is being developed to produce electricity and fresh water, and is expected to locate near population zone. In the aspect of safety, it is required to protect the public and environment from the possible releases of fission products and to prevent the fresh water from the contamination of radioactivity. Thus, in this study, the safety characteristics of the integral reactor adopting passive and inherent safety features significantly different from existing nuclear power plants were investigated. Also, safety requirements applicable to the NDP were analyzed based on the regulatory requirements for current light water reactor and advanced reactor designs, and user requirements for small-medium size reactors. Based on these analyses, some safety concerns to be considered in the design stage have been identified and discussed. They include the use of proven technology for new safety features, systematic event classification and selection, strengthening containment function, and the safety impacts on desalination-related systems. The study presents the general safety requirements applicable to licensing of an integral reactor and suggests additional regulatory requirements, which need to be developed, based on the direction to resolution of the safety concerns. The efforts to identify and technically resolve the safety concerns in the design stage will provide the early confidence of SMART safety and the technical basis to evaluate the safety to designers and reviewers in the future. Suggestion on the development of additional regulatory requirements will contribute for the regulator to taking actions for licensing of an integral reactor. 66 refs., 5 figs., 24 tabs. (Author)

  4. Safety culture assessment developed by JANTI

    International Nuclear Information System (INIS)

    Hamada, Jun

    2009-01-01

    Japan's JCO accident in September 1999 provided a real-life example of what can happen when insufficient attention is paid to safety culture. This accident brought to light the importance of safety culture and reinforced the movement to foster a safety culture. Despite this, accidents and inappropriate conduct have continued to occur. Therefore, there is a strong demand to instill a safety culture throughout the nuclear power industry. In this context, Japan's nuclear power regulator, the Nuclear and Industrial Safety Agency (NISA), decided to include in its safety inspections assessments of the safety culture found in power utilities' routine safety operations to get signs of deterioration in the organizational climate. In 2007, NISA constructed guidelines for their inspectors to carry out these assessments. At the same time, utilities have embarked on their own independent safety culture initiatives, such as revising their technical specifications and building effective PDCA cycle to promote safety culture. In concert with these developments, JANTI has also instituted safety culture assessments. (author)

  5. Sign up to Safety: developing a safety improvement plan.

    Science.gov (United States)

    Dight, Carol; Peters, Hayley

    2015-04-01

    The Sign up to Safety (SutS) programme was launched in June 2014 by health secretary Jeremy Hunt. It focuses on listening to patients, carers and staff, learning from what they say when things go wrong, and then taking action to improve patient safety. The programme aims to make the NHS the safest healthcare system in the world by creating a culture devoted to continuous learning and improvement (NHS England 2014). Musgrove Park Hospital, part of Taunton and Somerset NHS Foundation Trust, was one of 12 NHS organisations that signed up to the SutS programme, making public its commitment to the national pledges to be 'open and transparent' and to develop a safety improvement plan. This paper describes the development of the strategy.

  6. Development of Safety Analysis Technology for Integral Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sim, S. K. [Korea Atomic Energy Research Institute, Taejeon (Korea); Seul, K. W.; Kim, W. S.; Kim, W. K.; Yun, Y. G.; Ahn, H. J.; Lee, J. S.; Sin, A. D. [Korea Institute of Nuclear Safety, Taejeon (Korea)

    2000-03-01

    The Nuclear Desalination Plant(NDP) is being developed to produce electricity and fresh water, and is expected to locate near population zone. In the aspect of safety, it is required to protect the public and environment from the possible releases of fission products and to prevent the fresh water from the contamination of radioactivity. Thus, in a present study, the safety characteristics of the integral reactor adopting passive and inherent safety features significantly different from existing nuclear power plants were investigated based on the design of foreign and domestic integral reactors. Also, safety requirements applicable to the NDP were analyzed based on the regulatory requirements for current and advanced reactor designs, and use requirements for small-medium size reactors. Based on these analyses, some safety concerns to be considered in the design stage have been identified. They includes the use of proven technology for new safety systems, the systematic classification and selection of design basis accidents, and the safety assurance of desalination-related systems. These efforts to identify and resolve the safety concerns in the design stage will provide the early confidence of SMART safety to designers, and the technical basis to evaluate the safety to reviewers in the future. 8 refs., 20 figs., 4 tabs. (Author)

  7. EULAR points to consider when establishing, analysing and reporting safety data of biologics registers in rheumatology

    DEFF Research Database (Denmark)

    Dixon, William G; Carmona, Loreto; Finckh, Axel

    2010-01-01

    upon safety data generated from observational drug registers makes it important to convert the lessons learned from such registers into recommendations for rheumatologists embarking upon the establishment of future registers, or analysing and reporting from new and existing registers.......The introduction of biological therapies for the treatment of rheumatic diseases has drawn attention to the limitations of traditional means of assessing drug safety. Consequently, a series of European academic biologics registers dedicated to this task have been established. Increasing reliance...

  8. Developing safety culture in nuclear power engineering

    International Nuclear Information System (INIS)

    Tevlin, S.A.

    2000-01-01

    The new issue (no. 11) of the IAEA publications series Safety Reports, devoted to the safety culture in nuclear engineering Safety culture development in the nuclear activities. Practical recommendations to achieve success, is analyzed. A number of recommendations of international experts is presented and basic general indicators of satisfactory and insufficient safety culture in the nuclear engineering are indicated. It is shown that the safety culture has two foundations: human behavior and high quality of the control system. The necessity of creating the confidence by the management at all levels of the enterprise, development of individual initiative and responsibility of the workers, which make it possible to realize the structural hierarchic system, including technical, human and organizational constituents, is noted. Three stages are traced in the process of introducing the safety culture. At the first stage the require,emts of scientific-technical documentation and provisions of the governmental, regional and control organs are fulfilled. At the second stage the management of the organization accepts the safety as an important direction in its activities. At the third stage the organization accomplishes its work, proceeding from the position of constant safety improvement. The general model of the safety culture development is considered [ru

  9. Towards an Industrial Application of Statistical Uncertainty Analysis Methods to Multi-physical Modelling and Safety Analyses

    International Nuclear Information System (INIS)

    Zhang, Jinzhao; Segurado, Jacobo; Schneidesch, Christophe

    2013-01-01

    Since 1980's, Tractebel Engineering (TE) has being developed and applied a multi-physical modelling and safety analyses capability, based on a code package consisting of the best estimate 3D neutronic (PANTHER), system thermal hydraulic (RELAP5), core sub-channel thermal hydraulic (COBRA-3C), and fuel thermal mechanic (FRAPCON/FRAPTRAN) codes. A series of methodologies have been developed to perform and to license the reactor safety analysis and core reload design, based on the deterministic bounding approach. Following the recent trends in research and development as well as in industrial applications, TE has been working since 2010 towards the application of the statistical sensitivity and uncertainty analysis methods to the multi-physical modelling and licensing safety analyses. In this paper, the TE multi-physical modelling and safety analyses capability is first described, followed by the proposed TE best estimate plus statistical uncertainty analysis method (BESUAM). The chosen statistical sensitivity and uncertainty analysis methods (non-parametric order statistic method or bootstrap) and tool (DAKOTA) are then presented, followed by some preliminary results of their applications to FRAPCON/FRAPTRAN simulation of OECD RIA fuel rod codes benchmark and RELAP5/MOD3.3 simulation of THTF tests. (authors)

  10. Developing IAM for Life Cycle Safety Assessment

    NARCIS (Netherlands)

    Toxopeus, Marten E.; Lutters, Diederick; Nee, Andrew Y.C.; Song, Bin; Ong, Soh-Khim

    2013-01-01

    This publication discusses aspects of the development of an impact assessment method (IAM) for safety. Compared to the many existing IAM’s for environmentally oriented LCA, this method should translate the impact of a product life cycle on the subject of safety. Moreover, the method should be

  11. Safety management in research and development organisation

    International Nuclear Information System (INIS)

    Nivedha, T.

    2016-01-01

    Health and safety is one of the most important aspects of an organizations smooth and effective functioning. It depends on the safety management, health management, motivation, leadership and training, welfare facilities, accident statistics, policy, organization and administration, hazard control and risk analysis, monitoring, statistics and reporting. Workplace accidents are increasingly common, main causes are untidiness, noise, too hot or cold environments, old or poorly maintained machines, and lack of training or carelessness of employees. One of the biggest issues facing employers today is the safety of their employees. This study aims at analyzing the occupational health and safety of Research organization in Indira Gandhi Centre for Atomic Research by gathering information on health management, safety management, motivation, leadership and training, welfare facilities, accident statistics, organization and administration, hazard control and risk analysis, monitoring, statistics and reporting. Data were collected by using questionnaires which were developed on health and safety management system. (author)

  12. Collaborating with nurse leaders to develop patient safety practices.

    Science.gov (United States)

    Kanerva, Anne; Kivinen, Tuula; Lammintakanen, Johanna

    2017-07-03

    Purpose The organisational level and leadership development are crucial elements in advancing patient safety, because patient safety weaknesses are often caused by system failures. However, little is known about how frontline leader and director teams can be supported to develop patient safety practices. The purpose of this study is to describe the patient safety development process carried out by nursing leaders and directors. The research questions were: how the chosen development areas progressed in six months' time and how nursing leaders view the participatory development process. Design/methodology/approach Participatory action research was used to engage frontline nursing leaders and directors into developing patient safety practices. Semi-structured group interviews ( N = 10) were used in data collection at the end of a six-month action cycle, and data were analysed using content analysis. Findings The participatory development process enhanced collaboration and gave leaders insights into patient safety as a part of the hospital system and their role in advancing it. The chosen development areas advanced to different extents, with the greatest improvements in those areas with simple guidelines to follow and in which the leaders were most participative. The features of high-reliability organisation were moderately identified in the nursing leaders' actions and views. For example, acting as a change agent to implement patient safety practices was challenging. Participatory methods can be used to support leaders into advancing patient safety. However, it is important that the participants are familiar with the method, and there are enough facilitators to steer development processes. Originality/value Research brings more knowledge of how leaders can increase their effectiveness in advancing patient safety and promoting high-reliability organisation features in the healthcare organisation.

  13. Development of a safety management protocol.

    Science.gov (United States)

    2008-09-01

    The UC Berkeley Traffic Safety Center (TSC) has produced this report under a contract from the California Department of Transportation : (Caltrans). The aim is to address workplace injuries and accidents among Caltrans employees and develop recommend...

  14. Developing safety in the nuclear fuel cycle

    International Nuclear Information System (INIS)

    Brown, M.L.

    1996-01-01

    The nuclear fuel cycle had its origins in the new technology developed in the 1940s and 50s involving novel physical and chemical processes. At the front end of the cycle, mining, milling and fuel fabrication all underwent development, but in general the focus of process development and safety concerns was the reprocessing stage, with radiation, contamination and criticality the chief hazards. Safety research is not over and there is still work to be done in advancing technical knowledge to new generation nuclear fuels such as Mixed Oxide Fuel and in refining knowledge of margins and of potential upset conditions. Some comments are made on potential areas for work. The NUCEF facility will provide many useful data to aid safety analysis and accident prevention. The routine operations in such plants, basically chemical factories, requires industrial safety and in addition the protection of workers against radiation or contamination. The engineering and management measures for this were novel and the early operation of such plants pioneering. Later commissioning and operating experience has improved routine operating safety, leading to a new generation of factories with highly developed worker protection, engineering safeguards and safety management systems. Ventilation of contamination control zones, remote operation and maintenance, and advanced neutron shielding are engineering examples. In safety management, dose control practices, formally controlled operating procedures and safety cases, and audit processes are comparable with, or lead, best industry practice in other hazardous industries. Nonetheless it is still important that the knowledge and experience from operating plants continue to be gathered together to provide a common basis for improvement. The NEA Working Group on Fuel Cycle Safety provides a forum for much of this interchange. Some activities in the Group are described in particular the FINAS incident reporting system. (J.P.N.)

  15. Unique differences in applying safety analyses for a graphite moderated, channel reactor

    International Nuclear Information System (INIS)

    Moffitt, R.L.

    1993-06-01

    Unlike its predecessors, the N Reactor at the Hanford Site in Washington State was designed to produce electricity for civilian energy use as well as weapons-grade plutonium. This paper describes the major problems associated with applying safety analysis methodologies developed for commercial light water reactors (LWR) to a unique reactor like the N Reactor. The focus of the discussion is on non-applicable LWR safety standards and computer modeling/analytical variances of standards. The approaches used to resolve these problems to develop safety standards and limits for the N Reactor are described

  16. Safety critical software development qualification

    International Nuclear Information System (INIS)

    Marron, J. E.

    2006-01-01

    With the increasing use of digital systems in control applications, customers must acquire appropriate expectations for software development and quality assurance procedures. Purchasers and users of digital systems need to understand the benefits to the supplier of effective quality systems. These systems consist not only of procedures but tools that enable automation. Without the use of automation, quality can not be assured. A software and systems quality program starts with the documents you are very familiar with. But these documents must define more than the final system. They must address specific development environment characteristics and testing capabilities. Starting with the RFP, some of the items that should be introduced are Software Configuration Management, regression testing and defect tracking. The digital system customer is in the best position to enforce the use of software and systems quality programs by including them in project requirements as early as the Purchase Order. The customer's understanding of the full scope and implementation of a software quality program is essential to achieving the quality necessary in nuclear projects, and, incidentally, completing those projects on schedule. (authors)

  17. Global road safety online course development.

    Science.gov (United States)

    2017-06-01

    The Global Road Safety Online Curriculum Development project involved the adaptation of in-person classroom materials and development of new materials to be used in an online setting. A short-course format was selected to pilot the course, and four t...

  18. Thermal Safety Analyses for the Production of Plutonium-238 at the High Flux Isotope Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hurt, Christopher J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Freels, James D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hobbs, Randy W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jain, Prashant K. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Maldonado, G. Ivan [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-08-01

    There has been a considerable effort over the previous few years to demonstrate and optimize the production of plutonium-238 (238Pu) at the High Flux Isotope Reactor (HFIR). This effort has involved resources from multiple divisions and facilities at the Oak Ridge National Laboratory (ORNL) to demonstrate the fabrication, irradiation, and chemical processing of targets containing neptunium-237 (237Np) dioxide (NpO2)/aluminum (Al) cermet pellets. A critical preliminary step to irradiation at the HFIR is to demonstrate the safety of the target under irradiation via documented experiment safety analyses. The steady-state thermal safety analyses of the target are simulated in a finite element model with the COMSOL Multiphysics code that determines, among other crucial parameters, the limiting maximum temperature in the target. Safety analysis efforts for this model discussed in the present report include: (1) initial modeling of single and reduced-length pellet capsules in order to generate an experimental knowledge base that incorporate initial non-linear contact heat transfer and fission gas equations, (2) modeling efforts for prototypical designs of partially loaded and fully loaded targets using limited available knowledge of fabrication and irradiation characteristics, and (3) the most recent and comprehensive modeling effort of a fully coupled thermo-mechanical approach over the entire fully loaded target domain incorporating burn-up dependent irradiation behavior and measured target and pellet properties, hereafter referred to as the production model. These models are used to conservatively determine several important steady-state parameters including target stresses and temperatures, the limiting condition of which is the maximum temperature with respect to the melting point. The single pellet model results provide a basis for the safety of the irradiations, followed by parametric analyses in the initial prototypical designs

  19. Development of safety analysis technology for integral reactor

    International Nuclear Information System (INIS)

    Kim, Hee Cheol; Kim, K. K.; Kim, S. H.

    2002-04-01

    The state-of-the-arts for the integral reactor was performed to investigate the safety features. The safety and performance of SMART were assessed using the technologies developed during the study. For this purpose, the computer code system and the analysis methodology were developed and the safety and performance analyses on SMART basic design were carried out for the design basis event and accident. The experimental facilities were designed for the core flow distribution test and the self-pressurizing pressurizer performance test. The tests on the 2-phase critical flow with non-condensable gas were completed and the results were used to assess the critical flow model. Probabilistic Safety Assessment(PSA) was carried out to evaluate the safety level and to optimize the design by identifying and remedying any weakness in the design. A joint study with KINS was carried out to promote licensing environment. The generic safety issues of integral reactors were identified and the solutions were formulated. The economic evaluation of the SMART desalination plant and the activities related to the process control were carried out in the scope of the study

  20. Safety demonstration analyses at JAERI for severe accident during overland transport of fresh nuclear fuel

    International Nuclear Information System (INIS)

    Nomura, Yasushi; Kitao, Kohichi; Karasawa, Kiyonori; Yamada, Kenji; Takahashi, Satoshi; Watanabe, Kohji; Okuno, Hiroshi; Miyoshi, Yoshinori

    2005-01-01

    It is expected in the near future that more and more fresh nuclear fuel will be transported in a variety of transport packages to cope with increasing demand from nuclear fuel cycle facilities. Accordingly, safety demonstration analyses are planned and conducted at JAERI under contract with the Ministry of Economy, Trade and Industry of Japan. These analyses are conducted in a four year plan from 2001 to 2004 to verify integrity of packaging against leakage of radioactive material in the case of a severe accident postulated to occur during transportation, for the purpose of gaining acceptance of such nuclear fuel activities. In order to create the accident scenarios, actual transportation routes were surveyed, accident or incident records were tracked, international radioactive material transport regulations such as IAEA rules were investigated and thus, accident conditions leading to mechanical damages and thermal failure were determined to characterize the scenarios. As a result, the worst-case conditions of run-off-the-road accidents were set up to define the impact against a concrete or asphalt surface. For fire accident scenarios to be set up, collisions were assumed to occur with an oil tanker carrying lots of inflammable material in open air, or with a commonly used two-ton-truck inside a tunnel without ventilation. Then the cask models were determined for these safety demonstration analyses to represent those commonly used for fresh nuclear fuel transported throughout Japan. Following the postulated accident scenarios, the mechanical damages were analyzed by using the general-purpose finite element code LS-DYNA with three-dimensional elements. It was found that leak tightness of the package be maintained even in the severe impact scenario. Then the thermal safety was analyzed by using the general-purpose finite element code ABAOUS with three-dimensional elements to describe cask geometry. As a result of the thermal analyses, the integrity of the containment

  1. Cyber safety education in developing countries

    CSIR Research Space (South Africa)

    Von Solms, R

    2015-07-01

    Full Text Available Cyber safety has become critical in today's world. Young children specifically need to be educated to operate in a safe manner in cyberspace and to protect themselves in the process. Unfortunately, African and developing countries do not necessarily...

  2. Cyber safety education in developing countries

    CSIR Research Space (South Africa)

    Von Solms, R

    2015-01-01

    Full Text Available Cyber safety has become critical in today's world. Young children specifically need to be educated to operate in a safe manner in cyberspace and to protect themselves in the process. Unfortunately, African and developing countries do not necessarily...

  3. Nuclear power development, safety and environmental problems

    International Nuclear Information System (INIS)

    Kasak, F.; Kadlec, J.

    1977-09-01

    The current state is described of power production by conventional power plants and the problems of burning fossil fuels are discussed. A survey is presented of the development of world nuclear power production and of the planned construction of nuclear power plants in Czechoslovakia. The questions of the safety of nuclear installations and their environmental impacts in normal operation and in case of accident are outlined. In the analysis of these aspects of nuclear power production the probability data on the potential hazards of operating nuclear reactors as published in the Rasmussen Safety Report are discussed. (O.K.)

  4. Assessing the validity of road safety evaluation studies by analysing causal chains.

    Science.gov (United States)

    Elvik, Rune

    2003-09-01

    This paper discusses how the validity of road safety evaluation studies can be assessed by analysing causal chains. A causal chain denotes the path through which a road safety measure influences the number of accidents. Two cases are examined. One involves chemical de-icing of roads (salting). The intended causal chain of this measure is: spread of salt --> removal of snow and ice from the road surface --> improved friction --> shorter stopping distance --> fewer accidents. A Norwegian study that evaluated the effects of salting on accident rate provides information that describes this causal chain. This information indicates that the study overestimated the effect of salting on accident rate, and suggests that this estimate is influenced by confounding variables the study did not control for. The other case involves a traffic club for children. The intended causal chain in this study was: join the club --> improve knowledge --> improve behaviour --> reduce accident rate. In this case, results are rather messy, which suggests that the observed difference in accident rate between members and non-members of the traffic club is not primarily attributable to membership in the club. The two cases show that by analysing causal chains, one may uncover confounding factors that were not adequately controlled in a study. Lack of control for confounding factors remains the most serious threat to the validity of road safety evaluation studies.

  5. Development of a measure of safety climate

    Directory of Open Access Journals (Sweden)

    N. R. Barnes

    1990-06-01

    Full Text Available A measure of safety climate was developed to aid management in identifying safety problems and responding proactively to safety issues; to assess the general mood of the workforce to safety; and as a standard for comparison with other organizations. The measure of safety climate was based on items extracted from the Chamber of Mines "Loss Control" audit manual. Reliability analysis performed on the scale indicated consistently high reliability coefficients across three ethnic groups. Factor analysis gave support for the construct validity of the scale. Opsomming 'n Meting vir veiligheidsklimaat is ontwikkel ten einde bestuur in staat te stel om veiligheidsprobleme te identifiseer en om pro-aktiefop te tree; om die algemene gevoel van die werkskragte rakende veiligheid te bepaal en om 'n maatstaf vir vergelyking met ander organisasies daar te stel. 'n Betroubaarheidssanalise wat op die skaal uitgevoer is het daarop gedui dat daar konsekwent hoe betroubaarheidskoefisiënte vir drie etniese groepe verkry word. 'n Faktoranalise het die konstrukgeldigheid van die skaal bevestig. The author acknowledges the financial assistance provided by the Human Sciences Research Council for this research.

  6. Scoping analyses for the safety injection system configuration for Korean next generation reactor

    International Nuclear Information System (INIS)

    Bae, Kyoo Hwan; Song, Jin Ho; Park, Jong Kyoon

    1996-01-01

    Scoping analyses for the Safety Injection System (SIS) configuration for Korean Next Generation Reactor (KNGR) are performed in this study. The KNGR SIS consists of four mechanically separated hydraulic trains. Each hydraulic train consisting of a High Pressure Safety Injection (HPSI) pump and a Safety Injection Tank (SIT) is connected to the Direct Vessel Injection (DVI) nozzle located above the elevation of cold leg and thus injects water into the upper portion of reactor vessel annulus. Also, the KNGR is going to adopt the advanced design feature of passive fluidic device which will be installed in the discharge line of SIT to allow more effective use of borated water during the transient of large break LOCA. To determine the feasible configuration and capacity of SIT and HPSl pump with the elimination of the Low Pressure Safety Injection (LPSI) pump for KNGR, licensing design basis evaluations are performed for the limiting large break LOCA. The study shows that the DVI injection with the fluidic device SlT enhances the SIS performance by allowing more effective use of borated water for an extended period of time during the large break LOCA

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

    Energy Technology Data Exchange (ETDEWEB)

    WILLIAMS, J.C.

    2003-11-15

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

  8. Developing software for safety-critical applications

    International Nuclear Information System (INIS)

    Chudleigh, M.

    1989-01-01

    The effective implementation of many safety-critical systems involves microprocessors running software which needs to be of very high integrity. This article describes some of the problems of producing such software and the place of software within the total system. A development strategy is proposed based on three principles: the goal of defect-free development, the use of mathematical formalism, and the use of an independent team for testing. (author)

  9. Probabilistic evaluation of scenarios in long-term safety analyses. Results of the project ISIBEL; Probabilistische Bewertung von Szenarien in Langzeitsicherheitsanalysen. Ergebnisse des Vorhabens ISIBEL

    Energy Technology Data Exchange (ETDEWEB)

    Buhmann, Dieter; Becker, Dirk-Alexander; Laggiard, Eduardo; Ruebel, Andre; Spiessl, Sabine; Wolf, Jens

    2016-07-15

    In the frame of the project ISIBEL deterministic analyses on the radiological consequences of several possible developments of the final repository were performed (VSG: preliminary safety analysis of the site Gorleben). The report describes the probabilistic evaluation of the VSG scenarios using uncertainty and sensitivity analyses. It was shown that probabilistic analyses are important to evaluate the influence of uncertainties. The transfer of the selected scenarios in computational cases and the used modeling parameters are discussed.

  10. Safety demonstration analyses for severe accident of fresh nuclear fuel transport packages at JAERI

    International Nuclear Information System (INIS)

    Yamada, K.; Watanabe, K.; Nomura, Y.; Okuno, H.; Miyoshi, Y.

    2004-01-01

    It is expected in the near future that more and more fresh nuclear fuel will be transported in a variety of transport packages to cope with increasing demand from nuclear fuel cycle facilities. Accordingly, safety demonstration analyses of these methods are planned and conducted at JAERI under contract with the Ministry of Economy, Trade and Industry of Japan. These analyses are conducted part of a four year plan from 2001 to 2004 to verify integrity of packaging against leakage of radioactive material in the case of a severe accident envisioned to occur during transportation, for the purpose of gaining public acceptance of such nuclear fuel activities. In order to create the accident scenarios, actual transportation routes were surveyed, accident or incident records were tracked, international radioactive material transport regulations such as IAEA rules were investigated and, thus, accident conditions leading to mechanical damage and thermal failure were selected for inclusion in the scenario. As a result, the worst-case conditions of run-off-the-road accidents were incorporated, where there is impact against a concrete or asphalt surface. Fire accidents were assumed to occur after collision with a tank truck carrying lots of inflammable material or destruction by fire after collision inside a tunnel. The impact analyses were performed by using three-dimensional elements according to the general purpose impact analysis code LS-DYNA. Leak-tightness of the package was maintained even in the severe impact accident scenario. In addition, the thermal analyses were performed by using two-dimensional elements according to the general purpose finite element method computer code ABAQUS. As a result of these analyses, the integrity of the inside packaging component was found to be sufficient to maintain a leak-tight state, confirming its safety

  11. Development of Safety Assessment Information System (SAIS)

    International Nuclear Information System (INIS)

    Park, Byung Shik; Lee, Kyung Jin; Lee, Byung Chul; Song, Tae Young; Lee, Chang Ho

    2007-01-01

    Many reports and documents about nuclear power plant safety analysis like a Periodic Safe Review (PSR), Periodic Safety Analysis (PSA) and Severe Accident Management Guideline (SAMG) come to be drawn up from KHNP. Since these are not arranged easy to look up, the systematic arrangement of data was necessary. The solution against hereupon is to store database, and it was developed with the name, SAIS, by FNC Tech. Co. together with NETEC KHNP. In this web program it is easy to manage (registration, search and statistics) data. And the authorized user can approach this system. This was developed, and was verified under the development environment of; - Web Server : Apache 2.2.5 - Program Language : PHP 5.2 - DBMS : Oracle 10g

  12. Development of design and safety analysis supporting system for casks

    International Nuclear Information System (INIS)

    Ohsono, Katsunari; Higashino, Akira; Endoh, Shuji

    1993-01-01

    Mitsubishi heavy Industries has developed a design and safety analysis supporting system 'CADDIE' (Cask Computer Aided Design, Drawing and Integrated Evaluation System), with the following objectives: (1) Enhancement of efficiency of the design and safety analysis (2) Further advancement of design quality (3) Response to the diversification of design requirements. The features of this system are as follows: (1) The analysis model data common to analyses is established, and it is prepared automatically from the model made by CAD. (2) The input data for the analysis code is available by simple operation of conversation type from the analysis model data. (3) The analysis results are drawn out in diagrams by output generator, so as to facilitate easy observation. (4) The data of material properties, fuel assembly data, etc. required for the analyses are made available as a data base. (J.P.N.)

  13. Preliminary standard review guide for Environmental Restoration/Decontamination and Decommissioning safety analyses

    International Nuclear Information System (INIS)

    Ellingson, D.R.

    1993-06-01

    The review guide is based on the shared experiences, approaches, and philosophies of the Environmental Restoration/Decontamination and Decommissioning (ER/D ampersand D) subgroup members. It is presented in the form of a review guide to maximize the benefit to both the safety analyses practitioner and reviewer. The guide focuses on those challenges that tend to be unique to ER/D ampersand D cleanup activities. Some of these experiences, approaches, and philosophies may find application or be beneficial to a broader spectrum of activities such as terminal cleanout or even new operations. Challenges unique to ER/D ampersand D activities include (1) consent agreements requiring activity startup on designated dates; (2) the increased uncertainty of specific hazards; and (3) the highly variable activities covered under the broad category of ER/D ampersand D. These unique challenges are in addition to the challenges encountered in all activities; e.g., new and changing requirements and multiple interpretations. The experiences in approaches, methods, and solutions to the challenges are documented from the practitioner and reviewer's perspective, thereby providing the viewpoints on why a direction was taken and the concerns expressed. Site cleanup consent agreements with predetermined dates for restoration activity startup add the dimension of imposed punitive actions for failure to meet the date. Approval of the safety analysis is a prerequisite to startup. Actions that increase expediency are (1) assuring activity safety; (2) documenting that assurance; and (3) acquiring the necessary approvals. These actions increase the timeliness of startup and decrease the potential for punitive action. Improvement in expediency has been achieved by using safety analysis techniques to provide input to the line management decision process rather than as a review of line management decisions. Expediency is also improved by sharing the safety input and resultant decisions with

  14. Scanning electron microscopic analyses of Ferrocyanide tank wastes for the Ferrocyanide safety program

    International Nuclear Information System (INIS)

    Callaway, W.S.

    1995-09-01

    This is Fiscal Year 1995 Annual Report on the progress of activities relating to the application of scanning electron microscopy in addressing the Ferrocyanide Safety Issue associated with Hanford Site high-level radioactive waste tanks. The status of the FY 1995 activities directed towards establishing facilities capable of providing SEM based micro-characterization of ferrocyanide tank wastes is described. A summary of key events in the SEM task over FY 1995 and target activities in FY 1996 are presented. A brief overview of the potential applications of computer controlled SEM analytical data in light of analyses of ferrocyanide simulants performed by an independent contractor is also presented

  15. Further development of the methodology for the realization of safety analyses concerning the controllability of operational malfunctions and accidents. Report on the working package 1. Review and development of safety-related assessment for final repositories for wastes with negligible heat generation and the provision of the necessary set of tools using the example of the final repository Konrad; Weiterentwicklung der Methodik fuer die Durchfuehrung von Sicherheitsanalysen zur Beherrschung von Betriebsstoerungen und Stoerfaellen. Bericht zum Arbeitspaket 1. Untersuchung und Entwicklung von sicherheitstechnischen Bewertungen fuer Endlager fuer Abfaelle mit vernachlaessigbarer Waermeentwicklung und Bereitstellung des notwendigen Instrumentariums am Beispiel des Endlagers Konrad

    Energy Technology Data Exchange (ETDEWEB)

    Hartwig-Thurat, Eva; Uhlmann, Stephan

    2015-09-15

    In the research project on the ''Review and development of safety-related assessments of disposal facilities with negligible heat generation; development and provision of the necessary set of tools, using the example of the Konrad disposal facility'' (Untersuchung und Entwicklung von sicherheitstechnischen Bewertungen fuer Endlager fuer Abfaelle mit vernachlaessigbarer Waermeentwicklung; Entwicklung und Bereitstellung des notwendigen Instrumentariums am Beispiel des Endlagers Konrad - Forschungsvorhaben 3612R03410), the state of the art in science and technology of the safety-related assessments and sets of tools for building a safety case was examined. The reports pertaining to the two work packages described the further development of the methodology for accident analyses (WP 1) and of building a safety case (WP 2); also, comparisons were drawn on a national and international scale with the methods applied in the licensing procedure of the Konrad disposal facility. As part of the project, the report of Work Package 1 depicts the methodology of the operating safety analysis in order to control malfunctions and incidents (accident analysis) using the example of the Konrad mine accident analysis. Set of criteria in this connection is the state-of-the-art international and national comprehensive body of legislation identifying the incident requirements. In extracts complementary safety analysis procedures of other countries are presented where applicable. It becomes apparent, that the majority of the investigated countries use a deterministic accident analyses to identify incidents. Here, common international practice is to com-plement the deterministic accident analysis by a probabilistic analysis. This procedure acts on the IAEA (International Atomic Energy Agency) terms of reference using both deterministic and probabilistic methods for the determination of facility hazard potentials. Based on the Konrad mine method, aspects of incident

  16. Maintenance: organizational modes, activities and health and safety. Use of a French national survey and in-situ analyses.

    Science.gov (United States)

    Grusenmeyer, Corinne

    2014-12-01

    Maintenance activities are identified as critical both to operator safety and to systems safety and reliability. However, it is still difficult to identify maintenance workers in French occupational accident and disease statistics. Moreover, few analyses of these activities and of organizational changes in this field have been conducted. This paper presents two different approaches to this same issue. Analyses were aimed firstly at identifying the occupational exposures of these operators and at comparing them with occupational exposures of production staff and, secondly at developing understanding of normal real maintenance activities, i.e. maintenance activities that are normally actually carried out, while taking into account the socio-technical system and maintenance organization within which they lie. The use of the French SUMER 2003 survey shows that occupational exposures of maintenance staff to various constraints are more frequent than occupational exposures of their production colleagues. However, maintenance staff appear to have greater independence. Analyses were also conducted in a subcontracting urban public transport company, who outsources some maintenance work. Those analyses highlight a complex network of companies involved in maintenance activities, a substantial number of work interruptions and a significant fragmentation of the internal technicians' activities that can be cognitively costly, reduce anticipation possibilities and lead to incidents or accidents. Above all they underline internal technicians' contributions to the completion of outsourced interventions and interdependent relationships between the activities of the internal and the external technicians. Outsourcing maintenance interventions thus raises the question of risks associated with the interdependence of actual work activities undertaken by the different types of staff, since they contribute to the same maintenance intervention. This study therefore pinpoints the need to

  17. Urban development in developing countries: analysing current policies for Mumbai

    OpenAIRE

    Byrd, Hugh; Mandal, Anindita

    2011-01-01

    Urban areas in developing countries have been growing exponentially. In a list of cities with 5 million people or more, Mumbai did not feature in 1950, but in 1975 it occupied the 15th position which changed to 5th in 2000 and is expected to be 3rd by 2015. The rapid and continued growth of Mumbai and its sprawling metropolitan regions has put considerable strain on the city’s infrastructure and caused environmental degradation. Not only is Mumbai one of the most crowded cities in the wor...

  18. OSH for development : occupational safety and health for development

    OpenAIRE

    2009-01-01

    This book is written for professionals in developing countries, for occupational safety and health specialists, production engineers, managers and trade union representatives. The book links occupational safety and health with production and productivity, showing the positive relationship between good and safe working conditions and sustainable high productivity. The book makes use of research results, and is intended to be used in connection with training activities, but does not concentrate...

  19. Development of the reactor safety film

    International Nuclear Information System (INIS)

    Sheheen, N.N.; Hodson, P.J.

    1981-01-01

    The first computer-generated film of LASL's Reactor Safety efforts was developed using the ANIMATE framework, a program that adds visual capabilities to MAPPER. Numerous software limitations had to be overcome within a very limited production schedule. A significant achievement was the 15,000-vector-per-frame sequence depicting a pressurized water reactor core with parts flashing while pumps circulate fluid through the system

  20. Development of French technical safety regulations: safety fundamental rules

    International Nuclear Information System (INIS)

    Lebouleux, P.

    1983-01-01

    The technical regulation related to nuclear safety in France is made of a set of regulation texts, of a different nature, that define the requirements for the construction, commissioning and operating of nuclear facilities. Simultaneously, the safety authorities (Service Central de Surete des Installations Nucleaires: SCSIN) issue recommendations or guides which are not strictly speaking regulations in the juridicial sense; they are called Regles Fondamentales de Surete (RFS). The RFS set up and detail the conditions, the respect of which is deemed to be complying with the French regulation practice, for the subject to which they relate. Their purpose is to make known rules judged acceptable by safety authorities, thus making the safety review easier. The RFS program is described. A RFS - or a letter - can also give the result of the examination of the constructor and operator codes (RCC) by safety authorities

  1. Development and formation of safety cultures

    International Nuclear Information System (INIS)

    Merry, M.W.J.; Rycraft, H.S.

    1995-01-01

    The Thermal Oxide Reprocessing Plant (THORP) is the largest project ever undertaken by British Nuclear Fuels plc (BNFL) and its success is important for the future of the company. The company recognised at the planning stage that to be profitable, THORP had to operate both safely and with a smaller workforce. The establishment of an appropriate culture which saw safety and productivity as essential and complimentary at the beginning of the life of the plant was therefore vital for the future success of THORP The key factors in the THORP Culture formation were : The recruitment policy; the training policy; measures taken to ensure participation from the workforce; teamworking support; communication initiatives; clear statement of cultural principles; clear and demonstrable leadership. The current stage of evolution has seen some positive results namely: A clear commitment to involving all personnel in problem solving and task organisation, including safety; a confident workforce with an improved ability to communicate; the capability of the majority of the workforce to work as a team; safety awareness of the workforce is generally high along with an awareness of environmental, commercial and (political) external issues affecting the THORP business; a commitment to continuous improvement. The development of the safety culture within THORP has also had challenges, some as a result of the composite nature of the workforce, and others as side effects of the culture shaping measures. Management have recognised these, and using the results of attitude surveys, are working with the workforce to overcome their effects. Clear recognition has been achieved that the establishment of positive behaviours is a key. step in generating the culture required summarising, there is recognition that the design of safety management systems and improvement programmes, should be based on the principles of human psychology and behaviour. which includes wide participation by the workforce

  2. Developments in safety standards and regulation

    International Nuclear Information System (INIS)

    Harbison, S.A.

    1994-01-01

    This paper explains, in broad terms, how regulatory control is exercised over licensed nuclear installations in the UK and how HSE has developed its safety standards to support its regulatory approach. It first sets out the scope of HSE's regulatory responsibilities, which NII exercises on its behalf, and briefly describes the licensing process and compliance monitoring through inspection over the life of a nuclear plant. It also refers to the role of assessment in NII's decision-making processes, and the part played in this by the consideration of costs and safety benefits. It then moves on to consider the challenges that HSE/NII are likely to face from the changing nuclear industry in the second half of the 1990s. (author)

  3. Development and Validation of a Safety Attitude Scale for Coal Miners in China

    Directory of Open Access Journals (Sweden)

    Xiang Wu

    2017-11-01

    Full Text Available Safety attitude is of vital importance to accident prevention, and the high accident rate in the coal mining industry makes it urgent to undertake research on coal miners’ safety attitude. However, the current literature still lacks a valid and reliable safety attitude measurement scale for coal miners, which stands as a barrier against their safety attitude improvement. In this study, a scale is developed that can be used to measure coal miners’ safety attitude. The preliminary scale was based on an extended literature review. Empirical data were then collected from 725 coal miners using the preliminary scale. Both exploratory and confirmatory factor analyses were undertaken to validate and improve the scale. The final scale, which consists of 17 items, contains four dimensions: management safety commitment, team safety climate, fatalism and work pressure. Results show that this safety attitude scale can effectively measure the safety attitude of coal miners, showing high psychological measurement validity. This paper contributes to the occupational safety research by developing the factor structure and indicator system of coal miners’ safety attitude, thus providing more profound interpretation of this crucial construct in the safety research domain. The measurement scale serves as an important tool for safety attitude benchmarking among different coal mining enterprises and, thus, can boost the overall safety improvement of the whole industry. These findings can facilitate improvement of both theories and practices related to occupational safety attitude.

  4. Swiss regulatory use of databanks for nuclear power plant life management, surveillance and safety analyses

    International Nuclear Information System (INIS)

    Tipping, Ph.; Beutler, R.; Schoen, G.; Noeggerath, J.

    2002-01-01

    Full text: As operational time is accumulated, the overall safety and performance of nuclear power plants (NPPs) will tend to be characterised by those areas in which structures, systems and components (SSCs) have not performed as well, or as reliably, as expected. The reasons for non-availability of equipment in NPPs due to SSC material malfunction or unsatisfactory performance, leading to events or even accidents, are varied and they must be analysed in order to obtain the root causes. Once the root causes are identified, corresponding measures can be applied in order to improve reliability and therefore safety. The root cause information obtained, if brought into user-friendly databanks (DBs), can be used to follow NPP performance trends, to check whether a repair or replacement has been effective, to focus regulatory attention and NPP surveillance on known weak-spots and to serve as an advance indicator where potential problems may arise. Using the DBs, similar occurrences of failures or problems in other NPPs can be identified and generic issues recognised early on and preventative action taken. The following describes the Swiss Federal Nuclear Safety Inspectorate's (HSK) DB concepts for keeping track of NPP safety and lifetime management issues. Typical sources of data for the Inspectorate's DBs are, for example, the IAEA/NEA Incident Reporting System (IRS) reports, US-NRC Generic Letters, the Swiss NPP's own reports (monthly, annual and normal outage) and, more importantly, the document that these NPPs must issue to the Inspectorate whenever a reportable event takes place. Specifically, the reporting of events in the NPPs is laid down in the Inspectorate's Guideline (R-15 'Reporting Guideline Concerning The Operation of Nuclear Power Plants'). In this Guideline, reportable events are defined and the criteria for assessing the degree of importance or impact on nuclear safety are given. In this manner, a standard and consistent approach to data collection is

  5. Development and assessment of best estimate integrated safety analysis code

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Bub Dong; Lee, Young Jin; Hwang, Moon Kyu (and others)

    2007-03-15

    Improvement of the integrated safety analysis code MARS3.0 has been carried out and a multi-D safety analysis application system has been established. Iterative matrix solver and parallel processing algorithm have been introduced, and a LINUX version has been generated to enable MARS to run in cluster PCs. MARS variables and sub-routines have been reformed and modularised to simplify code maintenance. Model uncertainty analyses have been performed for THTF, FLECHT, NEPTUN, and LOFT experiments as well as APR1400 plant. Participations in international cooperation research projects such as OECD BEMUSE, SETH, PKL, BFBT, and TMI-2 have been actively pursued as part of code assessment efforts. The assessment, evaluation and experimental data obtained through international cooperation projects have been registered and maintained in the T/H Databank. Multi-D analyses of APR1400 LBLOCA, DVI Break, SLB, and SGTR have been carried out as a part of application efforts in multi-D safety analysis. GUI based 3D input generator has been developed for user convenience. Operation of the MARS Users Group (MUG) was continued and through MUG, the technology has been transferred to 24 organisations. A set of 4 volumes of user manuals has been compiled and the correction reports for the code errors reported during MARS development have been published.

  6. Development of NUMO safety case for geological disposal

    International Nuclear Information System (INIS)

    Suzuki, Satoru; Deguchi, Akira

    2016-01-01

    NUMO has developed a generic safety ease based on the latest knowledge to show the feasibility and safety of geological disposal in Japan. The NUMO safety case has been developed to provide a basic structure for subsequent safety cases that would be applied to any selected site, emphasising practical approaches and methodology, which will be applicable for the conditions/constraints during an actual siting process. This paper will provide a brief overview of the NUMO safety case. (author)

  7. Development of an approach for the analysis of network technologies in safety related instrumentation and control systems with respect to the propagation and effect of postulated failures; Entwicklung eines Ansatzes zur Analyse der Netzwerktechnologien in sicherheitsrelevanten Leittechniksystemen hinsichtlich Verbreitung und Auswirkung postulierter Fehler

    Energy Technology Data Exchange (ETDEWEB)

    Herb, Joachim; Jopen, Manuela; Lindner, Falk; Piljugin, Ewgenij; Vogt, Pascal

    2015-06-15

    So far, safety related instrumentation and control (I and C) functions in nuclear power plants, such as controlling of safety systems, were mostly performed by conventional (analog) I and C equipment. For some years now, I and C systems and equipment in nuclear power plants worldwide, but also in Germany, are modernized by computer-based I and C systems. In signal processing of the computer-based I and C systems, modern network technologies are used both for internal and external communication, whereas the reliability and safety for information transfer and processing plays an important role. National and interna-tional operational experience shows a significant influence of communication in a net-worked I and C system on its reliability. The aim of the GRS within the project 361R01351 ''Development of an approach for an analysis of network technologies in safety related I and C systems in view of distribution and effect of postulated failures'' was to improve the expertise in the field of network communication, to investigate phenomenologically potential sources of failures and fault propagation paths (Network failures) in a generic I and C system as well as to develop methodic approaches for analyses of propagation and effect of postulated failures in typical networks. The GRS conducted extensive research in the field of ''Data communication in digital I and C systems''. In this report, the basic principles of data communication of computer-based I and C systems are presented. This includes, among other things, network topolo-gies, communication protocols and standards as well as generic failures. Additionally, the properties of different analysis methods and its applicability for reliability analyses of network communication in computer-based I and C systems are discussed. Based on state of the art evaluation, an analysis approach was developed, which takes into account the specific properties of network communication and

  8. New developments enhancing MCNP for criticality safety

    International Nuclear Information System (INIS)

    Hendricks, J.S.; McKinney, G.W.; Forster, R.A.

    1993-01-01

    Since the early 80's MCNP has had three estimates of k eff : collision, absorption, and track length. MCNP has also had collision and absorption estimators of removal lifetime. These are calculated for every cycle and are averaged over the cycles as simple averages and covariance weighted averages. Correlation coefficients between estimators are also calculated. These criticality estimators are all in addition to the extensive summary information and tally edits used in shielding and other problems. A number of significant new developments have been made to enhance the MCNP Monte Carlo radiation transport code for criticality safety applications. These are available in the newly released MCNP4A version of the code

  9. Framework Development Supporting the Safety Portal

    Energy Technology Data Exchange (ETDEWEB)

    Prescott, Steven Ralph [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kvarfordt, Kellie Jean [Idaho National Lab. (INL), Idaho Falls, ID (United States); Vang, Leng [Idaho National Lab. (INL), Idaho Falls, ID (United States); Smith, Curtis Lee [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-07-01

    In a collaborating scientific research arena it is important to have an environment where analysts have access to a shared repository of information, documents, and software tools, and be able to accurately maintain and track historical changes in models. The new Safety Portal cloud-based environment will be accessible remotely from anywhere regardless of computing platforms given that the platform has available Internet access and proper browser capabilities. Information stored at this environment would be restricted based on user assigned credentials. This report discusses current development of a cloud-based web portal for PRA tools.

  10. Reentry safety for the Topaz II Space Reactor: Issues and analyses

    Energy Technology Data Exchange (ETDEWEB)

    Connell, L.W.; Trost, L.C.

    1994-03-01

    This report documents the reentry safety analyses conducted for the TOPAZ II Nuclear Electric Propulsion Space Test Program (NEPSTP). Scoping calculations were performed on the reentry aerothermal breakup and ground footprint of reactor core debris. The calculations were used to assess the risks associated with radiologically cold reentry accidents and to determine if constraints should be placed on the core configuration for such accidents. Three risk factors were considered: inadvertent criticality upon reentry impact, atmospheric dispersal of U-235 fuel, and the Special Nuclear Material Safeguards risks. Results indicate that the risks associated with cold reentry are very low regardless of the core configuration. Core configuration constraints were therefore not established for radiologically cold reentry accidents.

  11. Reentry safety for the Topaz II Space Reactor: Issues and analyses

    International Nuclear Information System (INIS)

    Connell, L.W.; Trost, L.C.

    1994-03-01

    This report documents the reentry safety analyses conducted for the TOPAZ II Nuclear Electric Propulsion Space Test Program (NEPSTP). Scoping calculations were performed on the reentry aerothermal breakup and ground footprint of reactor core debris. The calculations were used to assess the risks associated with radiologically cold reentry accidents and to determine if constraints should be placed on the core configuration for such accidents. Three risk factors were considered: inadvertent criticality upon reentry impact, atmospheric dispersal of U-235 fuel, and the Special Nuclear Material Safeguards risks. Results indicate that the risks associated with cold reentry are very low regardless of the core configuration. Core configuration constraints were therefore not established for radiologically cold reentry accidents

  12. Safety analyses of reactivity initiated accidents for planned power uprate of NPP Paks

    International Nuclear Information System (INIS)

    Kereszturi, A.; Hegyi, Gy.; Panka, I.; Telbisz, M.

    2004-01-01

    In the last years NPP Paks initiated several projects aiming at the introduction of new fuel types resulting in increased maximum allowed power and more economic fuel cycles. Modifications of the absorber assembly coupler part, increased average enrichment, modification of the lattice pitch, profiled enrichment are the most important changes. The increased power and parallel application of new fuel types require the renewal of the relevant chapters of the Safety Analysis Report, which is necessary because of the modified core nuclear properties, reloading schemes, power histories, thermal-mechanical conditions. The Fuel System and Accident Analyses are the most important chapters from this point of view. The fulfillment of the fuel design basis requirements, summarized briefly also in the paper, must be investigated during the normal and the selected accidental conditions. Core design, thermal hydraulic and reactor physics accident analysis, stationary and transient fuel behavior calculations are necessary for proving the fulfillment of the acceptance criteria detailed in the Appendices. The characteristics of the different codes, the data transfer between them are detailed. Due to the modified reloading schemes and reactor physics characteristics, reactivity initiated accidents requiring 3D kinetic analyses coupled to the system thermal hydraulic calculation are of special importance from this point of view. The analysis of the initiating event starting of one inactive loop is presented as a characteristic example. The following conclusions were drawn: Sound basis of licensing methodology, safety analysis, and necessary computer codes for the VVER power uprate and parallel fuel modernization is available; The potential of the computer codes is demonstrated by the analyses of the initiating event startup of one inactive loop, that requiring coupled 3D neutronics and thermal hydraulic calculations; The asymmetric flux distribution must be taken into account for the

  13. Development of regulatory technology for thermal-hydraulic safety analysis

    International Nuclear Information System (INIS)

    Bang, Young Seok; Lee, S. H.; Ryu, Y. H.

    2001-02-01

    The present study aims to develop the regulation capability in thermal-hydraulic safety analysis which was required for the reasonable safety regulation in the current NPP, the next generation reactors, and the future-type reactors. The fourth fiscal year of the first phase of the research was focused on the following research topics: Investigation on the current status of the thermal-hydraulic safety analysis technology outside and inside of the country; Review on the improved features of the thermal-hydraulic safety analysis regulatory audit code, RELAP5/MOD3; Assessments of code with LOFT L9-3 ATWS experiment and LSTF SB-SG-10 multiple SGTR experiment; Application of the RELAP5/CANDU code to analyses of SLB and LBLOCA and evaluation of its effect on safety; Application of the code to IAEA PHWR ISP analysis; Assessments of RELAP5 and TRAC with UPTF downcomer injection test and Analysis of LBLOCA with RELAP5 for the performance evaluation of KNGR DVI; Setup of a coupled 3-D kinetics and thermal-hydraulics and application it to a reactivity accident analysis; and Extension of database and improvement of plant input decks. For supporting the resolution of safety issues, loss of RHR event during midloop operation was analyzed for Kori Unit 3, issues on high burnup fuel were reviewed and performance of FRAPCON-3 assessed. Also MSLB was analyzed to figure out the sensitivity of downcomer temperature supporting the PTS risk evaluation of Kori Unit 1. Thermal stratification in pipe was analyzed using the method proposed. And a method predicting the thermal-hydraulic performance of IRWST of KNGR was explored. The PWR ECCS performance criteria was issued as a MOST Article 200-19.and a regulatory guide on evaluation methodology was improved to cover concerns raised from the related licensing review process

  14. Safety

    International Nuclear Information System (INIS)

    2001-01-01

    This annual report of the Senior Inspector for the Nuclear Safety, analyses the nuclear safety at EDF for the year 1999 and proposes twelve subjects of consideration to progress. Five technical documents are also provided and discussed concerning the nuclear power plants maintenance and safety (thermal fatigue, vibration fatigue, assisted control and instrumentation of the N4 bearing, 1300 MW reactors containment and time of life of power plants). (A.L.B.)

  15. Safety analyses of potential exposure in medical irradiation plants by Fuzzy Fault Tree

    International Nuclear Information System (INIS)

    Casamirra, Maddalena; Castiglia, Francesco; Giardina, Mariarosa; Tomarchio, Elio

    2008-01-01

    The results of Fuzzy Fault Tree (FFT) analyses of various accidental scenarios, which involve the operators in potential exposures inside an High Dose Rate (HDR) remote after-loading systems for use in brachytherapy, are reported. To carry out fault tree analyses by means of fuzzy probabilities, the TREEZZY2 computer code is used. Moreover, the HEART (Human Error Assessment and Reduction Technique) model, properly modified on the basis of the fuzzy approach, has been employed to assess the impact of performances haping and error-promoting factors in the context of the accidental events. The assessment of potential dose values for some identified accidental scenarios allows to consider, for a particular event, a fuzzy uncertainty range in potential dose estimate. The availability of lower and upper limits allows evaluating the possibility of optimization of the installation from the point of view of radiation protection. The adequacy of the training and information program for staff and patients (and their family members) and the effectiveness of behavioural rules and safety procedures were tested. Some recommendations on procedures and equipment to reduce the risk of radiological exposure are also provided. (author)

  16. Development of a new methodology for quantifying nuclear safety culture

    International Nuclear Information System (INIS)

    Han, Kiyoon; Jae, Moosung

    2017-01-01

    The present study developed a Safety Culture Impact Assessment Model (SCIAM) which consists of a safety culture assessment methodology and a safety culture impact quantification methodology. The SCIAM uses a safety culture impact index (SCII) to monitor the status of safety culture of NPPs periodically and it uses relative core damage frequency (RCDF) to present the impact of safety culture on the safety of NPPs. As a result of applying the SCIAM to the reference plant (Kori 3), the standard for the healthy safety culture of the reference plant is suggested. SCIAM might contribute to improve the safety of NPPs (Nuclear Power Plants) by monitoring the status of safety culture periodically and presenting the standard of healthy safety culture.

  17. Numerical and experimental analyses of lighting columns in terms of passive safety

    Science.gov (United States)

    Jedliński, Tomasz Ireneusz; Buśkiewicz, Jacek

    2018-01-01

    Modern lighting columns have a very beneficial influence on road safety. Currently, the columns are being designed to keep the driver safe in the event of a car collision. The following work compares experimental results of vehicle impact on a lighting column with FEM simulations performed using the Ansys LS-DYNA program. Due to high costs of experiments and time-consuming research process, the computer software seems to be very useful utility in the development of pole structures, which are to absorb kinetic energy of the vehicle in a precisely prescribed way.

  18. Japanese standard method for safety evaluation using best estimate code based on uncertainty and scaling analyses with statistical approach

    International Nuclear Information System (INIS)

    Mizokami, Shinya; Hotta, Akitoshi; Kudo, Yoshiro; Yonehara, Tadashi; Watada, Masayuki; Sakaba, Hiroshi

    2009-01-01

    Current licensing practice in Japan consists of using conservative boundary and initial conditions(BIC), assumptions and analytical codes. The safety analyses for licensing purpose are inherently deterministic. Therefore, conservative BIC and assumptions, such as single failure, must be employed for the analyses. However, using conservative analytical codes are not considered essential. The standard committee of Atomic Energy Society of Japan(AESJ) has drawn up the standard for using best estimate codes for safety analyses in 2008 after three-years of discussions reflecting domestic and international recent findings. (author)

  19. Development of a nuclear ship safety philosophy

    International Nuclear Information System (INIS)

    Thompson, T.E.

    1978-01-01

    A unique safety philosophy must be recognized and accepted as an integral part of the design and operation of a nuclear ship. For the nuclear powered ship, the ultimate safety of the reactor and therefore the crew and the environment lies with the safety of the ship itself. The basis for ship safety is its ability to navigate and survive the conditions or the environment in which it may find itself. The subject of traditional ship safety is examined along with its implication for reactor protection and safety. Concepts of reactor safety are also examined. These two philosophies are combined in a manner so as to provide a sound philosophy for the safety of nuclear ships, their crews, and the environment

  20. Development of Network Protocol for the Integrated Safety System

    International Nuclear Information System (INIS)

    Park, S. W.; Baek, J. I.; Lee, S. H.; Park, C. S.; Park, K. H.; Shin, J. M.

    2007-06-01

    Communication devices in the safety system of nuclear power plants are distinguished from those developed for commercial purposes in terms of a strict requirement of safety. The concept of safety covers the determinability, the reliability, and the separation/isolation to prevent the undesirable interactions among devices. The safety also requires that these properties be never proof less. Most of the current commercialized communication products rarely have the safety properties. Moreover, they can be neither verified nor validated to satisfy the safety property of implementation process. This research proposes the novel architecture and protocol of a data communication network for the safety system in nuclear power plants

  1. Development of Network Protocol for the Integrated Safety System

    Energy Technology Data Exchange (ETDEWEB)

    Park, S. W.; Baek, J. I.; Lee, S. H.; Park, C. S.; Park, K. H.; Shin, J. M. [Hannam Univ., Daejeon (Korea, Republic of)

    2007-06-15

    Communication devices in the safety system of nuclear power plants are distinguished from those developed for commercial purposes in terms of a strict requirement of safety. The concept of safety covers the determinability, the reliability, and the separation/isolation to prevent the undesirable interactions among devices. The safety also requires that these properties be never proof less. Most of the current commercialized communication products rarely have the safety properties. Moreover, they can be neither verified nor validated to satisfy the safety property of implementation process. This research proposes the novel architecture and protocol of a data communication network for the safety system in nuclear power plants.

  2. Analysis of adverse events as a contribution to safety culture in the context of practice development

    Science.gov (United States)

    Hoffmann, Susanne; Frei, Irena Anna

    2017-01-01

    Background: Analysing adverse events is an effective patient safety measure. Aim: We show, how clinical nurse specialists have been enabled to analyse adverse events with the „Learning from Defects-Tool“ (LFD-Tool). Method: Our multi-component implementation strategy addressed both, the safety knowledge of clinical nurse specialists and their attitude towards patient safety. The culture of practice development was taken into account. Results: Clinical nurse specialists relate competency building on patient safety due to the application of the LFD-tool. Applying the tool, fosters the reflection of adverse events in care teams. Conclusion: Applying the „Learning from Defects-Tool“ promotes work-based learning. Analysing adverse events with the „Learning from Defects-Tool“ contributes to the safety culture in a hospital.

  3. Propagation of Isotopic Bias and Uncertainty to Criticality Safety Analyses of PWR Waste Packages

    Energy Technology Data Exchange (ETDEWEB)

    Radulescu, Georgeta [ORNL

    2010-06-01

    predicted spent fuel compositions (i.e., determine the penalty in reactivity due to isotopic composition bias and uncertainty) for use in disposal criticality analysis employing burnup credit. The method used in this calculation to propagate the isotopic bias and bias-uncertainty values to k{sub eff} is the Monte Carlo uncertainty sampling method. The development of this report is consistent with 'Test Plan for: Isotopic Validation for Postclosure Criticality of Commercial Spent Nuclear Fuel'. This calculation report has been developed in support of burnup credit activities for the proposed repository at Yucca Mountain, Nevada, and provides a methodology that can be applied to other criticality safety applications employing burnup credit.

  4. Solubility of radionuclides in a bentonite environment for provisional safety analyses for SGT-E2

    International Nuclear Information System (INIS)

    Berner, U.

    2014-08-01

    Within stage 2 of the sectoral plan for deep geological repositories for radioactive waste in Switzerland provisional safety analyses are carried out. In the case of the repository for spent fuel and vitrified high level waste considered, retention mechanisms include the concentration limits of safety relevant elements in the pore water of the buffer material (bentonite). The present work describes the solubility limits of the safety relevant elements Be, C inorg , Cl, K, Ca, Co, Ni, Se, Sr, Zr, Nb, Mo, Tc, Pd, Ag, Sn, I, Cs, Sm, Eu, Ho, Pb, Po, Ra, Ac, Th, Pa, U, Np, Pu, Am and Cm in the pore water of bentonite after diffusive solution exchange with the host rock Opalinus Clay. The term solubility limit denotes the maximum amount of an element dissolving in the pore solution of the considered chemical reference system. Chemical equilibrium thermodynamics is the classical tool used for quantifying such considerations. For a given solid phase equilibrium thermodynamics predict the amount of substance dissolving in the solution and describe the speciation of the considered element in solution. The principles of chemical equilibrium will also be the primary work hypothesis in the present work. Solubility calculations were performed with the most recent version of GEMS/PSI (GEMS3.2 v.890) using the PSI/Nagra Chemical Thermodynamic Data Base 12/07, which is an update of the former Nagra/PSI Chemical Thermodynamic Data Base 01/01. The database was complemented with datasets from the ThermoChimie v. 7b for elements that were not considered in the mentioned update (Ag, Co, Sm, Ho, Pa, Be), with data from Iupac (Pb) and with data from the literature (Mo). Differing sources for thermodynamic data are noted. Reference values as well as lower and upper guideline values are evaluated. For many formation constants of solids and solutes uncertainties are known and allow conveying lower and upper guideline values. In many cases it is not clear whether the most stable solid is

  5. 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.

  6. Development of safety performance indicators for HANARO

    International Nuclear Information System (INIS)

    Wu, Jong-Sup; Jung, Hoan-Sung; Ahn, Guk-Hoon; Lee, Kye-Hong; Lim, In-Cheol

    2007-01-01

    The nuclear facilities need an extensive basis for ensuring their safety. An operating organization should conduct its operation and utilization important to the safety in accordance with approved procedures and regulations. The general aims of a management system for nuclear facilities are to improve the safety performance through a planning, control and supervision of safety related activities and to foster a strong safety culture. The effectiveness of a management system can be monitored and measured to confirm the ability of its processes to achieve the intended safety performance by an assessment of the operational performance. The Operational Safety Performance Indicators, also known as SPI, help an organization define and measure a progress with regard to safety activity goals. The elements of a SPI are quantifiable measurements that reflect the critical success factors of an organizational safety. Since 1995, efforts have been directed towards the elaboration of a framework for the establishment of an operational safety performance indicator program in nuclear power plants (NPP). IAEA-TECDOC-1141, 'Operational safety performance indicators for NPP' attempted to provide a frame work for an identification of performance indicators which have a relationship to the desired safety attributes, and therefore, to a safe plant operation. Three key attributes of a smooth operation, an operation with a low risk, and an operation with a positive safety attitude, were recommended, which are associated with a safe operation. Because these attributes cannot be directly measured, an indicator structure is expanded further until a level of easily quantifiable or directly measurable indicators is identified. The intention of this approach is to use quantitative information provided by the specific indicators and to analyze performance trends relative to established goals. The safety activities in HANARO have been continuously conducted to enhance its safe operation. HANARO

  7. Assessing progress in the development of safety culture

    International Nuclear Information System (INIS)

    Rotaru, Ioan; Ghita, Sorin

    1999-01-01

    visible prescriptive formula for developing a strong safety culture. However, a prerequisite is genuine and consistent commitment by the top management of an organization to improving safety . Providing this commitment exists, the best recommendation is to due something tangible and visible to improve safety, preferably involving employees from the outset. The choice of practices for developing an improved safety culture should take account of the existing national and organizational culture in order to ensure effective implementation. The importance of the learning process has been emphasized. A mechanism is necessary to ensure that international experience of practices to develop a strong safety culture is shared on a regular and frequent basis. The maintenance and improvement of a safety culture is a process of continuous evolution. Indicators are available to assess positive progress in this evolution and to detect a weakening safety culture. (authors)

  8. Producing health, producing safety. Developing a collective safety culture in radiotherapy

    International Nuclear Information System (INIS)

    Nascimento, Adelaide

    2009-01-01

    This research thesis aims at a better understanding of safety management in radiotherapy and at proposing improvements for patient safety through the development of a collective safety culture. A first part presents the current context in France and abroad, addresses the transposition of other safety methods to the medical domain, and discusses the peculiarities of radiotherapy in terms of risks and the existing quality-assurance approaches. The second part presents the theoretical framework by commenting the intellectual evolution with respect to system safety and the emergence of the concept of safety culture, and by presenting the labour collective aspects and their relationship with system safety. The author then comments the variety of safety cultures among the different professions present in radiotherapy, highlights the importance of the collective dimension in correcting discrepancies at the end of the treatment process, and highlights how physicians take their colleagues work into account. Recommendations are made to improve patient safety in radiotherapy

  9. Development of inspection safety evaluation technology

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Seok Chul; Yoon, Yeo Chang; Kim, Jong Soo; Lee, Tae Young; Kim, Chang Ryol; Lee, Hyung Sub; Kim, Jong Soo

    1995-12-01

    The purpose of this project is to protection nation inspector`s over exposure from radiation that can be occurred by inspection activity at nuclear facilities and its environment, and to ensure the safety of inspection activity at the nuclear facilities. To effectively carry out the domestic inspection task to be enforced from 1996, the evaluation for special radiation exposure rate of nuclear facilities, air and surface contamination level, and measurement and monitoring of water contamination level were made to determine whether these measured values exceeded permissible limitations, and to protect the inspector`s over exposure from radiation at domestic nuclear facilities. Management of inspector`s exposure was carried out under assistance of the Department of Health Physics. Performance tests of two gamma detectors, one neutron detector, alpha and beta detector, and gamma spectroscopy analyzer were carried out to control dose on extremity, the characteristic test for extremity dosimeter was carried out and the theoretical calculation of gamma dose conversion factors based on ANSI N13.32 standard was performed. Under the 93+2 program, IAEA began to recognize the necessity of environmental observation technology development of air-borne particulates travelled from long distance location. Associated with the necessity of this technology development, a proposal of international joint research for development of the special radiation measurement and analysis has been prepared. (author). 21 tabs., 24 figs., 20 refs.

  10. History of aviation safety; the satisfying sighs of relief due to developments in Aviation safety

    NARCIS (Netherlands)

    Stoop, J.A.A.M.

    2014-01-01

    ”Aviation safety is an Integral part of my career. Being part of TU Delft’s impressive record of research on Aviation safety, my career has been with a sense of purpose and a responsibility to equip students to deal with the status quo challenges on Aviation safety, developments, Investigations and

  11. Provisional safety analyses for SGT stage 2 -- Models, codes and general modelling approach

    International Nuclear Information System (INIS)

    2014-12-01

    In the framework of the provisional safety analyses for Stage 2 of the Sectoral Plan for Deep Geological Repositories (SGT), deterministic modelling of radionuclide release from the barrier system along the groundwater pathway during the post-closure period of a deep geological repository is carried out. The calculated radionuclide release rates are interpreted as annual effective dose for an individual and assessed against the regulatory protection criterion 1 of 0.1 mSv per year. These steps are referred to as dose calculations. Furthermore, from the results of the dose calculations so-called characteristic dose intervals are determined, which provide input to the safety-related comparison of the geological siting regions in SGT Stage 2. Finally, the results of the dose calculations are also used to illustrate and to evaluate the post-closure performance of the barrier systems under consideration. The principal objective of this report is to describe comprehensively the technical aspects of the dose calculations. These aspects comprise: · the generic conceptual models of radionuclide release from the solid waste forms, of radionuclide transport through the system of engineered and geological barriers, of radionuclide transfer in the biosphere, as well as of the potential radiation exposure of the population, · the mathematical models for the explicitly considered release and transport processes, as well as for the radiation exposure pathways that are included, · the implementation of the mathematical models in numerical codes, including an overview of these codes and the most relevant verification steps, · the general modelling approach when using the codes, in particular the generic assumptions needed to model the near field and the geosphere, along with some numerical details, · a description of the work flow related to the execution of the calculations and of the software tools that are used to facilitate the modelling process, and · an overview of the

  12. RECOMMENDED TRITIUM OXIDE DEPOSITION VELOCITY FOR USE IN SAVANNAH RIVER SITE SAFETY ANALYSES

    Energy Technology Data Exchange (ETDEWEB)

    Lee, P.; Murphy, C.; Viner, B.; Hunter, C.; Jannik, T.

    2012-04-03

    The Defense Nuclear Facilities Safety Board (DNFSB) has recently questioned the appropriate value for tritium deposition velocity used in the MELCOR Accident Consequence Code System Ver. 2 (Chanin and Young 1998) code when estimating bounding dose (95th percentile) for safety analysis (DNFSB 2011). The purpose of this paper is to provide appropriate, defensible values of the tritium deposition velocity for use in Savannah River Site (SRS) safety analyses. To accomplish this, consideration must be given to the re-emission of tritium after deposition. Approximately 85% of the surface area of the SRS is forested. The majority of the forests are pine plantations, 68%. The remaining forest area is 6% mixed pine and hardwood and 26% swamp hardwood. Most of the path from potential release points to the site boundary is through forested land. A search of published studies indicate daylight, tritiated water (HTO) vapor deposition velocities in forest vegetation can range from 0.07 to 2.8 cm/s. Analysis of the results of studies done on an SRS pine plantation and climatological data from the SRS meteorological network indicate that the average deposition velocity during daylight periods is around 0.42 cm/s. The minimum deposition velocity was determined to be about 0.1 cm/s, which is the recommended bounding value. Deposition velocity and residence time (half-life) of HTO in vegetation are related by the leaf area and leaf water volume in the forest. For the characteristics of the pine plantation at SRS the residence time corresponding to the average, daylight deposition velocity is 0.4 hours. The residence time corresponding to the night-time deposition velocity of 0.1 cm/s is around 2 hours. A simple dispersion model which accounts for deposition and re-emission of HTO vapor was used to evaluate the impact on exposure to the maximally exposed offsite individual (MOI) at the SRS boundary (Viner 2012). Under conditions that produce the bounding, 95th percentile MOI exposure

  13. The development of technologies of safety analysis for LMR ('03)

    International Nuclear Information System (INIS)

    Lee, Y. B.; Suk, S. D.; Chang, W. P.; Kwon, Y. M.; Jeong, H. Y.; Ha, K. W.; Heo, S.

    2004-03-01

    The developmental objectives of the project, 'The development of safety analysis techniques in LMR', are the code development for the subchannel blockage analysis, the code development for the system transient analysis, the code development for the HCDA(Hypothetical Core Disruptive Accident) analysis, the preliminary safety analysis for KALIMER-600 equipped with the components of new concepts, and the establishment of data base. The purpose of the analysis for subchannel blockage in the subassembly of LMR is to represent quantitatively that the maximum damage due to the accident is within the safety criteria. The computational program should be developed to simulate the thermal hydraulic phenomena and to verify the safety of LMR for the accident. For the purpose, the hybrid scheme has been implemented into the MATRA-LMR code based on the upwind scheme to analyze the various flow fields occurred in the subchannel blockage accident. The turbulent mixing models using the CFX code were assessed to compute more precisely the heat transfer between subchannels. Through this assessment, empirical correction factors of 1.7 for the heat conduction, 0.006 for the turbulent mixing coefficient were obtained. The distributed resistance model instead of wire forcing function has been developed to represent the more exact flow field due to wire-wrap. Other models, such as heat conductor model and various turbulent mixing model, have been implemented into the MATRA-LMR. The ORNL THORS 19-Pin FFM-5B tests have been assessed to validate above new models using the improved MATRA-LMR. The results using MATRA-LMR were well agreed with the experimental data. The subchannel blockage accidents which assumed to be occurred at the three locations for the conceptual plant of KALIMER-600 have been analysed according to blockage size using the MATRA-LMR code. The results of calculations for the design basis events which 6 subchannels were blocked showed the margins of the 290 7.dog. C up to the

  14. Establishment of safety analysis system for emergency operating procedures of Kori 3 and 4 and YGN 1 and 2 and reference analyses

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Won Jae; Kim, Hee Cheol; Ha, Kwi Seok; Chung, Bub Dong; Jeong, Jae Jun

    1999-02-01

    This report describes the establishment of safety analysis system for emergency operating procedures(EOPs) of Kori 3 and 4 and YGN 1 and 2 and the results of reference analyses. MARS1.3.1 code has been selected as a realistic system analysis tool of the safety analysis system, and the reactor system has been modeled using a fine nodding scheme in order to capture the major thermal-hydraulic phenomena that might occur during the transients. Full power steady state operating conditions are generated based on the plant operation data. Then, the reference analyses have been carried out for the accidents that can represent the typical EOPtransients, that is, a small-break loss-of-coolant-accident,a main steam line break accident and a steam generator tube rupture accident from the full power operation. For the realistic simulation of plant transient responses, reactor control and protection systems and safety systems are modeled based on their realistic performances. Also, the operator actions are modeled based on the current EOP actions. Through the reference analyses, the soundness of the established safety analysis system and the system modeling has been verified and the effectiveness of the current EOP has been partly justified. In conclusion, the safety analysis system established through this study can be used for the generation of technical background in the development and improvement of EOP actions and in the operator training. (Author). 11 refs., 6 tabs., 48 figs.

  15. Developing a generic environmental safety case

    International Nuclear Information System (INIS)

    Bailey, Lucy

    2014-01-01

    The Nuclear Decommissioning Authority (NDA) has been charged with implementing the United Kingdom government's policy for the long-term management of higher activity radioactive waste by planning, building and operating a geological disposal facility (GDF). Within the NDA, we - the Radioactive Waste Management Directorate (RWMD) - are tasked with the development of a GDF. The UK government has also decided that a process of voluntarism and partnership will be followed to identify a suitable site for the GDF. To date there is no volunteer community and the site selection process to find a volunteer host community is under review. RWMD has an ongoing role to provide advice to UK radioactive waste producers on the conditioning and packaging of wastes and to undertake disposability assessments of waste packaging proposals to determine their suitability for eventual disposal in a GDF. We also need to demonstrate our confidence that a GDF would be safe. Therefore RWMD has published a generic Environmental Safety Case (ESC) (NDA, 2010) to demonstrate that we are confident that a GDF could be developed to meet the guidelines set down by the environmental regulators (EA/NIEA, 2009) in a range of geological settings. The ESC includes reference case calculations that are used as a benchmark for disposability assessments. (author)

  16. Conclusions and Recommendations of the IAEA International Conference on Topical Issues in Nuclear Safety: Ensuring Safety for Sustainable Nuclear Development

    International Nuclear Information System (INIS)

    El-Shanawany, Mamdouh

    2011-01-01

    Over 200 participants from 33 countries and three international organizations came and actively participated and contributed to focused discussions and the success of the conference. The following points summarize the key conclusions and recommendations of the conference with respect to nuclear safety. 1. The nuclear safety approach is based on the philosophy developed in the 60's: defense in depth principles and deterministic criteria. When properly applied and completed by probabilistic analyses and operational experience feedback, it continues to be a successful approach. However, guarding against the risk of accidents requires constant vigilance and high technical competence and a never ending fight against complacency. In this context, having a strong leadership with a commitment to continuous improvement and a vision of sustained excellence is a key element of nuclear safety. Continuous improvement in safety also should be pursued through scientific research and operational experience feedback. 2. An accident anywhere is of concern to all Member States. Therefore, it is in the interest of all Member States to share and collaborate on safety matters. Participation of all Member States in international nuclear safety instruments and conventions, including liability for nuclear damage, is considered beneficial to global safety. The Convention on Nuclear Safety, the Joint Convention, international cooperation through IAEA and other organizations, bilateral or multilateral arrangements are important elements for establishing networks for sharing and transferring knowledge. It is acknowledged that the IAEA's Safety Fundamentals and Safety Requirements provide a sound foundation for high level nuclear safety. IAEA Safety Standards should be the basis for the establishment and maintenance of safety infrastructure. The IAEA's peer reviews and services such as IRRS, OSART, Site Evaluation and Reactor Safety Reviews provide also a valuable platform for sharing

  17. Main Conclusions and Recommendations of International Conference on Topical Issues in Nuclear Installation Safety: Ensuring Safety for Sustainable Nuclear Development

    International Nuclear Information System (INIS)

    El-Shanawany, Mamdouh

    2011-01-01

    Over 200 participants from 33 countries and three international organizations came and actively participated and contributed to focused discussions and the success of the conference. The following points summarize the key conclusions and recommendations of the conference with respect to nuclear safety. 1. The nuclear safety approach is based on the philosophy developed in the 60's: defense in depth principles and deterministic criteria. When properly applied and completed by probabilistic analyses and operational experience feedback, it continues to be a successful approach. However, guarding against the risk of accidents requires constant vigilance and high technical competence and a never ending fight against complacency. In this context, having a strong leadership with a commitment to continuous improvement and a vision of sustained excellence is a key element of nuclear safety. Continuous improvement in safety also should be pursued through scientific research and operational experience feedback. 2. An accident anywhere is of concern to all Member States. Therefore, it is in the interest of all Member States to share and collaborate on safety matters. Participation of all Member States in international nuclear safety instruments and conventions, including liability for nuclear damage, is considered beneficial to global safety. The Convention on Nuclear Safety, the Joint Convention, international cooperation through IAEA and other organizations, bilateral or multilateral arrangements are important elements for establishing networks for sharing and transferring knowledge. It is acknowledged that the IAEA's Safety Fundamentals and Safety Requirements provide a sound foundation for high level nuclear safety. IAEA Safety Standards should be the basis for the establishment and maintenance of safety infrastructure. The IAEA's peer reviews and services such as IRRS, OSART, Site Evaluation and Reactor Safety Reviews provide also a valuable platform for sharing

  18. Economic Developments on Perceived Safety, Violence, and Economic Benefits

    Directory of Open Access Journals (Sweden)

    Anthony Fabio

    2015-01-01

    Full Text Available Background. Emerging research highlights the promise of community- and policy-level strategies in preventing youth violence. Large-scale economic developments, such as sports and entertainment arenas and casinos, may improve the living conditions, economics, public health, and overall wellbeing of area residents and may influence rates of violence within communities. Objective. To assess the effect of community economic development efforts on neighborhood residents’ perceptions on violence, safety, and economic benefits. Methods. Telephone survey in 2011 using a listed sample of randomly selected numbers in six Pittsburgh neighborhoods. Descriptive analyses examined measures of perceived violence and safety and economic benefit. Responses were compared across neighborhoods using chi-square tests for multiple comparisons. Survey results were compared to census and police data. Results. Residents in neighborhoods with the large-scale economic developments reported more casino-specific and arena-specific economic benefits. However, 42% of participants in the neighborhood with the entertainment arena felt there was an increase in crime, and 29% of respondents from the neighborhood with the casino felt there was an increase. In contrast, crime decreased in both neighborhoods. Conclusions. Large-scale economic developments have a direct influence on the perception of violence, despite actual violence rates.

  19. Experiment-specific analyses in support of code development

    International Nuclear Information System (INIS)

    Ott, L.J.

    1990-01-01

    Experiment-specific models have been developed since 1986 by Oak Ridge National Laboratory Boiling Water Reactor (BWR) severe accident analysis programs for the purpose of BWR experimental planning and optimum interpretation of experimental results. These experiment-specific models have been applied to large integral tests (ergo, experiments) which start from an initial undamaged core state. The tests performed to date in BWR geometry have had significantly different-from-prototypic boundary and experimental conditions because of either normal facility limitations or specific experimental constraints. These experiments (ACRR: DF-4, NRU: FLHT-6, and CORA) were designed to obtain specific phenomenological information such as the degradation and interaction of prototypic components and the effects on melt progression of control-blade materials and channel boxes. Applications of ORNL models specific to the ACRR DF-4 and KfK CORA-16 experiments are discussed and significant findings from the experimental analyses are presented. 32 refs., 16 figs

  20. Measuring patient safety in a UK dental hospital: development of a dental clinical effectiveness dashboard.

    Science.gov (United States)

    Pemberton, M N; Ashley, M P; Shaw, A; Dickson, S; Saksena, A

    2014-10-01

    Patient safety is an important marker of quality for any healthcare organisation. In 2008, the British Government white paper entitled High quality care for all, resulting from a review led by Lord Darzi, identified patient safety as a key component of quality and discussed how it might be measured, analysed and acted upon. National and local clinically curated metrics were suggested, which could be displayed via a 'clinical dashboard'. This paper explains the development of a clinical effectiveness dashboard focused on patient safety in an English dental hospital and how it has helped us identify relevant patient safety issues in secondary dental care.

  1. Development and measurement of perioperative patient safety indicators

    NARCIS (Netherlands)

    Emond, Y. E.; Stienen, J. J.; Wollersheim, H. C.; Bloo, G. J.; Damen, J.; Westert, G. P.; Boermeester, M. A.; Pols, M. A.; Calsbeek, H.; Wolff, A. P.

    2015-01-01

    To improve perioperative patient safety, hospitals are implementing evidence-based perioperative safety guidelines. To facilitate this process, it is important to provide insight into current practice. For this purpose, we aimed to develop patient safety indicators. The RAND-modified Delphi method

  2. Utilisation of best estimate system codes and best estimate methods in safety analyses of VVER reactors in the Czech Republic

    International Nuclear Information System (INIS)

    Macek, Jiri; Kral, Pavel

    2010-01-01

    The content of the presentation was as follows: Conservative versus best estimate approach, Brief description and selection of methodology, Description of uncertainty methods, Examples of the BE methodology. It is concluded that where BE computer codes are used, uncertainty and sensitivity analyses should be included; if best estimate codes + uncertainty are used, the safety margins increase; and BE + BSA is the next step in licensing analyses. (P.A.)

  3. Development and Validation of Career Development Guidelines by Task/Activity Analysis of Occupational Safety and Health Professions: Industrial Hygiene and Safety Professional. Final Report. Technical Report XII.

    Science.gov (United States)

    Vernon, Ralph J.; And Others

    This report summarizes research findings which resulted in development of curricula for occupational safety and health professions based on task/activity analyses and related performance objectives. The first seven chapters focus on the seven objectives. Chapter 1, Literature Review and Selection of Employers, concerns tasks required for…

  4. Development and applications of a safety assessment system for promoting safety culture in nuclear power plants

    International Nuclear Information System (INIS)

    Takano, Ken-ichi; Hasegawa, Naoko; Hirose, Ayako; Hayase, Ken-ichi

    2004-01-01

    For past five years, CRIEPI has been continuing efforts to develop and make applications of a 'safety assessment system' which enable to measure the safety level of organization. This report describe about frame of the system, assessment results and its reliability, and relation between labor accident rate in the site and total safety index (TSI), which can be obtained by the principal factors analysis. The safety assessment in this report is based on questionnaire survey of employee. The format and concrete questionnaires were developed using existing literatures including organizational assessment tools. The tailored questionnaire format involved 124 questionnaire items. The assessment results could be considered as a well indicator of the safety level of organization, safety management, and safety awareness of employee. (author)

  5. Developing an integrated dam safety program

    International Nuclear Information System (INIS)

    Nielsen, N. M.; Lampa, J.

    1996-01-01

    An effort has been made to demonstrate that dam safety is an integral part of asset management which, when properly done, ensures that all objectives relating to safety and compliance, profitability, stakeholders' expectations and customer satisfaction, are achieved. The means to achieving this integration of the dam safety program and the level of effort required for each core function have been identified using the risk management approach to pinpoint vulnerabilities, and subsequently to focus priorities. The process is considered appropriate for any combination of numbers, sizes and uses of dams, and is designed to prevent exposure to unacceptable risks. 5 refs., 1 tab

  6. Results of the safety analyses for the Greifswald and Stendal WWER nuclear power plants

    International Nuclear Information System (INIS)

    Milhem, J.L.

    1993-03-01

    Following a brief introduction of the design features of the three types of the WWER reactors, the paper deals with the main issues of the safety-related design and the most important recommendations which have been derived for upgrading measures. Furthermore some operational safety aspects of the VVER-1000 will be discussed in some detail

  7. How can BIM support Construction Safety Management? Development of SIM

    Directory of Open Access Journals (Sweden)

    Chan Isabelle Y S

    2016-01-01

    Full Text Available BIM is continuously changing the way construction can be approached, from design, scheduling, costing to facilities management. However, even though construction safety is a worldwide issue, and though previous, yet limited, studies have proven its great potential, BIM has not been extended to include automated construction safety management functions, i.e., hazard identification, assessment and control for construction works. In fact, as many as 71% of safety incidents can be prevented by safety considerations at design stage. Automation via BIM is the key to enhance effectiveness and efficiency of safety management, but development of Safety Information Modelling (SIM is the key pre-requisite. This paper aims to outline a research project for developing both information and technical requirements for SIM via questionnaire survey, focus group, and real case studies. The project results, to be validated by real projects, will act as solid platform for developing:- i Practical guidance to construction and safety professionals on the types of safety management a SIM can conduct, and the ways of utilizing the safety information in a SIM model [based on the systematic SIM information requirements (e.g., object identification, geometry attributes, safety rules]; and ii Overall guidance to software developers to develop appropriate SIM tools for different scenarios [based on the systematic SIM technical requirements (e.g., model information exchange, visualization, conflict analysis process for risk assessment, reliability].

  8. Histological and chemical analyses of mesiodens development and mineralization.

    Science.gov (United States)

    Muhamad, Awady; Moti, Moskovitz; Ornit, Cohen; Uri, Zilberman

    2018-03-01

    This study aimed to compare the developmental timing and mineralization quality of mesiodentes, i.e., supernumerary teeth located mainly in the midline of the maxilla between the central incisors, with the developmental timing and mineralization quality of permanent and primary central incisors. Sixteen mesiodentes, nine permanent and seven primary central incisors were collected. The location of the neonatal line was determined using a light microscope at 10× or 20× enlargements. Chemical composition of the enamel at two locations was analyzed using energy dispersive X-ray spectrometer. Neonatal lines were observed in eight out of 16 mesiodentes, in all primary central incisors and in none of the permanent central incisors. Chemical analyses showed that mesiodentes mineralization was impaired, resulting in higher amount of organic ions and reduced inorganic ions. Discriminant analysis showed minimal overlap of mesiodentes with either primary or permanent centrals. Mesiodentes development begins before birth in 50% of the cases but later than the primary centrals. Mineralization of mesiodens is impaired with less mineral content and higher organic content. The results showed that mesiodentes are a special group of teeth with defective morpho-differentiation and mineralization, with little similarity to primary or permanent central incisors. Copyright © 2017 Elsevier Ltd. All rights reserved.

  9. Development of safety performance indicators in Japan

    International Nuclear Information System (INIS)

    Ohashi, H.; Tamao, S.; Tanaka, J.; Sawayama, T.

    2001-01-01

    For the purpose of safety regulations of operating nuclear power stations in Japan, the regulatory authorities utilize two types of regulations. One is the direct regulation, such as periodical inspection to inspect the function and performance of equipment important to safety, and the other is the audit type regulation such as preservation inspection to audit the compliance with the safety preservation rules. As performance indicators are expected to be an effective tool to evaluate the activities by audit type regulations, NUPEC is studying a comprehensive set of operational performance indicators to meet the effective evaluation method for the safety preservation activities in the audit type regulations under the frame of current safety regulation system. The study includes the establishment of comprehensive operational performance indicators applicable in Japan, the effective application of performance indicators to the current Japanese regulation, the clarification of the applicable scope of utilization, the possibility of applying the performance indicators. This report describes the present status of our performance indicator studies. After the completion of these studies the regulatory authorities will evaluate if and how the new set of comprehensive performance indicators could be introduced to Japanese regulatory scheme. (author)

  10. Development of digital image correlation method to analyse crack ...

    Indian Academy of Sciences (India)

    The crack formation tendencies on masonry wall can be observed at an earlier stage by this proposed method. These results show a great application potential of the DIC technique for various situations such as inspecting shrinkage-induced cracks in fresh concrete, masonry and reinforced concrete structures, and safety of ...

  11. RAMONA-4B development for SBWR safety studies

    Energy Technology Data Exchange (ETDEWEB)

    Rohatgi, U.S.; Aronson, A.L.; Cheng, H.S.; Khan, H.J.; Mallen, A.N.

    1993-12-31

    The Simplified Boiling Water Reactor (SBWR) is a revolutionary design of a boiling-water reactor. The reactor is based on passive safety systems such as natural circulation, gravity flow, pressurized gas, and condensation. SBWR has no active systems, and the flow in the vessel is by natural circulation. There is a large chimney section above the core to provide a buoyancy head for natural circulation. The reactor can be shut down by either of four systems; namely, scram, Fine Motion Control Rod Drive (FMCRD), Alternate Rod Insertion (ARI), and Standby Liquid Control System (SLCS). The safety injection is by gravity drain from the Gravity Driven Cooling System (GDCS) and Suppression Pool (SP). The heat sink is through two types of heat exchangers submerged in the tank of water. These heat exchangers are the Isolation Condenser (IC) and the Passive Containment Cooling System (PCCS). The RAMONA-4B code has been developed to simulate the normal operation, reactivity transients, and to address the instability issues for SBWR. The code has a three-dimensional neutron kinetics coupled to multiple parallel-channel thermal-hydraulics. The two-phase thermal hydraulics is based on a nonhomogeneous nonequilibrium drift-flux formulation. It employs an explicit integration to solve all state equations (except for neutron kinetics) in order to predict the instability without numerical damping. The objective of this project is to develop a Sun SPARC and IBM RISC 6000 based RAMONA-4B code for applications to SBWR safety analyses, in particular for stability and ATWS studies.

  12. Highway safety design workshops.

    Science.gov (United States)

    2010-11-01

    Highway safety is an ongoing concern for the Texas Department of Transportation (TxDOT). As part of its : proactive commitment to improving highway safety, TxDOT is moving toward including quantitative safety : analyses earlier in the project develop...

  13. Assessing progress in the development of safety culture

    International Nuclear Information System (INIS)

    Rotaru, I.; Ghita, S.; Biro, L.

    2002-01-01

    This paper is focussed on the organizational culture and learning processes required for the implementation of all aspects of safety culture. There is no prescriptive formula for improving safety culture. However, some common characteristics and practices are emerging that can be adopted by organizations in order to make progress. The paper refers to some approaches that have been successful in a number of countries. The experience of the international nuclear industry in the development and improvement of safety culture could be extended and found useful in other nuclear activities, irrespective of scale. The examples given of specific practice cover a wide range of activities including analysis of events, the regulatory approach on safety culture, employee participation and safety performance measures. Many of these practices may be relevant to smaller organizations and could contribute to improving safety culture, whatever the size of the organization. The most effective approach is to pursue a range of practices that can be mutually supportive in the development of a progressive safety culture, supported by professional standards, organizational and management commitment. Some guidance is also given on the assessment of safety culture and on the detection of a weakening safety culture. Few suggestions for accelerating the safety culture development and improvement process are also provided. (author)

  14. Accomplishment of 10-year research in NUCEF and future development. Criticality safety research

    International Nuclear Information System (INIS)

    Miyoshi, Yoshinori

    2005-01-01

    Since 1995, static and transient critical experiments on low enriched uranyl nitrate solution have been performed using two solution type criticality facilities, STACY and TRACY constructed in NUCEF. The obtained fundamental and systematic data on aqueous solution were used to validate the criticality safety calculation codes and to develop the transient analyses codes for criticality accident evaluation. This paper describes the outline of the criticality safety research conducted in NUCEF. (author)

  15. Patient Safety in Pediatrics: a Developing Discipline

    NARCIS (Netherlands)

    C. van der Starre (Cynthia)

    2011-01-01

    markdownabstract__Abstract__ The publication of the breakthrough report “To Err is Human” by the Institute of Medicine was the launch of patient safety initiatives all over the world. In the intensive care unit (ICU) of the Erasmus MC-Sophia Children’s Hospital this resulted in the institution

  16. Development of safety goals for nuclear power plants

    International Nuclear Information System (INIS)

    Rathbun, D.; Modarres, M.

    1987-01-01

    This article presents an overview of the historical development of safety goals for nuclear power plants, including the July 30, 1986, safety goal policy. In the course of the Nuclear Regulatory Commission's evaluation of safety goals, several issues have emerged. In particular, the benefit-cost guideline and the core-melt design objective have been subjects of continuing scrutiny and debate. These and other issues remaining to be resolved for completing the development of a final policy are discussed. Finally, the current status of the safety goals policy is discussed

  17. Development of Safety Review Guidance for Research and Training Reactors

    International Nuclear Information System (INIS)

    Oh, Kju-Myeng; Shin, Dae-Soo; Ahn, Sang-Kyu; Lee, Hoon-Joo

    2007-01-01

    The KINS already issued the safety review guidance for pressurized LWRs. But the safety review guidance for research and training reactors were not developed. So, the technical standard including safety review guidance for domestic research and training reactors has been applied mutates mutandis to those of nuclear power plants. It is often difficult for the staff to effectively perform the safety review of applications for the permit by the licensee, based on peculiar safety review guidance. The NRC and NSC provide the safety review guidance for test and research reactors and European countries refer to IAEA safety requirements and guides. The safety review guide (SRG) of research and training reactors was developed considering descriptions of the NUREG- 1537 Part 2, previous experiences of safety review and domestic regulations for related facilities. This study provided the safety review guidance for research and training reactors and surveyed the difference of major acceptance criteria or characteristics between the SRG of pressurized light water reactor and research and training reactors

  18. Biosphere analyses for the safety assessment SR-Site - synthesis and summary of results

    International Nuclear Information System (INIS)

    Saetre, Peter

    2010-12-01

    This report summarises nearly 20 biosphere reports and gives a synthesis of the work performed within the SR-Site Biosphere project, i.e. the biosphere part of SR-Site. SR-Site Biosphere provides the main project with dose conversion factors (LDFs), given a unit release rate, for calculation of human doses under different release scenarios, and assesses if a potential release from the repository would have detrimental effects on the environment. The intention of this report is to give sufficient details for an overview of methods, results and major conclusions, with references to the biosphere reports where methods, data and results are presented and discussed in detail. The philosophy of the biosphere assessment was to make estimations of the radiological risk for humans and the environment as realistic as possible, based on the knowledge of present-day conditions at Forsmark and the past and expected future development of the site. This was achieved by using the best available knowledge, understanding and data from extensive site investigations from two sites. When sufficient information was not available, uncertainties were handled cautiously. A systematic identification and evaluation of features and processes that affect transport and accumulation of radionuclides at the site was conducted, and the results were summarised in an interaction matrix. Data and understanding from the site investigation was an integral part of this work, the interaction matrix underpinned the development of the radionuclide model used in the biosphere assessment. Understanding of the marine, lake and river and terrestrial ecosystems at the site was summarized in a conceptual model, and relevant features and process have been characterized to capture site specific parameter values. Detailed investigations of the structure and history of the regolith at the site and simulations of regolith dynamics were used to describe the present day state at Forsmark and the expected development of

  19. Development of Safety Kit for Industrial Radiography Application

    International Nuclear Information System (INIS)

    Mohd Noorul Ikhsan Ahmad; Amry Amin Abas

    2011-01-01

    A safety kit for industrial radiography has been developed. The safety kit that consist of a set of technical rod and various size of base that can be used in radiograph of pipe with diameter between half and one and half inch with utilization of collimator. With the kit, radiographers will not having anymore problem to use collimator in their work. The paper discuss about the technical measures of the safety kit and the importance of introducing it to the industry. (author)

  20. Development of an environmental safety case guidance manual

    International Nuclear Information System (INIS)

    Wellstead, Matthew John

    2014-01-01

    NDA RWMD is currently considering the scope, purpose and structure of a safety case manual that covers the development of nuclear operational, transport and environmental safety cases for a geological disposal facility in the United Kingdom. This paper considers the Environmental Safety Case (ESC) input into such a manual (herein referred to as the 'ESC Manual'), looking at the drivers and benefits that a guidance manual in this area may provide. (authors)

  1. Time Headway Modelling of Motorcycle-Dominated Traffic to Analyse Traffic Safety Performance and Road Link Capacity of Single Carriageways

    Directory of Open Access Journals (Sweden)

    D. M. Priyantha Wedagama

    2017-04-01

    Full Text Available This study aims to develop time headway distribution models to analyse traffic safety performance and road link capacities for motorcycle-dominated traffic in Denpasar, Bali. Three road links selected as the case study are Jl. Hayam Wuruk, Jl.Hang Tuah, and Jl. Padma. Data analysis showed that between 55%-80% of motorists in Denpasar during morning and evening peak hours paid less attention to the safe distance with the vehicles in front. The study found that Lognormal distribution models are best to fit time headway data during morning peak hours while either Weibull (3P or Pearson III distributions is for evening peak hours. Road link capacities for mixed traffic predominantly motorcycles are apparently affected by the behaviour of motorists in keeping safe distance with the vehicles in front. Theoretical road link capacities for Jl. Hayam Wuruk, Jl. Hang Tuah and Jl. Padma are 3,186 vehicles/hour, 3,077 vehicles/hour and 1935 vehicles/hour respectively.

  2. Criticality safety and sensitivity analyses of PWR spent nuclear fuel repository facilities

    NARCIS (Netherlands)

    Maucec, M; Glumac, B

    Monte Carlo criticality safety and sensitivity calculations of pressurized water reactor (PWR) spent nuclear fuel repository facilities for the Slovenian nuclear power plant Krsko are presented. The MCNP4C code was deployed to model and assess the neutron multiplication parameters of pool-based

  3. EULAR points to consider when establishing, analysing and reporting safety data of biologics registers in rheumatology

    DEFF Research Database (Denmark)

    Dixon, William G; Carmona, Loreto; Finckh, Axel

    2010-01-01

    The introduction of biological therapies for the treatment of rheumatic diseases has drawn attention to the limitations of traditional means of assessing drug safety. Consequently, a series of European academic biologics registers dedicated to this task have been established. Increasing reliance...

  4. Guidelines for nuclear power plant safety issue prioritization information development. Supplement 2

    Energy Technology Data Exchange (ETDEWEB)

    Andrews, W.B.; Gallucci, R.H.V.; Konzek, G.J.; Heaberlin, S.W.; Fecht, B.A.; Allen, C.H.; Allen, R.D.; Bickford, W.E., Carbaugh, E.H.; Lewis, J.R.

    1983-12-01

    This is the third in a series of reports to document the use of a methodology developed by the Pacific Northwest Laboratory to calculate, for prioritization purposes, the risk, dose and cost impacts of implementing resolutions to reactor safety issues (NUREG/CR-2800, Andrews et al. 1983). This report contains results of issue-specific analyses for 31 issues. Each issue was considered within the constraints of available information as of summer 1983, and two staff-weeks of labor. The results are referenced, as one consideration in setting priorities for reactor safety issues, in NUREG-0933, A Prioritization of Generic Safety Issues.

  5. Guidelines for nuclear power plant safety issue prioritization information development. Supplement 2

    International Nuclear Information System (INIS)

    Andrews, W.B.; Gallucci, R.H.V.; Konzek, G.J.; Heaberlin, S.W.; Fecht, B.A.; Allen, C.H.; Allen, R.D.; Bickford, W.E.; Carbaugh, E.H.; Lewis, J.R.

    1983-12-01

    This is the third in a series of reports to document the use of a methodology developed by the Pacific Northwest Laboratory to calculate, for prioritization purposes, the risk, dose and cost impacts of implementing resolutions to reactor safety issues (NUREG/CR-2800, Andrews et al. 1983). This report contains results of issue-specific analyses for 31 issues. Each issue was considered within the constraints of available information as of summer 1983, and two staff-weeks of labor. The results are referenced, as one consideration in setting priorities for reactor safety issues, in NUREG-0933, A Prioritization of Generic Safety Issues

  6. Guidelines for nuclear power plant safety issue prioritization information development. Supplement 4

    Energy Technology Data Exchange (ETDEWEB)

    Tabatabai, A.S.; Fecht, B.A.; Powers, T.B.; Bickford, W.E.; Andrews, W.B.; Gallucci, R.H.V.; Bian, S.H.; Daling, P.M.; Eschbach, E.J.; Allen, C.H.

    1986-07-01

    This is the fifth in a series of reports to document the use of a methodology developed by the Pacific Northwest Laboratory to calculate, for prioritization purposes, the risk, dose and cost impacts of implementing resolutions to reactor safety issues (NUREG/CR-2800, Andrews et al. 1983). This report contains results of issue-specific analyses for 23 issues. Each issue was considered within the constraints of available information as of winter 1986, and two staff-weeks of labor. The results are referenced, as one consideration in setting priorities for reactor safety issues, in NUREG-0933, ''A Prioritization of Generic Safety Issues.''

  7. Development of a Safety Management Web Tool for Horse Stables.

    Science.gov (United States)

    Leppälä, Jarkko; Kolstrup, Christina Lunner; Pinzke, Stefan; Rautiainen, Risto; Saastamoinen, Markku; Särkijärvi, Susanna

    2015-11-12

    Managing a horse stable involves risks, which can have serious consequences for the stable, employees, clients, visitors and horses. Existing industrial or farm production risk management tools are not directly applicable to horse stables and they need to be adapted for use by managers of different types of stables. As a part of the InnoEquine project, an innovative web tool, InnoHorse, was developed to support horse stable managers in business, safety, pasture and manure management. A literature review, empirical horse stable case studies, expert panel workshops and stakeholder interviews were carried out to support the design. The InnoHorse web tool includes a safety section containing a horse stable safety map, stable safety checklists, and examples of good practices in stable safety, horse handling and rescue planning. This new horse stable safety management tool can also help in organizing work processes in horse stables in general.

  8. Cyber safety education in developing countries

    CSIR Research Space (South Africa)

    Von Solms, R

    2015-07-01

    Full Text Available on the potential risks that are faced when using Internet communication tools, such as the social media, chat rooms, online gaming, email and instant messaging [1],[2] When users are educated about these risks, the users' vulnerability to malware attacks, cyber... Union's initiatives include ins@fe [5], Safer Social Networking Principles for the EU [6] and European Framework for Safer Mobile Use by Younger Teenagers and Children [7]. As early as 2009, education on Online Safety was included in school curricula...

  9. Towards a Road Safety Development Index (RSDI) : Development of an International Index to Measure Road Safety Performance

    OpenAIRE

    Al Haji, Ghazwan

    2005-01-01

    Aim. This study suggests a set of methodologies to combine different indicators of road safety into a single index. The RSDI is a simple and quick composite index, which may become a significant measurement in comparing, ranking and determining road safety levels in different countries and regions worldwide. Design. One particular concern in designing a Road Safety Development Index (RSDI) is to come up with a comprehensive set of exposure and risk indicators which includes as far as possible...

  10. Development and implementation of setpoint tolerances for special safety systems

    International Nuclear Information System (INIS)

    Oliva, A.F.; Balog, G.; Parkinson, D.G.; Archinoff, G.H.

    1991-01-01

    The establishment of tolerances and impairment limits for special safety system setpoints is part of the process whereby the plant operator demonstrates to the regulatory authority that the plant operates safely and within the defined plant licensing envelope. The licensing envelope represents the set of limits and plant operating state and for which acceptably safe plant operation has been demonstrated by the safety analysis. By definition, operation beyond this envelope contributes to overall safety system unavailability. Definition of the licensing envelope is provided in a wide range of documents including the plant operating licence, the safety report, and the plant operating policies and principles documents. As part of the safety analysis, limits are derived for each special safety system initiating parameter such that the relevant safety design objectives are achieved for all design basis events. If initiation on a given parameter occurs at a level beyond its limit, there is a potential reduction in safety system effectiveness relative to the performance credited in the plant safety analysis. These safety system parameter limits, when corrected for random and systematic instrument errors and other errors inherent in the process of periodic testing or calibration, are then used to derive parameter impairment levels and setpoint tolerances. This paper describes the methodology that has evolved at Ontario Hydro for developing and implementing tolerances for special safety system parameters (i.e., the shutdown systems, emergency coolant injection system and containment system). Tolerances for special safety system initiation setpoints are addressed specifically, although many of the considerations discussed here will apply to performance limits for other safety system components. The first part of the paper deals with the approach that has been adopted for defining and establishing setpoint limits and tolerances. The remainder of the paper addresses operational

  11. Patient involvement in patient safety: Protocol for developing an intervention using patient reports of organisational safety and patient incident reporting

    Science.gov (United States)

    2011-01-01

    Background Patients have the potential to provide a rich source of information on both organisational aspects of safety and patient safety incidents. This project aims to develop two patient safety interventions to promote organisational learning about safety - a patient measure of organisational safety (PMOS), and a patient incident reporting tool (PIRT) - to help the NHS prevent patient safety incidents by learning more about when and why they occur. Methods To develop the PMOS 1) literature will be reviewed to identify similar measures and key contributory factors to error; 2) four patient focus groups will ascertain practicality and feasibility; 3) 25 patient interviews will elicit approximately 60 items across 10 domains; 4) 10 patient and clinician interviews will test acceptability and understanding. Qualitative data will be analysed using thematic content analysis. To develop the PIRT 1) individual and then combined patient and clinician focus groups will provide guidance for the development of three potential reporting tools; 2) nine wards across three hospital directorates will pilot each of the tools for three months. The best performing tool will be identified from the frequency, volume and quality of reports. The validity of both measures will be tested. 300 patients will be asked to complete the PMOS and PIRT during their stay in hospital. A sub-sample (N = 50) will complete the PMOS again one week later. Health professionals in participating wards will also be asked to complete the AHRQ safety culture questionnaire. Case notes for all patients will be reviewed. The psychometric properties of the PMOS will be assessed and a final valid and reliable version developed. Concurrent validity for the PIRT will be assessed by comparing reported incidents with those identified from case note review and the existing staff reporting scheme. In a subsequent study these tools will be used to provide information to wards/units about their priorities for patient

  12. Patient involvement in patient safety: Protocol for developing an intervention using patient reports of organisational safety and patient incident reporting

    Directory of Open Access Journals (Sweden)

    Armitage Gerry

    2011-05-01

    Full Text Available Abstract Background Patients have the potential to provide a rich source of information on both organisational aspects of safety and patient safety incidents. This project aims to develop two patient safety interventions to promote organisational learning about safety - a patient measure of organisational safety (PMOS, and a patient incident reporting tool (PIRT - to help the NHS prevent patient safety incidents by learning more about when and why they occur. Methods To develop the PMOS 1 literature will be reviewed to identify similar measures and key contributory factors to error; 2 four patient focus groups will ascertain practicality and feasibility; 3 25 patient interviews will elicit approximately 60 items across 10 domains; 4 10 patient and clinician interviews will test acceptability and understanding. Qualitative data will be analysed using thematic content analysis. To develop the PIRT 1 individual and then combined patient and clinician focus groups will provide guidance for the development of three potential reporting tools; 2 nine wards across three hospital directorates will pilot each of the tools for three months. The best performing tool will be identified from the frequency, volume and quality of reports. The validity of both measures will be tested. 300 patients will be asked to complete the PMOS and PIRT during their stay in hospital. A sub-sample (N = 50 will complete the PMOS again one week later. Health professionals in participating wards will also be asked to complete the AHRQ safety culture questionnaire. Case notes for all patients will be reviewed. The psychometric properties of the PMOS will be assessed and a final valid and reliable version developed. Concurrent validity for the PIRT will be assessed by comparing reported incidents with those identified from case note review and the existing staff reporting scheme. In a subsequent study these tools will be used to provide information to wards/units about their

  13. Recommended Tritium Oxide Deposition Velocity For Use In Savannah River Site Safety Analyses

    Energy Technology Data Exchange (ETDEWEB)

    Lee, P. L.; Murphy, C. E.; Viner, B. J.; Hunter, C. H.

    2012-07-31

    This report documents the results of examining the deposition velocity of water to forests, the residence time of HTO in forests, and the relation between deposition velocity and residence time with specific consideration given to the topography and experimental work performed at SRS. A simple mechanistic model is used to obtain plausible deposition velocity and residence time values where experimental data are not available and recommendations are made for practical application in a safety analysis model.

  14. An approach of sensitivity and uncertainty analyses methods installation in a safety calculation

    International Nuclear Information System (INIS)

    Pepin, G.; Sallaberry, C.

    2003-01-01

    Simulation of the migration in deep geological formations leads to solve convection-diffusion equations in porous media, associated with the computation of hydrogeologic flow. Different time-scales (simulation during 1 million years), scales of space, contrasts of properties in the calculation domain, are taken into account. This document deals more particularly with uncertainties on the input data of the model. These uncertainties are taken into account in total analysis with the use of uncertainty and sensitivity analysis. ANDRA (French national agency for the management of radioactive wastes) carries out studies on the treatment of input data uncertainties and their propagation in the models of safety, in order to be able to quantify the influence of input data uncertainties of the models on the various indicators of safety selected. The step taken by ANDRA consists initially of 2 studies undertaken in parallel: - the first consists of an international review of the choices retained by ANDRA foreign counterparts to carry out their uncertainty and sensitivity analysis, - the second relates to a review of the various methods being able to be used in sensitivity and uncertainty analysis in the context of ANDRA's safety calculations. Then, these studies are supplemented by a comparison of the principal methods on a test case which gathers all the specific constraints (physical, numerical and data-processing) of the problem studied by ANDRA

  15. Biosphere analyses for the safety assessment SR-Site - synthesis and summary of results

    Energy Technology Data Exchange (ETDEWEB)

    Saetre, Peter (comp.)

    2010-12-15

    This report summarises nearly 20 biosphere reports and gives a synthesis of the work performed within the SR-Site Biosphere project, i.e. the biosphere part of SR-Site. SR-Site Biosphere provides the main project with dose conversion factors (LDFs), given a unit release rate, for calculation of human doses under different release scenarios, and assesses if a potential release from the repository would have detrimental effects on the environment. The intention of this report is to give sufficient details for an overview of methods, results and major conclusions, with references to the biosphere reports where methods, data and results are presented and discussed in detail. The philosophy of the biosphere assessment was to make estimations of the radiological risk for humans and the environment as realistic as possible, based on the knowledge of present-day conditions at Forsmark and the past and expected future development of the site. This was achieved by using the best available knowledge, understanding and data from extensive site investigations from two sites. When sufficient information was not available, uncertainties were handled cautiously. A systematic identification and evaluation of features and processes that affect transport and accumulation of radionuclides at the site was conducted, and the results were summarised in an interaction matrix. Data and understanding from the site investigation was an integral part of this work, the interaction matrix underpinned the development of the radionuclide model used in the biosphere assessment. Understanding of the marine, lake and river and terrestrial ecosystems at the site was summarized in a conceptual model, and relevant features and process have been characterized to capture site specific parameter values. Detailed investigations of the structure and history of the regolith at the site and simulations of regolith dynamics were used to describe the present day state at Forsmark and the expected development of

  16. Playing by the rules? Analysing incremental urban developments

    NARCIS (Netherlands)

    Karnenbeek, van Lilian; Janssen-Jansen, Leonie

    2018-01-01

    Current urban developments are often considered outdated and static, and the argument follows that they should become more adaptive. In this paper, we argue that existing urban development are already adaptive and incremental. Given this flexibility in urban development, understanding changes in the

  17. Nuclear criticality safety experiments, calculations, and analyses - 1958 to 1982. Volume 2. Summaries. Complilation of papers from the Transactions of the American Nuclear Society

    International Nuclear Information System (INIS)

    Koponen, B.L.; Hampel, V.E.

    1982-01-01

    This compilation contains 688 complete summaries of papers on nuclear criticality safety as presented at meetings of the American Nuclear Society (ANS). The selected papers contain criticality parameters for fissile materials derived from experiments and calculations, as well as criticality safety analyses for fissile material processing, transport, and storage. The compilation was developed as a component of the Nuclear Criticality Information System (NCIS) now under development at the Lawrence Livermore National Laboratory. The compilation is presented in two volumes: Volume 1 contains a directory to the ANS Transaction volume and page number where each summary was originally published, the author concordance, and the subject concordance derived from the keyphrases in titles. Volume 2 contains-in chronological order-the full-text summaries, reproduced here by permission of the American Nuclear Society from their Transactions, volumes 1-41

  18. Nuclear criticality safety experiments, calculations, and analyses - 1958 to 1982. Volume 2. Summaries. Complilation of papers from the Transactions of the American Nuclear Society

    Energy Technology Data Exchange (ETDEWEB)

    Koponen, B.L.; Hampel, V.E.

    1982-10-21

    This compilation contains 688 complete summaries of papers on nuclear criticality safety as presented at meetings of the American Nuclear Society (ANS). The selected papers contain criticality parameters for fissile materials derived from experiments and calculations, as well as criticality safety analyses for fissile material processing, transport, and storage. The compilation was developed as a component of the Nuclear Criticality Information System (NCIS) now under development at the Lawrence Livermore National Laboratory. The compilation is presented in two volumes: Volume 1 contains a directory to the ANS Transaction volume and page number where each summary was originally published, the author concordance, and the subject concordance derived from the keyphrases in titles. Volume 2 contains-in chronological order-the full-text summaries, reproduced here by permission of the American Nuclear Society from their Transactions, volumes 1-41.

  19. Reactor safety computer code development at INEL

    International Nuclear Information System (INIS)

    Johnsen, G.W.

    1985-01-01

    This report provides a brief overview of the computer code development programs being conducted at EG and G Idaho, Inc. on behalf of US Nuclear Regulatory Commission and the Department of Energy, Idaho Operations Office. Included are descriptions of the codes being developed, their development status as of the date of this report, and resident code development expertise

  20. Contamination, decontamination and radiochemical safety analyses of the RA reactor (Report 1966)

    International Nuclear Information System (INIS)

    Maksimovic, Z.

    1966-12-01

    This contract is concerned with development of methods for detection of fission products i the heavy water and quantitative radiochemical analysis for detecting one fission product which enables reliable verification of heavy water contamination by fission products and estimation of contamination level. Qualitative and quantitative radiometry measurements of fission products in water are shown on page 4. Page 6 shows study of contamination and decontamination of water on the laboratory level. Experiments have shown that the majority of fission products was adsorbed on the uranium oxide and that the iodine isotopes are partly in water (non-adsorbed). Gamma spectrometry analyses showed 131 I moves to distillate with the initial quantities of distilled water. decontamination factors compared to the total activity of fission products in distillator and distillate are not higher than ∼10 3 . Decontamination of water contaminated by uranium oxide and fission products in the distillation device of the RA reactor is shown on page 8. Experiments demanded special preparation due to high activity of uranium (1.7 g of uranium irradiated in the reactor for 10 days at neutron flux 1.10 13 n.cm 2 /s. Prior preparations for transport and dissolution of irradiated metal uranium as well as sampling were needed. Distillation was done under lower pressure and temperature to avoid possible contamination of the environment bu fission products and iodine. Decontamination factors are shown in Table. Contamination and decontamination of stainless steel on the laboratory level are described on page 5. It was found that the deposition of activity on the stainless steel plates is inhomogeneous showing that the uranium oxide and fission products are deposited on the rough metal surfaces. According to literature data and our laboratory studies decontamination was done by nitric acid solution (2MHNO 3 ). Since the heavy water system of the RA reactor was made of stainless teel (except the

  1. Analysed with Shanghai international fashion the development of creative industry

    Directory of Open Access Journals (Sweden)

    Jinjin Ma

    2017-05-01

    Full Text Available Rapid development of economy, and promote people to enter the era of knowledge economy. Under this background, the global economy especially the economic model of developed countries began to industrial restructuring and structural adjustment, and fashion creative industry economy is the product of the change. It embodies a nation in such aspects as culture, science and technology and creative design of soft power, to some extent, also represents a national industry's international competitiveness, is one of the most important industry in the development of leading industry. In the globalization trend of strengthening, today, the increasingly fierce competition in the international fashion scale and degree, the development of creative industry has become a measure of a country or a city comprehensive competitiveness of one of the important symbol. Therefore, many countries and regions all over the industry as a strategic industry and pillar industry to develop. Along with the rapid economic and social development as well as the consumer demand is rising, fashion creative industry gradually become Shanghai currently one of the most promising new industries. Especially in the face of the global economic downturn, China's transformation of the mode of development environment, development fashion creative industry will help speed up the Shanghai industrial structure transformation, beneficial to stimulate consumer demand, to improve the Shanghai international influence, for the Shanghai a new round of development, the construction of "four centers" and one of the breach of the international metropolis.

  2. Transit safety retrofit package development : applications requirements document.

    Science.gov (United States)

    2014-05-01

    This Application Requirements Document for the Transit Safety Retrofit Package (TRP) Development captures the : system, hardware and software requirements towards fulfilling the technical objectives stated within the contract. : To achieve the object...

  3. Risk and safety analyses for disposal of alpha-contaminated waste in INEL

    International Nuclear Information System (INIS)

    Smith, T.

    1982-01-01

    The author first discusses the context, objectives, and scope of the risk analysis. Then he gives some background on the waste and how its managed, including the alternatives for long-term management. These are followed by risk evaluation approach, results, and 7 conclusions and problems. One of his conclusions is that a 100 nCi/g limit would provide adequate safety margins. Raising the limit to 100 nCi/g would allow about 20% of the stored waste to be diverted to near-surface disposal. He added that analyzing waste packages at 10 nCi/g is not now practical. 21 figures

  4. Recent developments in Ontario Hydro's safety emergency preparedness plan program

    International Nuclear Information System (INIS)

    Leung, T.; Smith, G.

    1996-01-01

    Emergency Preparedness Plans (EPPs) are essential during dam emergencies or failures; the absence of legislation about dam safety does not exempt dam owners of legal liabilities and responsibilities for dam safety. Ontario Hydro instituted a dam safety EPP program in 1992. An overview of the Plan, and its recent integration with overall site emergency planning was described, with case studies to illustrate development, implementation, testing and training processes. A recent development was the network version of the EPP, which not only streamlined maintenance of the Plan, but also increased its effectiveness by providing instant access to the most up-to-date version of the Plan. 6 refs., 5 figs

  5. Challenges on innovations of newly-developed safety analysis codes

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Yanhua [Shanghai Jiao Tong Univ. (China). School of Nuclear Science and Engineering; Zhang, Hao [State Nuclear Power Software Development Center, Beijing (China). Beijing Future Science and Technology City

    2016-05-15

    With the development of safety analysis method, the safety analysis codes meet more challenges. Three challenges are presented in this paper, which are mathematic model, code design and user interface. Combined with the self-reliance safety analysis code named COSINE, the ways of meeting these requirements are suggested, that is to develop multi-phases, multi-fields and multi-dimension models, to adopt object-oriented code design ideal and to improve the way of modeling, calculation control and data post-processing in the user interface.

  6. Difficulties in using Material Safety Data Sheets to analyse occupational exposures to contact allergens

    DEFF Research Database (Denmark)

    Friis, Ulrik F; Menné, Torkil; Flyvholm, Mari-Ann

    2015-01-01

    BACKGROUND: Information on the occurrence of contact allergens and irritants is crucial for the diagnosis of occupational contact dermatitis. Material Safety Data Sheets (MSDS) are important sources of information concerning exposures in the workplace. OBJECTIVE: From a medical viewpoint, to eval......BACKGROUND: Information on the occurrence of contact allergens and irritants is crucial for the diagnosis of occupational contact dermatitis. Material Safety Data Sheets (MSDS) are important sources of information concerning exposures in the workplace. OBJECTIVE: From a medical viewpoint......, to evaluate the information available from MSDSs, and to ascertain whether MSDS are easy to obtain, whether they serve their purpose, and whether they provide sufficient information regarding allergens to enable correct diagnosis. METHODS: MSDS and ingredients labelling were collected from consecutive...... patients and reviewed. If it was suspected that the MSDS were incomplete, the manufacturer, supplier, salesperson or workplace was contacted to gather more information. RESULTS: Twenty-five per cent (79/316) of patients provided material for the exposure assessment. One or more shortcomings were found...

  7. The development and validation of a psychological contract of safety scale.

    Science.gov (United States)

    Walker, Arlene

    2010-08-01

    This paper builds on previous research by the author and describes the development and validation of a new measure of the psychological contract of safety. The psychological contract of safety is defined as the beliefs of individuals about reciprocal safety obligations inferred from implicit and explicit promises. A psychological contract is established when an individual believes that perceived employer and employee safety obligations are contingent on each other. A pilot test of the measure is first undertaken with participants from three different occupations: nurses, construction workers, and meat processing workers (N=99). Item analysis is used to refine the measure and provide initial validation of the scale. A larger validation study is then conducted with a participant sample of health care workers (N=424) to further refine the measure and to determine the psychometric properties of the scale. Item and correlational analyses produced the final employer and employee obligations scales, consisting of 21 and 17 items, respectively. Factor analyses identified two underlying dimensions in each scale comparable to that previously established in the organizational literature. These transactional and relational-type obligations provided construct validity of the scale. Internal consistency ratings using Cronbach's alpha found the components of the psychological contract of safety measure to be reliable. The refined and validated psychological contract of safety measure will allow investigation of the positive and negative outcomes associated with fulfilment and breach of the psychological contract of safety in future research. 2010 Elsevier Ltd. All rights reserved.

  8. Nuclear safety culture in Finland and Sweden - Developments and challenges

    International Nuclear Information System (INIS)

    Reiman, T.; Pietikaeinen, E.; Kahlbom, U.; Rollenhagen, C.

    2011-02-01

    The project aimed at studying the concept of nuclear safety culture and the Nordic nuclear branch safety culture. The project also aimed at looking how the power companies and the regulators view the current responsibilities and role of subcontractors in the Nordic nuclear safety culture as well as to inspect the special demands for safety culture in subcontracting chains. Interview data was collected in Sweden (n = 14) and Finland (n = 16) during 2009. Interviewees represented the major actors in the nuclear field (regulators, power companies, expert organizations, waste management organizations). Results gave insight into the nature and evaluation of safety culture in the nuclear industry. Results illustrated that there is a wide variety of views on matters that are considered important for nuclear safety within the Nordic nuclear community. However, the interviewees considered quite uniformly such psychological states as motivation, mindfulness, sense of control, understanding of hazards and sense of responsibility as important for nuclear safety. Results also gave insight into the characteristics of Nordic nuclear culture. Various differences in safety cultures in Finland and Sweden were uncovered. In addition to the differences, historical reasons for the development of the nuclear safety cultures in Finland and Sweden were pointed out. Finally, results gave implications that on the one hand subcontractors can bring new ideas and improvements to the plants' practices, but on the other hand the assurance of necessary safety attitudes and competence of the subcontracting companies and their employees is considered as a challenge. The report concludes that a good safety culture requires a deep and wide understanding of nuclear safety including the various accident mechanisms of the power plants as well as a willingness to continuously develop one's competence and understanding. An effective and resilient nuclear safety culture has to foster a constant sense of

  9. Nuclear safety culture in Finland and Sweden - Developments and challenges

    Energy Technology Data Exchange (ETDEWEB)

    Reiman, T.; Pietikaeinen, E. (Technical Research Centre of Finland, VTT (Finland)); Kahlbom, U. (RiskPilot AB (Sweden)); Rollenhagen, C. (Royal Institute of Technology (KTH) (Sweden))

    2011-02-15

    The project aimed at studying the concept of nuclear safety culture and the Nordic nuclear branch safety culture. The project also aimed at looking how the power companies and the regulators view the current responsibilities and role of subcontractors in the Nordic nuclear safety culture as well as to inspect the special demands for safety culture in subcontracting chains. Interview data was collected in Sweden (n = 14) and Finland (n = 16) during 2009. Interviewees represented the major actors in the nuclear field (regulators, power companies, expert organizations, waste management organizations). Results gave insight into the nature and evaluation of safety culture in the nuclear industry. Results illustrated that there is a wide variety of views on matters that are considered important for nuclear safety within the Nordic nuclear community. However, the interviewees considered quite uniformly such psychological states as motivation, mindfulness, sense of control, understanding of hazards and sense of responsibility as important for nuclear safety. Results also gave insight into the characteristics of Nordic nuclear culture. Various differences in safety cultures in Finland and Sweden were uncovered. In addition to the differences, historical reasons for the development of the nuclear safety cultures in Finland and Sweden were pointed out. Finally, results gave implications that on the one hand subcontractors can bring new ideas and improvements to the plants' practices, but on the other hand the assurance of necessary safety attitudes and competence of the subcontracting companies and their employees is considered as a challenge. The report concludes that a good safety culture requires a deep and wide understanding of nuclear safety including the various accident mechanisms of the power plants as well as a willingness to continuously develop one's competence and understanding. An effective and resilient nuclear safety culture has to foster a constant

  10. Development and use of safety indicators at STUK

    International Nuclear Information System (INIS)

    Tiippana, Petteri

    2002-01-01

    This paper presents general outline of the use and development of YTO Indicator System at the department of Nuclear Reactor Regulation (YTO) in the Radiation and Nuclear Safety Authority, STUK. The purpose of the YTO Indicator System is to be a complementary tool in the nuclear safety regulation in addition to inspections and safety reviews. The development project of indicators and indicator system has shown that properly chosen and defined indicators provide an objective tool for the regulator to control nuclear safety and also to evaluate it's own activities and effectiveness. Development of indicator system has been a long project and has not been finished yet. YTO Indicator System is divided into two main areas; safety of nuclear facilities and regulatory activities. Safety of nuclear facilities is divided into 3 areas based on the concept of defence in depth; safety and quality culture, operational events and physical barriers. Regulatory activities is also divided into 3 areas; working processes, resource management and regeneration and ability to work. These areas are measured using several indicators. At the moment some of these indicators are included in YTO's management system to measure whether internal goals are achieved or not. (author)

  11. A human error taxonomy for analysing healthcare incident reports: assessing reporting culture and its effects on safety perfomance

    DEFF Research Database (Denmark)

    Itoh, Kenji; Omata, N.; Andersen, Henning Boje

    2009-01-01

    The present paper reports on a human error taxonomy system developed for healthcare risk management and on its application to evaluating safety performance and reporting culture. The taxonomy comprises dimensions for classifying errors, for performance-shaping factors, and for the maturity...

  12. LOCA, LOFA and LOVA analyses pertaining to NET/ITER safety design guidance

    International Nuclear Information System (INIS)

    Ebert, E.; Raeder, J.

    1991-01-01

    The analyses presented pertain to loss of coolant accidents (LOCA), loss of coolant flow accidents (LOFA) and loss of vacuum accidents (LOVA). These types of accidents may jeopardise components and plasma vessel integrity and cause radioactivity mobilisation. The analyses reviewed have been performed under the assumption that the plasma facing components are protected by a carbon based armour. Accidental temperatures and pressure transients are quantified, the possibility of reaction products combustion is investigated and worst case accidental public doses are assessed. On this basis, design recommendations are given and design features such as low plasma facing components armour temperatures (on almost the entire surface) and inert gas adjacent to the vacuum vessel have been implemented. (orig.)

  13. [Supervised exercise training in patients with pulmonary arterial hypertension - analyses of the effectiveness and safety].

    Science.gov (United States)

    Saxer, S; Rhyner, M; Treder, U; Speich, R; van Gestel, A J R

    2012-02-01

    Both in today's scientific research and in clinical practice, there exists a need to address the uncertainty concerning the effectiveness and safety of cardiopulmonary exercise training (CPET) in patients with pulmonary arterial hypertension (PAH). It is commonly believed that CPET may be dangerous for patients with PAH, because increasing pressure on the pulmonary arteries may worsen right-sided heart failure. Recently, the first clinical trials on exercise training in patients with pulmonary hypertension reported promising results. Extension of the walking distance at the 6-minute walk test improved quality of life, endurance capacity and a reduction in symptoms were observed after CPET. Furthermore, CPET was well tolerated by the patients in five clinical trials. In conclusion, it may be postulated that CPET is an effective therapy in patients with PAH and was tendentially well tolerated by the patients.

  14. Generic Safety Requirements for Developing Safe Insulin Pump Software

    Science.gov (United States)

    Zhang, Yi; Jetley, Raoul; Jones, Paul L; Ray, Arnab

    2011-01-01

    Background The authors previously introduced a highly abstract generic insulin infusion pump (GIIP) model that identified common features and hazards shared by most insulin pumps on the market. The aim of this article is to extend our previous work on the GIIP model by articulating safety requirements that address the identified GIIP hazards. These safety requirements can be validated by manufacturers, and may ultimately serve as a safety reference for insulin pump software. Together, these two publications can serve as a basis for discussing insulin pump safety in the diabetes community. Methods In our previous work, we established a generic insulin pump architecture that abstracts functions common to many insulin pumps currently on the market and near-future pump designs. We then carried out a preliminary hazard analysis based on this architecture that included consultations with many domain experts. Further consultation with domain experts resulted in the safety requirements used in the modeling work presented in this article. Results Generic safety requirements for the GIIP model are presented, as appropriate, in parameterized format to accommodate clinical practices or specific insulin pump criteria important to safe device performance. Conclusions We believe that there is considerable value in having the diabetes, academic, and manufacturing communities consider and discuss these generic safety requirements. We hope that the communities will extend and revise them, make them more representative and comprehensive, experiment with them, and use them as a means for assessing the safety of insulin pump software designs. One potential use of these requirements is to integrate them into model-based engineering (MBE) software development methods. We believe, based on our experiences, that implementing safety requirements using MBE methods holds promise in reducing design/implementation flaws in insulin pump development and evolutionary processes, therefore improving

  15. Development of digital image correlation method to analyse crack ...

    Indian Academy of Sciences (India)

    The detection of crack development in a masonry wall forms an important study for investigating the earthquake resistance capability of the masonry structures. Traditionally, inspecting the structure and documenting the findings were done manually. The procedures are time-consuming, and the results are sometimes ...

  16. Development of digital image correlation method to analyse crack ...

    Indian Academy of Sciences (India)

    The wall deficiency is very difficult to detect in practical experiments because crack formation and development cannot be measured accurately. Therefore, the digital image correlation method is proposed in this study to observe the surface deformation of brick wall; the feasibility of applying this method for crack observation ...

  17. A nonlinear approach to analyse the development of tropical ...

    Indian Academy of Sciences (India)

    mation of thunderstorms, cumulus cloud development, hurricanes, tropical cyclones etc. In the present study, we wish to implement the concept of resonance in a convective sys- tem to perceive the sequence of abrupt changes or consequences like a hike or lowering of temperature, energy, pressure and so on.

  18. A nonlinear approach to analyse the development of tropical ...

    Indian Academy of Sciences (India)

    ous components constituting atmospheric circulation like vertical wind shear and tropical waves is studied with the help of the phenomenon called vibrational resonance. Also, we discuss how the enhancement in the response of the convective system causes various environmental factors leading to the development of ...

  19. The Development of an Analyses-Intensive Software for Improved ...

    African Journals Online (AJOL)

    The computer-aided software developed in this research work is used in designing cam systems by generating various follower motions and cam profiles. It is highly suited for extensive dynamics, kinematics and geometric design analysis based on some inherent features that are unique. The plate cam with either flat-face ...

  20. A nonlinear approach to analyse the development of tropical ...

    Indian Academy of Sciences (India)

    The development of atmospheric disturbances in the tropical region is explained using vibrational resonance, a nonlinear phenomenon. As the Lorenz system is the most plausible model to describe the convective process in a tropical region, the influence of vertical wind shear and tropical waves on the system leading to ...

  1. Transcriptomic and metabolite analyses of Cabernet Sauvignon grape berry development

    Science.gov (United States)

    Deluc, Laurent G; Grimplet, Jérôme; Wheatley, Matthew D; Tillett, Richard L; Quilici, David R; Osborne, Craig; Schooley, David A; Schlauch, Karen A; Cushman, John C; Cramer, Grant R

    2007-01-01

    Background Grape berry development is a dynamic process that involves a complex series of molecular genetic and biochemical changes divided into three major phases. During initial berry growth (Phase I), berry size increases along a sigmoidal growth curve due to cell division and subsequent cell expansion, and organic acids (mainly malate and tartrate), tannins, and hydroxycinnamates accumulate to peak levels. The second major phase (Phase II) is defined as a lag phase in which cell expansion ceases and sugars begin to accumulate. Véraison (the onset of ripening) marks the beginning of the third major phase (Phase III) in which berries undergo a second period of sigmoidal growth due to additional mesocarp cell expansion, accumulation of anthocyanin pigments for berry color, accumulation of volatile compounds for aroma, softening, peak accumulation of sugars (mainly glucose and fructose), and a decline in organic acid accumulation. In order to understand the transcriptional network responsible for controlling berry development, mRNA expression profiling was conducted on berries of V. vinifera Cabernet Sauvignon using the Affymetrix GeneChip® Vitis oligonucleotide microarray ver. 1.0 spanning seven stages of berry development from small pea size berries (E-L stages 31 to 33 as defined by the modified E-L system), through véraison (E-L stages 34 and 35), to mature berries (E-L stages 36 and 38). Selected metabolites were profiled in parallel with mRNA expression profiling to understand the effect of transcriptional regulatory processes on specific metabolite production that ultimately influence the organoleptic properties of wine. Results Over the course of berry development whole fruit tissues were found to express an average of 74.5% of probes represented on the Vitis microarray, which has 14,470 Unigenes. Approximately 60% of the expressed transcripts were differentially expressed between at least two out of the seven stages of berry development (28% of

  2. Transcriptomic and metabolite analyses of Cabernet Sauvignon grape berry development.

    Science.gov (United States)

    Deluc, Laurent G; Grimplet, Jérôme; Wheatley, Matthew D; Tillett, Richard L; Quilici, David R; Osborne, Craig; Schooley, David A; Schlauch, Karen A; Cushman, John C; Cramer, Grant R

    2007-11-22

    Grape berry development is a dynamic process that involves a complex series of molecular genetic and biochemical changes divided into three major phases. During initial berry growth (Phase I), berry size increases along a sigmoidal growth curve due to cell division and subsequent cell expansion, and organic acids (mainly malate and tartrate), tannins, and hydroxycinnamates accumulate to peak levels. The second major phase (Phase II) is defined as a lag phase in which cell expansion ceases and sugars begin to accumulate. Véraison (the onset of ripening) marks the beginning of the third major phase (Phase III) in which berries undergo a second period of sigmoidal growth due to additional mesocarp cell expansion, accumulation of anthocyanin pigments for berry color, accumulation of volatile compounds for aroma, softening, peak accumulation of sugars (mainly glucose and fructose), and a decline in organic acid accumulation. In order to understand the transcriptional network responsible for controlling berry development, mRNA expression profiling was conducted on berries of V. vinifera Cabernet Sauvignon using the Affymetrix GeneChip Vitis oligonucleotide microarray ver. 1.0 spanning seven stages of berry development from small pea size berries (E-L stages 31 to 33 as defined by the modified E-L system), through véraison (E-L stages 34 and 35), to mature berries (E-L stages 36 and 38). Selected metabolites were profiled in parallel with mRNA expression profiling to understand the effect of transcriptional regulatory processes on specific metabolite production that ultimately influence the organoleptic properties of wine. Over the course of berry development whole fruit tissues were found to express an average of 74.5% of probes represented on the Vitis microarray, which has 14,470 Unigenes. Approximately 60% of the expressed transcripts were differentially expressed between at least two out of the seven stages of berry development (28% of transcripts, 4,151 Unigenes

  3. Transcriptomic and metabolite analyses of Cabernet Sauvignon grape berry development

    Directory of Open Access Journals (Sweden)

    Schlauch Karen A

    2007-11-01

    Full Text Available Abstract Background Grape berry development is a dynamic process that involves a complex series of molecular genetic and biochemical changes divided into three major phases. During initial berry growth (Phase I, berry size increases along a sigmoidal growth curve due to cell division and subsequent cell expansion, and organic acids (mainly malate and tartrate, tannins, and hydroxycinnamates accumulate to peak levels. The second major phase (Phase II is defined as a lag phase in which cell expansion ceases and sugars begin to accumulate. Véraison (the onset of ripening marks the beginning of the third major phase (Phase III in which berries undergo a second period of sigmoidal growth due to additional mesocarp cell expansion, accumulation of anthocyanin pigments for berry color, accumulation of volatile compounds for aroma, softening, peak accumulation of sugars (mainly glucose and fructose, and a decline in organic acid accumulation. In order to understand the transcriptional network responsible for controlling berry development, mRNA expression profiling was conducted on berries of V. vinifera Cabernet Sauvignon using the Affymetrix GeneChip® Vitis oligonucleotide microarray ver. 1.0 spanning seven stages of berry development from small pea size berries (E-L stages 31 to 33 as defined by the modified E-L system, through véraison (E-L stages 34 and 35, to mature berries (E-L stages 36 and 38. Selected metabolites were profiled in parallel with mRNA expression profiling to understand the effect of transcriptional regulatory processes on specific metabolite production that ultimately influence the organoleptic properties of wine. Results Over the course of berry development whole fruit tissues were found to express an average of 74.5% of probes represented on the Vitis microarray, which has 14,470 Unigenes. Approximately 60% of the expressed transcripts were differentially expressed between at least two out of the seven stages of berry

  4. Initial development of a practical safety audit tool to assess fleet safety management practices.

    Science.gov (United States)

    Mitchell, Rebecca; Friswell, Rena; Mooren, Lori

    2012-07-01

    Work-related vehicle crashes are a common cause of occupational injury. Yet, there are few studies that investigate management practices used for light vehicle fleets (i.e. vehicles less than 4.5 tonnes). One of the impediments to obtaining and sharing information on effective fleet safety management is the lack of an evidence-based, standardised measurement tool. This article describes the initial development of an audit tool to assess fleet safety management practices in light vehicle fleets. The audit tool was developed by triangulating information from a review of the literature on fleet safety management practices and from semi-structured interviews with 15 fleet managers and 21 fleet drivers. A preliminary useability assessment was conducted with 5 organisations. The audit tool assesses the management of fleet safety against five core categories: (1) management, systems and processes; (2) monitoring and assessment; (3) employee recruitment, training and education; (4) vehicle technology, selection and maintenance; and (5) vehicle journeys. Each of these core categories has between 1 and 3 sub-categories. Organisations are rated at one of 4 levels on each sub-category. The fleet safety management audit tool is designed to identify the extent to which fleet safety is managed in an organisation against best practice. It is intended that the audit tool be used to conduct audits within an organisation to provide an indicator of progress in managing fleet safety and to consistently benchmark performance against other organisations. Application of the tool by fleet safety researchers is now needed to inform its further development and refinement and to permit psychometric evaluation. Copyright © 2012 Elsevier Ltd. All rights reserved.

  5. Development and application of digital safety system in NPPs

    International Nuclear Information System (INIS)

    Kwon, Keechoon; Kim, Changhwoi; Lee, Dongyoung

    2012-01-01

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

  6. 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

  7. Boron analyses in the reactor coolant system of French PWR by acid-base titration ([B]) and ICP-MS (10B atomic %): key to NPP safety

    International Nuclear Information System (INIS)

    Jouvet, Fabien; Roux, Sylvie; Carabasse, Stephanie; Felgines, Didier

    2012-09-01

    Boron is widely used by Nuclear Power Plants and especially by EDF Pressurized Water Reactors to ensure the control of the neutron rate in the reactor coolant system and, by this way, the fission reaction. The Boron analysis is thus a major factor of safety which enables operators to guarantee the permanent control of the reactor. Two kinds of analyses carried out by EDF on the Boron species, recently upgraded regarding new method validation standards and developed to enhance the measurement quality by reducing uncertainties, will be discussed in this topic: Acid-Base titration of Boron and Boron isotopic composition by Inductively Coupled Plasma Mass Spectrometer - ICP MS. (authors)

  8. Guidelines for nuclear power plant safety issue prioritization information development. Supplement 5

    Energy Technology Data Exchange (ETDEWEB)

    Daling, P.M.; Lavender, J.C. [Pacific Northwest National Lab., Richland, WA (United States)

    1996-07-01

    This is the sixth in a series of reports to document the development and use of a methodology developed by the Pacific Northwest Laboratory (PNL) to calculate, for prioritization purposes, the risk, dose, and cost impacts of implementing potential resolutions to reactor safety issues (see NUREG/CR-2800, Andrews, et al., 1983). This report contains the results of issue-specific analyses for 34 generic issues. Each issue was considered within the constraints of available information at the time the issues were examined and approximately 2 staff-weeks of labor. The results are referenced as one consideration in NUREG-0933, A Prioritization of Generic Safety Issues (Emrit, et al., 1983).

  9. The importance of probabilistic evaluations in connection with risk analyses according to technical safety laws

    International Nuclear Information System (INIS)

    Mathiak, E.

    1984-01-01

    The nuclear energy sector exemplifies the essential importance to be attached to the practical application of probabilistic evaluations (e.g. probabilistic reliability analyses) in connection with the legal risk assessment of technical systems and installations. The study is making use of a triad risk analysis and tries to reconcile the natural science and legal points of view. Without changing the definitions of 'risk' and 'hazard' in the legal sense of their meaning the publication discusses their reconcilation with the laws of natural science, their interpretation and application in view of the latter. (HSCH) [de

  10. Developing safety culture-rocket science or common sense?

    Energy Technology Data Exchange (ETDEWEB)

    Mahn, J.A.

    1998-08-01

    Despite evidence of significant management contributions to the causes of major accidents, recent events at Millstone Nuclear Power Station in the US and Ontario Hydro in Canada might lead one to conclude that the significance of safety culture, and the role of management in developing and maintaining an appropriate safety culture, is either not being understood or not being taken serious as integral to the safe operation of some complex, high-reliability operations. It is the purpose of this paper to address four aspects of management that are particularly important to safety culture, and to illustrate how development of an appropriate safety culture is more a matter of common sense than rocket science.

  11. Developing safety culture-rocket science or common sense?

    International Nuclear Information System (INIS)

    Mahn, J.A.

    1998-01-01

    Despite evidence of significant management contributions to the causes of major accidents, recent events at Millstone Nuclear Power Station in the US and Ontario Hydro in Canada might lead one to conclude that the significance of safety culture, and the role of management in developing and maintaining an appropriate safety culture, is either not being understood or not being taken serious as integral to the safe operation of some complex, high-reliability operations. It is the purpose of this paper to address four aspects of management that are particularly important to safety culture, and to illustrate how development of an appropriate safety culture is more a matter of common sense than rocket science

  12. Development of a Safety Assessment Information System for the Management of Periodic Safety Assessment Activities

    Energy Technology Data Exchange (ETDEWEB)

    Song, Tae Young [Nuclear Engineering and Technology Institute, Daejeon (Korea, Republic of)

    2007-07-01

    At present, the 10-year Periodic Safety Review(PSR) has been performing to confirm all the aspects of safety issues for all the operating plants in compliance with domestic nuclear law of article 23, subarticle 3. For each plant, in addition, Probabilistic Safety Assessment(PSA) and Severe Accident Management Guideline(SAMG) are being implemented and revised periodically to reflect the latest safety level according to principle fulfillment of severe accident policy statement. The assessment reports, as one of outcomes from these activities, are submitted into and reviewed by domestic regulatory body. During reviewing (in-office duty) and licensing (regulatory duty) process, a large number of outcomes of which most are the formal technical reports and licensing materials, are inevitably produced. Moreover, repeated review process over the plants can make them accumulated and produce a variety of documents additionally. This circumstance motivates to develop effective tool or system for the management of these reports and related technical documents for the future use in licensing process and for subsequent plant assessments. This paper presents the development status of Safety Assessment Information System(SAIS) which manages safety-related documents of PSR, PSA and SAMG for practical use for experienced engineers in charge of these areas.

  13. Risk-Informed Safety Margin Characterization Methods Development Work

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Curtis L [Idaho National Lab. (INL), Idaho Falls, ID (United States); Ma, Zhegang [Idaho National Lab. (INL), Idaho Falls, ID (United States); Riley, Tom [Idaho National Lab. (INL), Idaho Falls, ID (United States); Mandelli, Diego [Idaho National Lab. (INL), Idaho Falls, ID (United States); Nielsen, Joseph W [Idaho National Lab. (INL), Idaho Falls, ID (United States); Alfonsi, Andrea [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rabiti, Cristian [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-09-01

    This report summarizes the research activity developed during the Fiscal year 2014 within the Risk Informed Safety Margin and Characterization (RISMC) pathway within the Light Water Reactor Sustainability (LWRS) campaign. This research activity is complementary to the one presented in the INL/EXT-??? report which shows advances Probabilistic Risk Assessment Analysis using RAVEN and RELAP-7 in conjunction to novel flooding simulation tools. Here we present several analyses that prove the values of the RISMC approach in order to assess risk associated to nuclear power plants (NPPs). We focus on simulation based PRA which, in contrast to classical PRA, heavily employs system simulator codes. Firstly we compare, these two types of analyses, classical and RISMC, for a Boiling water reactor (BWR) station black out (SBO) initiating event. Secondly we present an extended BWR SBO analysis using RAVEN and RELAP-5 which address the comments and suggestions received about he original analysis presented in INL/EXT-???. This time we focus more on the stochastic analysis such probability of core damage and on the determination of the most risk-relevant factors. We also show some preliminary results regarding the comparison between RELAP5-3D and the new code RELAP-7 for a simplified Pressurized Water Reactors system. Lastly we present some conceptual ideas regarding the possibility to extended the RISMC capabilities from an off-line tool (i.e., as PRA analysis tool) to an online-tool. In this new configuration, RISMC capabilities can be used to assist and inform reactor operator during real accident scenarios.

  14. Consequence modeling for nuclear weapons probabilistic cost/benefit analyses of safety retrofits

    Energy Technology Data Exchange (ETDEWEB)

    Harvey, T.F.; Peters, L.; Serduke, F.J.D.; Hall, C.; Stephens, D.R.

    1998-01-01

    The consequence models used in former studies of costs and benefits of enhanced safety retrofits are considered for (1) fuel fires; (2) non-nuclear detonations; and, (3) unintended nuclear detonations. Estimates of consequences were made using a representative accident location, i.e., an assumed mixed suburban-rural site. We have explicitly quantified land- use impacts and human-health effects (e.g. , prompt fatalities, prompt injuries, latent cancer fatalities, low- levels of radiation exposure, and clean-up areas). Uncertainty in the wind direction is quantified and used in a Monte Carlo calculation to estimate a range of results for a fuel fire with uncertain respirable amounts of released Pu. We define a nuclear source term and discuss damage levels of concern. Ranges of damages are estimated by quantifying health impacts and property damages. We discuss our dispersal and prompt effects models in some detail. The models used to loft the Pu and fission products and their particle sizes are emphasized.

  15. DNA Analyses in Food Safety and Quality: Current Status and Expectations

    Science.gov (United States)

    Marchelli, Rosangela; Tedeschi, Tullia; Tonelli, Alessandro

    Food safety and quality are very important issues receiving a lot of attention in most countries by producers, consumers and regulatory and control authorities. In particular, DNA analysis in food is becoming popular not only in relation to genetically modified products (GMOs), in which DNA modification is the "clue" of the novelty, but also in other fields like microbiology and pathogen detection, which require long times for the cultivation and specially in cases in which the microorganisms are not cultivable like some viruses, as well as for authenticity and allergen detection. A new topic concerning "nutrigenetics and nutrigenomics" has also been mentioned, very important but still in its infancy, which could lead in the future to a personalized diet. In this chapter we have described the main areas of food research and fields of application where DNA analysis is being performed and the relative methods of detection, which are generally based on PCR. The possibility/opportunity to detect DNA without previous amplification (PCR-free) will be discussed. We have examined the following areas: (1) genetically modified foods (GMOs); (2) food allergens; (3) microbiological contaminations; (4) food authenticity; (5) nutrigenetics/nutrigenomics.

  16. Research and development program in reactor safety for NUCLEBRAS

    International Nuclear Information System (INIS)

    Pinheiro, R.B.; Resende Lobo, A.A. de; Horta, J.A.L.; Avelar Esteves, F. de; Lepecki, W.P.S.; Mohr, K.; Selvatici, E.

    1984-01-01

    With technical assistance from the IAEA, it was established recently an analytical and experimental Research and Development Program for NUCLEBRAS in the area of reactor safety. The main objectives of this program is to make possible, with low investments, the active participation of NUCLEBRAS in international PWR safety research. The analytical and experimental activities of the program are described with some detail, and the main results achieved up to now are presented. (Author) [pt

  17. Design premises for a KBS-3V repository based on results from the safety assessment SR-Can and some subsequent analyses

    International Nuclear Information System (INIS)

    2009-11-01

    The objective with this report is to: - provide design premises from a long term safety aspect of a KBS-3V repository for spent nuclear fuel, to form the basis for the development of the reference design of the repository. The design premises are used as input to the documents, called production reports, that present the reference design to be analysed in the long term safety assessment SR-Site. It is the aim that the production reports should verify that the chosen design complies with the design premises given in this report, whereas this report takes the burden of justifying why these design premises are relevant. The more specific aims and objectives with the production reports are provided in these reports. The following approach is used: - The reference design analysed in SR-Can is a starting point for setting safety related design premises for the next design step. - A few design basis cases, in accordance with the definition used in the regulation SSMFS 2008:211 and mainly related to the canister, can be derived from the results of the SR-Can assessment. From these it is possible to formulate some specific design premises for the canister. - The design basis cases involve several assumptions on the state of other barriers. These implied conditions are thus set as design premises for these barriers. - Even if there are few load cases on individual barriers that can be directly derived from the analyses, SR-Can provides substantial feedback on most aspects of the analysed reference design. This feedback is also formulated as design premises. - An important part of SR-Can Main report is the formulation and assessment of safety function indicator criteria. These criteria are a basis for formulating design premises, but they are not the same as the design premises discussed in the present report. Whereas the former should be upheld throughout the assessment period, the latter refer to the initial state and must be defined such that they give a margin for

  18. Person-centered endoscopy safety checklist: Development, implementation, and evaluation.

    Science.gov (United States)

    Dubois, Hanna; Schmidt, Peter T; Creutzfeldt, Johan; Bergenmar, Mia

    2017-12-28

    To describe the development and implementation of a person-centered endoscopy safety checklist and to evaluate the effects of a "checklist intervention". The checklist, based on previously published safety checklists, was developed and locally adapted, taking patient safety aspects into consideration and using a person-centered approach. This novel checklist was introduced to the staff of an endoscopy unit at a Stockholm University Hospital during half-day seminars and team training sessions. Structured observations of the endoscopy team's performance were conducted before and after the introduction of the checklist. In addition, questionnaires focusing on patient participation, collaboration climate, and patient safety issues were collected from patients and staff. A person-centered safety checklist was developed and introduced by a multi-professional group in the endoscopy unit. A statistically significant increase in accurate patient identity verification by the physicians was noted (from 0% at baseline to 87% after 10 mo, P importance of patient participation were rated more highly after the introduction of the checklist, but this did not result in statistical significance ( P = 0.07/ P = 0.08). The patients rated almost all items as very high both before and after the introduction of the checklist; hence, no statistical difference was noted. The intervention led to increased patient identity verification by physicians - a patient safety improvement. Clear evidence of enhanced person-centeredness or team work was not found.

  19. Development of the safety PLC for plant protection system

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Chang Hwoi; Lee, Dong Young [Korea Atomic Energy Research Institute, Taejeon (Korea, Republic of)

    2005-11-15

    The safety PLC (POSAFE-Q) is developing in the Korea Nuclear Instrumentation and Control System (KNICS) R and D project. The PLC satisfies Safety Class 1E, Quality Class 1, and Seismic Category I. The software such as RTOS and firmware are developed according to safety critical software life cycle. Especially, the formal method is applied to design SRS (Software Requirement Spec.) and SDS (Software Design Specification.) for error-free. The developed software according to software life cycle is verified by independent software V and V team. The overall response time from an input to the outputs shall be 50ms or less. The prototype for the POSAFE-Q was developed and functional testing and equipment qualification tests have been underway.

  20. A review of significant events analysed in general practice: implications for the quality and safety of patient care

    Directory of Open Access Journals (Sweden)

    Bradley Nick

    2009-09-01

    Full Text Available Abstract Background Significant event analysis (SEA is promoted as a team-based approach to enhancing patient safety through reflective learning. Evidence of SEA participation is required for appraisal and contractual purposes in UK general practice. A voluntary educational model in the west of Scotland enables general practitioners (GPs and doctors-in-training to submit SEA reports for feedback from trained peers. We reviewed reports to identify the range of safety issues analysed, learning needs raised and actions taken by GP teams. Method Content analysis of SEA reports submitted in an 18 month period between 2005 and 2007. Results 191 SEA reports were reviewed. 48 described patient harm (25.1%. A further 109 reports (57.1% outlined circumstances that had the potential to cause patient harm. Individual 'error' was cited as the most common reason for event occurrence (32.5%. Learning opportunities were identified in 182 reports (95.3% but were often non-specific professional issues not shared with the wider practice team. 154 SEA reports (80.1% described actions taken to improve practice systems or professional behaviour. However, non-medical staff were less likely to be involved in the changes resulting from event analyses describing patient harm (p Conclusion The study provides some evidence of the potential of SEA to improve healthcare quality and safety. If applied rigorously, GP teams and doctors in training can use the technique to investigate and learn from a wide variety of quality issues including those resulting in patient harm. This leads to reported change but it is unclear if such improvement is sustained.

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

    African Journals Online (AJOL)

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

  2. Inspire and develop people, two key competence for safety leadership

    International Nuclear Information System (INIS)

    Gonzalez, F.; Perez, O.; Fernandez, M.; Alvarez, N.; Villadoniga, J. I.

    2014-01-01

    Developing leadership skills in organizations is key to ensuring the sustainability of excellent results in industries with high standards of safety and reliability element. In order to have a model of development of specific leadership for these organizations, Tecnatom in 2011, we initiated an internal project to find and adapt a competency model to these requirements. (Author)

  3. Safety research on fusion DEMO in Japan: Toward development of safety strategy of a water-cooled DEMO

    Energy Technology Data Exchange (ETDEWEB)

    Nakamura, Makoto, E-mail: nakamura.makoto@jaea.go.jp [Japan Atomic Energy Agency, Rokkasho 039-3212, Aomori (Japan); Tobita, Kenji; Someya, Youji; Utoh, Hiroyasu; Sakamoto, Yoshiteru [Japan Atomic Energy Agency, Rokkasho 039-3212, Aomori (Japan); Gulden, Werner [Fusion for Energy, Garching D-85748 (Germany)

    2016-11-01

    Highlights: • This paper reports the current status of a safety research on water-cooled fusion DEMO in Japan. • We report analyses of two transients: (i) complete loss of decay heat removal and (ii) major ex-VV LOCA. • The MELCOR analysis has clarified the temperature histories of the DEMO components in complete loss of decay heat removal. • A strategy to reduce the pressure load to the final barrier confining radioactive materials is proposed against the major ex-VV LOCA. - Abstract: This paper reports the current status of a safety research on water-cooled fusion DEMO in Japan. A basic strategy of development of the safety guidelines is described for DEMO based on a water-cooled solid pebble bed blanket. Clarification of safety features of the DEMO in accident situations is a key issue to develop the guidelines. Recent achievements in understanding of the safety features of the water-cooled DEMO are reported. The MELCOR analysis has clarified the temperature histories of the DEMO components in a complete loss of decay heat removal event. The transient behavior of the first wall temperature is found to be essentially different from that of ITER. The pressure load to the tokamak cooling water system vault (TCWSV) is analyzed based on a simple model equation of the energy conservation. If the amount of the primary coolant is the same as that of Slim-CS, the previous small Japanese DEMO, the discharged water does not damage the TCWSV with the volume and pressure-tightness similar to those of pressurized light water reactors. It is shown that implementation of a pressure suppression system to the small TCWSV is effective to suppress the pressure load to the second confinement barrier.

  4. Development of small reactor safety criteria in Canada

    International Nuclear Information System (INIS)

    Ernst, P.C.; French, P.M.; Axford, D.J.; Snell, V.G.

    1990-01-01

    A number of new small reactor designs have been proposed in Canada over the last several years and some have reached the stage where licensing discussions have been initiated with the Atomic Energy Control Board (AECB). An inter-organizational Small Reactor Criteria (SRC) working group was formed in 1988 to propose safety and licensing criteria for these small reactors. Two levels of criteria are proposed. The first level forms a safety philosophy and the second is a set of criteria for specific reactor applications. The safety philosophy consists of three basic safety objectives together with evaluation criteria, and fourteen fundamental principles measured by specific criteria, which must be implemented to meet the safety objectives. Two of the fourteen principles are prime: defence in depth, and safety culture; the other twelve principles can be seen as deriving from them. A benefit of this approach is that the concepts of defence in depth and safety culture become well-defined. The objectives and principles are presented in the paper and their criteria are summarized. The second level of criteria, under development, will form a safety application set and will provide small reactor criteria in a number of general areas, such as regulatory process and safety assessment, as well as for specific reactor life-cycle activities, from siting through to decommissioning. The criteria are largely deterministic. However, the frequencies and consequences of postulated accidents are assessed against numerical criteria to assist in judging the acceptability of plant design, operation, and proposed siting. All criteria proposed are designed to be testable in some evidentiary fashion, readily enabling an assessment of compliance for a given proposal

  5. Offshore safety research and development programme. Project handbook 1997

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-02-01

    The Offshore Safety Division (OSD) of the Health and Safety Executive (HSE) undertakes Research and Development (R and D) in support of its regulatory responsibility for the safety of the offshore workforce and installations. The R and D programme undertaken has a broad technical scope ranging from `Collisions` to `Wells and Well Operations`. A strategy for this Research Programme has been published. The aim of this Handbook is to provide information on R and D projects which are currently in progress or have recently been completed. A similar handbook was published last year. To assist in the search for information, projects are grouped into sections according to their strategy area: collisions; decommissioning; diving operations; diver physiology; evacuation, escape and rescue; fire and blast; floating stability; human factors; materials performance; miscellaneous; moorings failure; occupational health; oceanography and fluid loading; pipelines; process systems and process control; safety systems; structural failure (concrete); structural failure (steel); and wells and well operations. (author)

  6. Development and measurement of perioperative patient safety indicators.

    Science.gov (United States)

    Emond, Y E; Stienen, J J; Wollersheim, H C; Bloo, G J; Damen, J; Westert, G P; Boermeester, M A; Pols, M A; Calsbeek, H; Wolff, A P

    2015-06-01

    To improve perioperative patient safety, hospitals are implementing evidence-based perioperative safety guidelines. To facilitate this process, it is important to provide insight into current practice. For this purpose, we aimed to develop patient safety indicators. The RAND-modified Delphi method was used to develop a set of patient safety indicators based on the perioperative guidelines. First, a core group of experts systematically selected recommendations from the guidelines. Then, an expert panel of representative professionals appraised the recommendations against safety criteria, prioritized them and reached consensus about 11 patient safety indicators. Measurability, applicability, improvement potential (based on current practice) and discriminatory capacity of each indicator were pilot tested in eight hospitals. Seven structure, two process and two outcome indicators were developed covering the entire perioperative care process. Most indicators showed good applicability (N=11), improvement potential (N=6) and discriminatory capacity (N=7). Four indicators were difficult to measure. Improvement opportunities concerned the use of perioperative stops, timely administration of antibiotics, availability of protocols on perioperative anticoagulants and on prospective risk analysis of medical equipment, presence of a surveillance system for postoperative wound infections, and a morbidity and mortality registration. Using a systematic, stepwise method 11 patient safety indicators were developed for internal assessment, monitoring and improvement of the perioperative care process. There was large variation in guideline adherence between and within hospitals, identifying opportunities for improvement in the quality of perioperative care. © The Author 2015. Published by Oxford University Press on behalf of the British Journal of Anaesthesia. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

  7. Development of the DOE Nuclear Criticality Safety Program Web site for the nuclear criticality safety professional

    International Nuclear Information System (INIS)

    Lee, C.K.; Huang, S.; Morman, J.A.; Garcia, A.S.

    2000-01-01

    Development of the US Department of Energy (DOE) Nuclear Criticality Safety (NCS) Program (NCSP) Web site is the result of the efforts of many members of the NCS community and is maintained by Lawrence Livermore National Laboratory (LLNL) under the direction of the NCSP Management Team. This World-Wide-Web resource was developed as part of the DOE response to Defense Nuclear Facilities Safety Board (DNFSB) Recommendation 97-2, which reflected the need to make criticality safety information available to a wide audience. The NCSP Web site provides information of interest to NCS professionals and includes links to other sites actively involved in the collection and dissemination of criticality safety information. To the extent possible, the hyperlinks on this Web site direct the user to the original source of the referenced material to ensure access to the latest, most accurate version. This site is intended to provide a central location for access to relevant NCS information in a user-friendly environment for the criticality safety community

  8. Safety-Related Optimization and Analyses of an Innovative Fast Reactor Concept

    OpenAIRE

    Vezzoni, Barbara; Gabrielli, Fabrizio; Rineiski, Andrei; Marchetti, Marco; Chen, Xue-Nong; Flad, Michael; Maschek, Werner; Boccaccini, Claudia Matzerath; Zhang, Dalin

    2012-01-01

    Since a fast reactor core with uranium-plutonium fuel is not in its most reactive configuration under operating conditions, redistribution of the core materials (fuel, steel, sodium) during a core disruptive accident (CDA) may lead to recriticalities and as a consequence to severe nuclear power excursions. The prevention, or at least the mitigation, of core disruption is therefore of the utmost importance. In the current paper, we analyze an innovative fast reactor concept developed within th...

  9. Development of IFC based fire safety assesment tools

    DEFF Research Database (Denmark)

    Taciuc, Anca; Karlshøj, Jan; Dederichs, Anne

    2016-01-01

    changes need to be implemented, involving supplementary work and costs with negative impact on the client. The aim of this project is to create a set of automatic compliance checking rules for prescriptive design and to develop a web application tool for performance based design that retrieves data from...... Building Information Models (BIM) to evacuate the safety level in the building during the conceptual design stage. The findings show that the developed tools can be useful in AEC industry. Integrating BIM from conceptual design stage for analyzing the fire safety level can ensure precision in further...

  10. Criticality-safety analyses of compacted and water-flooded. SP-100 reactors

    International Nuclear Information System (INIS)

    Brandon, D.I.; Sapir, J.L.

    1986-01-01

    Reactivity calculations were performed to determine the sensitivity of three liquid metal-cooled, fast reactor designs to various accident environments. The concepts, proposed for the SP-100 Space Nuclear Power Program, included one thermionic and two fuel-pin designs. Numerous models of each core were developed to analyze the effect of core compaction and of water-flooded lattice spreading. Results indicate that those designs incorporating in-core control are least affected by core compaction and that the thermonic concept can best withstand expansion of the flooded fuel element array

  11. Clinical pharmacology: special safety considerations in drug development and pharmacovigilance.

    Science.gov (United States)

    Atuah, Kwame N; Hughes, Dyfrig; Pirmohamed, Munir

    2004-01-01

    The dose of a drug is a major determinant of its safety, and establishing a safe dose of a novel drug is a prime objective during clinical development. The design of pre-marketing clinical trials precludes the representation of important subpopulations such as children, the elderly and people with co-morbidities. Therefore, postmarketing surveillance (PMS) activities are required to monitor the safety profile of drugs in real clinical practice. Furthermore, individual variations in pharmacogenetic profiles, the immune system, drug metabolic pathways and drug-drug interactions are also important factors in the occurrence of adverse drug reactions. Thus, the safety of a drug is a major clinical consideration before and after it is marketed. A multidisciplinary approach is required to enhance the safety profile of drugs at all stages of development, including PMS activities. Clinical pharmacology encompasses a range of disciplines and forms the backbone of drug safety consideration during clinical drug development. In this review we give an overview of the clinical drug development process and consider its limitations. We present a discussion of several aspects of clinical pharmacology and their application to enhancing drug safety. Pharmacokinetic-pharmacodynamic modelling provides a method of predicting a clinically safe dose; consideration of drug pharmacokinetics in special populations may enhance safe therapeutics in a wider spectrum of patients, while pharmacogenetics provides the possibility of genotype-specific therapeutics. Pharmacovigilance activities are also discussed. Given the complex nature and unpredictability of type B reactions, PMS activities are crucial in managing the risks drugs pose to the general population. The various aspects of clinical pharmacology discussed make a strong case for this field as the backbone of optimising and promoting safe development and use of drugs.

  12. Development of a patient safety climate scale in Japan.

    Science.gov (United States)

    Matsubara, Shinichi; Hagihara, Akihito; Nobutomo, Koichi

    2008-06-01

    Although patient safety climate is an important factor in preventing adverse events in health care organizations, there is no usable Japanese scale. To develop a Japanese scale to measure patient safety climate and to evaluate its psychometric properties. Cross-sectional questionnaire survey was conducted with respect to 9 non-academic general hospitals in Japan. A total of 1878 health care professionals (nurses, therapists, technicians, pharmacists and physicians) were included in a study. The eight dimensions measuring worker attitudes (free communication flow, continuous improvement, reporting/rules compliance and patient/family involvement) and organizational factors (supervisors' safety leadership, allied professionals' safety leadership, patient safety committee leadership and rules/equipment availability) were extracted by factor analysis. The internal consistency (measured by Cronbach's alpha) and repeatability (measured by intraclass correlation) were more than 0.70 for all subscales. In addition, the mean score, the within-group interrater reliability statistic (r(wg)) and the correlation coefficients of the mean score were confirmed at the workplace level. The scale showed acceptable dimensionality, reliability and validity. It also provided workplace team-evaluation and a tool for assessing the patient safety climate at the level of the workplace.

  13. Recent development in safety regulation of nuclear fuel cycle activities

    International Nuclear Information System (INIS)

    Kato, S.

    2001-01-01

    Through the effort of deliberation and legislation over five years, Japanese government structure was reformed this January, with the aim of realizing simple, efficient and transparent administration. Under the reform, the Agency for Nuclear and Industrial Safety (ANIS) was founded in the Ministry of Economy, Trade and Industry (METI) to be responsible for safety regulation of energy-related nuclear activities, including nuclear fuel cycle activities, and industrial activities, including explosives, high-pressure gasses and mining. As one of the lessons learned from the JCO criticality accident of September 1999, it was pointed out that the government's inspection function was not enough for fuel fabrication facilities. Accordingly, new statutory regulatory activities were introduced, namely, inspection of observance of safety rules and procedures for all kinds of nuclear operators and periodic inspection of fuel fabrication facilities. In addition, in order to cope with insufficient safety education and training of workers in nuclear facilities, licensees of nuclear facilities are required by law to specify safety education and training for their workers. ANIS is committed to enforce these new regulatory activities effectively and efficiently. In addition, it is going to be prepared, in its capacity as safety regulatory authority, for future development of Japanese fuel cycle activities, including commissioning of JNFL Rokkasho reprocessing plant and possible application for licenses for JNFL MOX fabrication plant and for spent fuel interim storage facilities. (author)

  14. Recent development in safety regulation of nuclear fuel cycle activities

    International Nuclear Information System (INIS)

    Kato, S.

    2002-01-01

    Through the effort of deliberation and legislation over five years, Japanese government structure was reformed this January, with the aim of realizing simple, efficient and transparent administration. Under the reform, the Agency for Nuclear and Industrial Safety (ANIS) was founded in the Ministry of Economy, Trade and Industry (METI) to be responsible for safety regulation of energy-related nuclear activities, including nuclear fuel cycle activities, and industrial activities, including explosives, high-pressure gasses and mining. As one of the lessons learned from the JCO criticality accident of September 1999, it was pointed out that government's inspection function was not enough for fuel fabrication facilities. Accordingly, new statutory regulatory activities were introduced, namely, inspection of observance of safety rules and procedures for all kinds of nuclear operators and periodic inspection of fuel fabrication facilities. In addition, in order to cope with insufficient safety education and training of workers in nuclear facilities, licensees of nuclear facilities are required by law to specify safety education and training for their workers. ANIS is committed to enforce these new regulatory activities effectively and efficiently. In addition, it is going to be prepared for, in its capacity of safety regulatory authority, future development of Japanese fuel cycle activities, including commissioning of JNFL Rokkasho reprocessing plant and possible application for licenses for JNFL MOX fabrication plant and for spent fuel interim storage facilities. (author)

  15. Development of safety performance indicators of regulatory interest (SAFPER) in Pakistan

    International Nuclear Information System (INIS)

    Khatoon, Abida

    2002-01-01

    Safety performance indicators provide a very useful tool for monitoring operational safety of a nuclear power plant. Utilities in many countries have developed plant specific indicators for the assessment of their performance and safety. Regulators can make use of some of these indicators for their regulatory assessment. In addition to these regulatory bodies in some countries have also developed programs for the formulation of safety performance indicators which are used in monitoring operational safety and regulatory decision making. Realizing its usefulness Directorate of Nuclear Safety and Radiation Protection (DNSRP-the regulatory body in Pakistan) has also initiated a country specific program for the development of Safety Performance Indicators (SAFPER) based on data provided by the utility and that collected during the course of regulatory inspections. Selected areas of NPP operation to be monitored are: - Significant events; - Safety systems performance; - Barriers integrity; - Environment protection; - Workers radiation safety; and - Emergency Preparedness. One of the objectives of this program is also to monitor the effectiveness of DNSRP regulatory activities. IAEA framework is taken as one of the bases for our program. Safety performance will be assessed on the basis of Performance Indicators and inspection findings. DNSRP program as shown in Appendix includes the indicators in use and under development. It is felt that the term Safety Performance Indicators may be termed as 'SAFPER Indicators' to be used by the Regulators, as it is clear from this presentation that utility safety performance indicators together with the regulatory effectiveness indicators constitute the measure for the adequate safety to the public and the environment. Additional research is still necessary for: - indicator definition for the proposed and under developed indicators; - data collection systems; - thresholds; - trend analysis; - goal setting (benefit from the trend can be

  16. To improve the safety of treatments in radiotherapy by developing a safety culture

    International Nuclear Information System (INIS)

    2008-01-01

    Following the radiotherapy accidents between 2004 and 2006, the I.R.S.N. deemed necessary to lead a study on the safety of treatments in radiotherapy and on the use and the adaptation to the medical domain of safety analysis approach developed for the nuclear installations. Of this study, six mains lines of investigation appear: Endow the radiotherapy services with real referential of safety, reinforce the robustness of the organization of radiotherapy services, improve the safety of the equipment and software at the design and operating stages, improve the management of the expertise and reinforce the operating feed back on incidents and accidents. The main learning from this study is the benefit that could be gained by fitting the safety analysis concepts and methods to the specificities of radiotherapy considering the organization of it collective work, the cooperation between actors stemming from different jobs as well as the interactions between actors and technical systems in the process of the treatments, when they are put into service and during their periodic checks. (author)

  17. Safety-Related Optimization and Analyses of an Innovative Fast Reactor Concept

    Directory of Open Access Journals (Sweden)

    Dalin Zhang

    2012-06-01

    Full Text Available Since a fast reactor core with uranium-plutonium fuel is not in its most reactive configuration under operating conditions, redistribution of the core materials (fuel, steel, sodium during a core disruptive accident (CDA may lead to recriticalities and as a consequence to severe nuclear power excursions. The prevention, or at least the mitigation, of core disruption is therefore of the utmost importance. In the current paper, we analyze an innovative fast reactor concept developed within the CP-ESFR European project, focusing on the phenomena affecting the initiation and the transition phases of an unprotected loss of flow (ULOF accident. Key phenomena for the initiation phase are coolant boiling onset and further voiding of the core that lead to a reactivity increase in the case of a positive void reactivity effect. Therefore, the first level of optimization involves the reduction, by design, of the positive void effect in order to avoid entering a severe accident. If the core disruption cannot be avoided, the accident enters into the transition phase, characterized by the progression of core melting and recriticalities due to fuel compaction. Dedicated features that enhance and guarantee a sufficient and timely fuel discharge are considered for the optimization of this phase.

  18. Developing glovebox robotics to meet the national robot safety standard and nuclear safety criteria

    International Nuclear Information System (INIS)

    McMahon, T.T.; Sievers, R.H.

    1991-09-01

    Development of a glove box based robotic system by the Lawrence Livermore National Laboratory (LLNL) is reported. Safety issues addressed include planning to meet the special constraints of operations within a hazardous material glove box and with hostile environments, compliance with the current and draft national robotic system safety standards, and eventual satisfaction of nuclear material handling requirements. Special attention has been required for the revision to the robot and control system models which antedate adoption of the present national safety standard. A robotic test bed, using non-radioactive surrogates is being activated at the Lawrence Livermore National Laboratory to develop the material handling system and the process interfaces for future special nuclear material processing applications. Part of this effort is to define, test, and revise adequate safety controls to ensure success when the system is eventually deployed at a DOE site. The current system is primarily for demonstration and testing, but will evolve into the baseline configuration from which the production system is to be derived. This results in special hazards associated with research activities which may not be present on a production line. Nuclear safety is of paramount importance and has been successfully addressed for 50 years in the DOE weapons production complex. It carries its particular requirements for robot systems and manual operations, as summarized below: Criticality must be avoided (materials cannot consolidate or accumulate to approach a critical mass). Radioactive materials must be confined. The public and workers must be protected from accountable radiation exposure. Nuclear material must be readily retrievable. Nuclear safety must be conclusively demonstrated through hazards analysis. 7 refs

  19. Analyses to support development of risk-informed separation distances for hydrogen codes and standards.

    Energy Technology Data Exchange (ETDEWEB)

    LaChance, Jeffrey L.; Houf, William G. (Sandia National Laboratories, Livermore, CA); Fluer, Inc., Paso Robels, CA; Fluer, Larry (Fluer, Inc., Paso Robels, CA); Middleton, Bobby

    2009-03-01

    The development of a set of safety codes and standards for hydrogen facilities is necessary to ensure they are designed and operated safely. To help ensure that a hydrogen facility meets an acceptable level of risk, code and standard development organizations are tilizing risk-informed concepts in developing hydrogen codes and standards.

  20. Radiation protection and safety: for nuclear technology development support

    International Nuclear Information System (INIS)

    San Pedro, A.; Hermida, J.; Fraga, H.

    1996-01-01

    A study about ICRP 60 changes as well as legal aspects and radiation protection activities in Uruguay was reported in this article. Special attention in the personnel dosimetry, radioactive sources control and their national inventory, SSDL, radioactive wastes management and SAMARI system can found among main activities developed in the Radiation Protection and Nuclear Safety Sector in National Nuclear Technology Direction

  1. Nutrition, Health and Safety in Early Childhood Development ...

    African Journals Online (AJOL)

    This article investigates the nutrition, health and safety status in Early Childhood Development (ECD) programmes and its impact thereof on the quality of care and education in Harare primary schools as perceived by the school heads, ECD teachers and parents. The study is part of a larger study on assessing the quality of ...

  2. NPA applications development in the nuclear safety authority framework

    International Nuclear Information System (INIS)

    Maselj, A.; Vojnovic, D.; Gregonc, M.

    1999-01-01

    Due to the present tasks of the SNSA (Slovenian Nuclear Safety Administration) there was a need to gain a tool for analysing the transients of the Krsko Nuclear Power Plant at the SNSA. Combining the RELAP5 code with graphical interface NPA (Nuclear Plant Analyzer), the SNSA management saw an opportunity to have a powerful instrument for analyses and assessments on a user friendly basis and without high costs. The Krsko NPP Analyzer is a joint project of the SNSA and the operator, the Krsko NPP. The RELAP5/Mod2.5 input deck was constructed by the Krsko NPP's experts and their subcontractors. In 1996 the work started with translation of input model from RELAP5/Mod2.5 version to Mod3.2. This was done by Tractebel which combined NPA masks with translated input deck and constructed new dynamic function and interactive commands between graphical MMI (Man Machine Interface) and simulation code. Since Tractebel performed similar activities for the Belgian plants, their experience was used through a transfer of knowledge to the SNSA. After this phase of the project, a user of the NPP Analyzer can run accidents as SBLOCA, Main Steam Line Break, Feed Water Break, SGTR, and many other transients activating and combining interactive commands, starting from a full power operation. This project has not been finished yet. Improvements of the input deck should be done. The Critical Safety Function window will be created. The analyzer will be a helpful tool during the training program for Accident Assessment Group, which will give to the experts basic knowledge of plant operation, its systems, and physical phenomena during a steady state and transients or accidents. Also a new dimension is added to the existing safety evaluations at the SNSA to confirm the requested level of nuclear safety at the Krsko NPP. (author)

  3. Developing and Strengthening of Safety Culture at Ukrainian NPPs: Experience of NNEGC “Energoatom”

    International Nuclear Information System (INIS)

    Sheyko, Y.; Kotin, P.

    2016-01-01

    safety of nuclear power plants and the development of a deep understanding of the importance of safety approach and the practical realisation of the principles of safety culture in production activities; • Creating an atmosphere of fruitful cooperation between management and staff, the improvement of collective action and of the behavior, developing a positive safety culture; Currently NAEC “Energoatom” is making efforts to improve the effectiveness of the implementation of these projects; to analyse the emerging issues in the implementation of project both at the pilot nuclear power plant and during its subsequent extension to the rest of the NPP; to conduct generalization, systematisation and integration of the results of these projects into a single management system of safety culture for NAEC “Energatom”. Realizing the importance of safety culture to achieve the goals of safety, as well as performing for many years a whole range of measures to improve safety and to improve the safety culture, NNEGC “Energoatom” considers the need for constant attention to safety culture at all organizational levels to be the key to success, and the main driving mechanism of progress and development in this area—wide awareness of international experience and achievements in improving the safety culture, their integration and implementation in your organization. (author)

  4. Development of main steam safety valve set pressure evaluating system

    International Nuclear Information System (INIS)

    Oketani, Koichiro; Manabe, Yoshihisa.

    1991-01-01

    A main steam safety valve set pressure test is conducted for all valves during every refueling outage in Japan's PWRs. Almost all operations of the test are manually conducted by a skilled worker. In order to obtain further reliability and reduce the test time, an automatic test system using a personnel computer has been developed in accordance with system concept. Quality assurance was investigated to fix system specifications. The prototype of the system was manufactured to confirm the system reliability. The results revealed that this system had high accuracy measurement and no adverse influence on the safety valve. This system was concluded to be applicable for actual use. (author)

  5. Developing the health, safety and environment excellence instrument.

    Science.gov (United States)

    Mohammadfam, Iraj; Saraji, Gebraeil Nasl; Kianfar, Ali; Mahmoudi, Shahram

    2013-01-07

    Quality and efficiency are important issues in management systems. To increase quality, to reach best results, to move towards the continuous improvement of system and also to make the internal and external customers satisfied, it is necessary to consider the system performance measurement. In this study the Health, Safety and Environment Excellence Instrument was represented as a performance measurement tool for a wide range of health, safety and environment management systems. In this article the development of the instrument overall structure, its parts, and its test results in three organizations are presented. According to the results, the scores ranking was the managership organization, the manufacturing company and the powerhouse construction project, respectively. The results of the instrument test in three organizations show that, on the whole, the instrument has the ability to measure the performance of health, safety and environment management systems in a wide range of organizations.

  6. Multi-dimensional Code Development for Safety Analysis of LMR

    Energy Technology Data Exchange (ETDEWEB)

    Ha, K. S.; Jeong, H. Y.; Kwon, Y. M.; Lee, Y. B

    2006-08-15

    A liquid metal reactor loaded a metallic fuel has the inherent safety mechanism due to the several negative reactivity feedback. Although this feature demonstrated through experiments in the EBR-II, any of the computer programs until now did not exactly analyze it because of the complexity of the reactivity feedback mechanism. A multi-dimensional detail program was developed through the International Nuclear Energy Research Initiative(INERI) from 2003 to 2005. This report includes the numerical coupling the multi-dimensional program and SSC-K code which is used to the safety analysis of liquid metal reactors in KAERI. The coupled code has been proved by comparing the analysis results using the code with the results using SAS-SASSYS code of ANL for the UTOP, ULOF, and ULOHS applied to the safety analysis for KALIMER-150.

  7. LOFT L2-3 scoping blowdown experiment safety analyses A, B, C, CD, CE, and CF

    International Nuclear Information System (INIS)

    Perryman, J.L.; Keeler, C.D.; Good, E.G.; Saukkoriipi, L.O.

    1978-01-01

    The consequences of various postulated single failures during loss-of-fluid test (LOFT) Loss-of-Coolant Experiment (LOCE) L2-3 were analyzed for the LOCE L2-3 blowdown scoping experiment safety analysis (ESA). The following six of the LOCE L2-3 blowdown scoping ESA were performed: (1) Analysis A--failure of high-pressure injection system (HPIS) A; (2) Analysis B--failure of accumulator A; (3) Analysis C--failure of low-pressure injection system (LPIS) A, assuming accumulator A initial conditions of 550 psig and 84 0 F; (4) Analysis CD--failure of LPIS A, assuming accumulator A initial conditions of 550 psig and 150 0 F; (5) Analysis CE--failure of LPIS A, assuming accumulator A initial conditions of 650 psig and 150 0 F; and (6) Analysis CF--failure of LPIS A, assuming accumulator A initial conditions of 650 psig and 115 0 F. RELAP4/MOD5 models with conservative off-nominal initial conditions, and evaluation model (EM) options were used during the blowdown-refill phase of the analyses. RELAP4/MOD6 models with conservative input parameters and code options selected were used for the reflood phase of the analyses

  8. Computer analyses for the design, operation and safety of new isotope production reactors: A technology status review

    International Nuclear Information System (INIS)

    Wulff, W.

    1990-01-01

    A review is presented on the currently available technologies for nuclear reactor analyses by computer. The important distinction is made between traditional computer calculation and advanced computer simulation. Simulation needs are defined to support the design, operation, maintenance and safety of isotope production reactors. Existing methods of computer analyses are categorized in accordance with the type of computer involved in their execution: micro, mini, mainframe and supercomputers. Both general and special-purpose computers are discussed. Major computer codes are described, with regard for their use in analyzing isotope production reactors. It has been determined in this review that conventional systems codes (TRAC, RELAP5, RETRAN, etc.) cannot meet four essential conditions for viable reactor simulation: simulation fidelity, on-line interactive operation with convenient graphics, high simulation speed, and at low cost. These conditions can be met by special-purpose computers (such as the AD100 of ADI), which are specifically designed for high-speed simulation of complex systems. The greatest shortcoming of existing systems codes (TRAC, RELAP5) is their mismatch between very high computational efforts and low simulation fidelity. The drift flux formulation (HIPA) is the viable alternative to the complicated two-fluid model. No existing computer code has the capability of accommodating all important processes in the core geometry of isotope production reactors. Experiments are needed (heat transfer measurements) to provide necessary correlations. It is important for the nuclear community, both in government, industry and universities, to begin to take advantage of modern simulation technologies and equipment. 41 refs

  9. Application of software to development of reactor-safety codes

    International Nuclear Information System (INIS)

    Wilburn, N.P.; Niccoli, L.G.

    1980-09-01

    Over the past two-and-a-half decades, the application of new techniques has reduced hardware cost for digital computer systems and increased computational speed by several orders of magnitude. A corresponding cost reduction in business and scientific software development has not occurred. The same situation is seen for software developed to model the thermohydraulic behavior of nuclear systems under hypothetical accident situations. For all cases this is particularly noted when costs over the total software life cycle are considered. A solution to this dilemma for reactor safety code systems has been demonstrated by applying the software engineering techniques which have been developed over the course of the last few years in the aerospace and business communities. These techniques have been applied recently with a great deal of success in four major projects at the Hanford Engineering Development Laboratory (HEDL): 1) a rewrite of a major safety code (MELT); 2) development of a new code system (CONACS) for description of the response of LMFBR containment to hypothetical accidents, and 3) development of two new modules for reactor safety analysis

  10. Ferrocyanide safety project: Task 3. 5 cyanide species analytical methods development

    Energy Technology Data Exchange (ETDEWEB)

    Bryan, S.A.; Pool, K.H.; Burger, L.L.; Carlson, C.D.; Hess, N.J.; Matheson, J.D.; Ryan, J.L.; Scheele, R.D.; Tingey, J.M.

    1993-01-01

    This report summarizes the results of studies conducted in FY 1992 to develop methods for the identification and quantification of cyanide species in ferrocyanide tank waste. Currently there are 24 high-level waste storage tanks at the Hanford Site that have been placed on a Ferrocyanide Tank Watchlist because they contain an estimated 1,000 g-moles or greater amount of precipitated ferrocyanide. This amount of ferrocyanide is of concern because the consequences of a potential explosion may exceed those reported previously in safety analyses. The threshold concentration of total cyanide within the tank waste matrix that is expected to be a safety concern is estimated at approximately 1 to 3 wt%. Methods for detection and speciation of ferrocyanide complexes in actual waste are needed to definitively measure and quantitate the amount of ferrocyanides present within actual waste tanks to a lower limit of at least 0.1 wt% in order to bound the safety concern.

  11. Ferrocyanide safety project: Task 3.5 cyanide species analytical methods development. FY 1992 annual report

    Energy Technology Data Exchange (ETDEWEB)

    Bryan, S.A.; Pool, K.H.; Burger, L.L.; Carlson, C.D.; Hess, N.J.; Matheson, J.D.; Ryan, J.L.; Scheele, R.D.; Tingey, J.M.

    1993-01-01

    This report summarizes the results of studies conducted in FY 1992 to develop methods for the identification and quantification of cyanide species in ferrocyanide tank waste. Currently there are 24 high-level waste storage tanks at the Hanford Site that have been placed on a Ferrocyanide Tank Watchlist because they contain an estimated 1,000 g-moles or greater amount of precipitated ferrocyanide. This amount of ferrocyanide is of concern because the consequences of a potential explosion may exceed those reported previously in safety analyses. The threshold concentration of total cyanide within the tank waste matrix that is expected to be a safety concern is estimated at approximately 1 to 3 wt%. Methods for detection and speciation of ferrocyanide complexes in actual waste are needed to definitively measure and quantitate the amount of ferrocyanides present within actual waste tanks to a lower limit of at least 0.1 wt% in order to bound the safety concern.

  12. Ferrocyanide safety project: Task 3.5 cyanide species analytical methods development

    International Nuclear Information System (INIS)

    Bryan, S.A.; Pool, K.H.; Burger, L.L.; Carlson, C.D.; Hess, N.J.; Matheson, J.D.; Ryan, J.L.; Scheele, R.D.; Tingey, J.M.

    1993-01-01

    This report summarizes the results of studies conducted in FY 1992 to develop methods for the identification and quantification of cyanide species in ferrocyanide tank waste. Currently there are 24 high-level waste storage tanks at the Hanford Site that have been placed on a Ferrocyanide Tank Watchlist because they contain an estimated 1,000 g-moles or greater amount of precipitated ferrocyanide. This amount of ferrocyanide is of concern because the consequences of a potential explosion may exceed those reported previously in safety analyses. The threshold concentration of total cyanide within the tank waste matrix that is expected to be a safety concern is estimated at approximately 1 to 3 wt%. Methods for detection and speciation of ferrocyanide complexes in actual waste are needed to definitively measure and quantitate the amount of ferrocyanides present within actual waste tanks to a lower limit of at least 0.1 wt% in order to bound the safety concern

  13. Radiation safety assessment and development of environmental radiation monitoring technology

    CERN Document Server

    Choi, B H; Kim, S G

    2002-01-01

    The Periodic Safety Review(PSR) of the existing nuclear power plants is required every ten years according to the recently revised atomic energy acts. The PSR of Kori unit 1 and Wolsong unit 1 that have been operating more than ten years is ongoing to comply the regulations. This research project started to develop the techniques necessary for the PSR. The project developed the following four techniques at the first stage for the environmental assessment of the existing plants. 1) Establishment of the assessment technology for contamination and accumulation trends of radionuclides, 2) alarm point setting of environmental radiation monitoring system, 3) Development of Radiation Safety Evaluation Factor for Korean NPP, and 4) the evaluation of radiation monitoring system performance and set-up of alarm/warn set point. A dynamic compartment model to derive a relationship between the release rates of gas phase radionuclides and the concentrations in the environmental samples. The model was validated by comparing ...

  14. Safety

    International Nuclear Information System (INIS)

    1998-01-01

    A brief account of activities carried out by the Nuclear power plants Jaslovske Bohunice in 1997 is presented. These activities are reported under the headings: (1) Nuclear safety; (2) Industrial and health safety; (3) Radiation safety; and Fire protection

  15. Safety analysis and the code development on radioactive waste disposal

    International Nuclear Information System (INIS)

    Uchida, Masahiro

    2011-01-01

    Regarding development of the safety analysis codes to be used for 'cross-check' (which is the evaluation of the validity of the safety analysis conducted by the licensee through cross comparison of the simulated result) of the sub-surface disposal conducted by the licensee, the codes are required to be capable of confirming the long term safety of the sub-surface disposal. The influence of the rainfall infiltration change on groundwater flow over the long term period due to climate change was studied. As a result, it was found that shoreline movement caused by the sea level change significantly influenced groundwater flow. Regarding development of the safety analysis codes to be used for 'cross-check' of the near surface disposal, it is important to efficiently simulate the groundwater flow with finely discretized mesh model. We therefore improved the memory allocation algorithm of the groundwater flow simulation code, TOUGH2 to be able to treat the large mesh model, such as several million cells. Modifications are made for the simulation support system, by adding the groundwater flow code 3D-SEEP which can treat land uplift and erosion and its associated modules. This modification not only improves efficiency but also allows to avoid human error. Moreover, sensitivity analysis of the unsaturated conditions such as infiltration rate on the migration of important nuclides of near surface disposal was conducted. As a result, influence of the unsaturated conditions on the exposed dose was evaluated. (author)

  16. SAFETY

    CERN Multimedia

    C. Schaefer and N. Dupont

    2013-01-01

      “Safety is the highest priority”: this statement from CERN is endorsed by the CMS management. An interpretation of this statement may bring you to the conclusion that you should stop working in order to avoid risks. If the safety is the priority, work is not! This would be a misunderstanding and misinterpretation. One should understand that “working safely” or “operating safely” is the priority at CERN. CERN personnel are exposed to different hazards on many levels on a daily basis. However, risk analyses and assessments are done in order to limit the number and the gravity of accidents. For example, this process takes place each time you cross the road. The hazard is the moving vehicle, the stake is you and the risk might be the risk of collision between both. The same principle has to be applied during our daily work. In particular, keeping in mind the general principles of prevention defined in the late 1980s. These principles wer...

  17. Improving occupational safety and health by integration into product development

    DEFF Research Database (Denmark)

    Broberg, Ole

    1996-01-01

    A cross-sectional case study was performed in a large company producing electro-mechanical products for industrial application. The objectives were: (i) to study the product development process and the role of key actors', (ii) to identify current practice on integrating occupational safety and h...... and studies of documents. A questionnaire regarding product development tasks and occupational safety and health were distributed to 30 design and production engineers. A total of 27 completed the questionnaire corresponding to a response rate of 90 per cent.......A cross-sectional case study was performed in a large company producing electro-mechanical products for industrial application. The objectives were: (i) to study the product development process and the role of key actors', (ii) to identify current practice on integrating occupational safety...... and health into the development process, especially the efforts and attitudes of design and production engineers', and (iii) to identify key actors'reflections on how to improve this integration. The study was based on qualitative as well as quantitative methods including interviews, questionnaires...

  18. Development of a Novel Nuclear Safety Culture Evaluation Method for an Operating Team Using Probabilistic Safety Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Han, Sangmin; Lee, Seung Min; Seong, Poong Hyun [KAIST, Daejeon (Korea, Republic of)

    2015-05-15

    IAEA defined safety culture as follows: 'Safety Culture is that assembly of characteristics and attitudes in organizations and individuals which establishes that, as an overriding priority, nuclear plant safety issues receive the attention warranted by their significance'. Also, celebrated behavioral scientist, Cooper, defined safety culture as,'safety culture is that observable degree of effort by which all organizational members direct their attention and actions toward improving safety on a daily basis' with his internal psychological, situational, and behavioral context model. With these various definitions and criteria of safety culture, several safety culture assessment methods have been developed to improve and manage safety culture. To develop a new quantitative safety culture evaluation method for an operating team, we unified and redefined safety culture assessment items. Then we modeled a new safety culture evaluation by adopting level 1 PSA concept. Finally, we suggested the criteria to obtain nominal success probabilities of assessment items by using 'operational definition'. To validate the suggested evaluation method, we analyzed the collected audio-visual recording data collected from a full scope main control room simulator of a NPP in Korea.

  19. Development of a Novel Nuclear Safety Culture Evaluation Method for an Operating Team Using Probabilistic Safety Analysis

    International Nuclear Information System (INIS)

    Han, Sangmin; Lee, Seung Min; Seong, Poong Hyun

    2015-01-01

    IAEA defined safety culture as follows: 'Safety Culture is that assembly of characteristics and attitudes in organizations and individuals which establishes that, as an overriding priority, nuclear plant safety issues receive the attention warranted by their significance'. Also, celebrated behavioral scientist, Cooper, defined safety culture as,'safety culture is that observable degree of effort by which all organizational members direct their attention and actions toward improving safety on a daily basis' with his internal psychological, situational, and behavioral context model. With these various definitions and criteria of safety culture, several safety culture assessment methods have been developed to improve and manage safety culture. To develop a new quantitative safety culture evaluation method for an operating team, we unified and redefined safety culture assessment items. Then we modeled a new safety culture evaluation by adopting level 1 PSA concept. Finally, we suggested the criteria to obtain nominal success probabilities of assessment items by using 'operational definition'. To validate the suggested evaluation method, we analyzed the collected audio-visual recording data collected from a full scope main control room simulator of a NPP in Korea

  20. [Home safety and severe mental disorders: Developing an evaluation tool].

    Science.gov (United States)

    Désormeaux-Moreau, Marjorie; Dumont, Claire; Aubin, Ginette; Larivière, Nadine

    2015-04-01

    Home safety evaluation is an important issue within the context of current perspectives on accommodation for people with a serious mental illness who favour a more independent way of life. This paper describes the development and content validation of the Evaluation de la sécurité a domicile et de la gestion des risques (ESGR), an occupational therapy assessment tool for people with a serious mental illness. The ESGR was developed from scientific knowledge and clinical experience. Assessing content validity was done in two phases and involved the consultation of 11 experts. In its current form, the ESGR includes 67 items organized into three categories (environment, occupation, person). According to the experts consulted, there is a clinical interest in using the ESGR to support occupational therapists in the assessment of home safety for people with serious mental illness. The statements are clear and representative of the concept and the target audience.

  1. The development of NPP operational safety training courses

    International Nuclear Information System (INIS)

    Lee, Chang Kun; Lee, Duk Sun; Lee, Byung Sun; Lee, Won Koo; Juhn, Heng Run; Moon, Byung Soo; Cho, Min Sik; Lee, Han Young; Moon, Hak Won; Seo, Yeon Ho

    1987-12-01

    The objective of the project is to develop a training course text for the betterment of reactor operation and assurance of its safety in general by providing training materials of the advanced compact nuclear simulator which will become operation in September 1988. Main scope and contents of the project are as follows: - compilation of basic data related to simulator operation and maintenance as well as the comparative analysis with respect to simulator materials in foreign countries - method of training by simulator - review the training status by simulator in foreign countries - development of training course in the field of reactor safety It is expected that the results will be reflected to the actual training and retraining of the reactor operating crew so as to improve and update their capabilities in training fashion. (Author)

  2. Safety-related operator actions: methodology for developing criteria

    International Nuclear Information System (INIS)

    Kozinsky, E.J.; Gray, L.H.; Beare, A.N.; Barks, D.B.; Gomer, F.E.

    1984-03-01

    This report presents a methodology for developing criteria for design evaluation of safety-related actions by nuclear power plant reactor operators, and identifies a supporting data base. It is the eleventh and final NUREG/CR Report on the Safety-Related Operator Actions Program, conducted by Oak Ridge National Laboratory for the US Nuclear Regulatory Commission. The operator performance data were developed from training simulator experiments involving operator responses to simulated scenarios of plant disturbances; from field data on events with similar scenarios; and from task analytic data. A conceptual model to integrate the data was developed and a computer simulation of the model was run, using the SAINT modeling language. Proposed is a quantitative predictive model of operator performance, the Operator Personnel Performance Simulation (OPPS) Model, driven by task requirements, information presentation, and system dynamics. The model output, a probability distribution of predicted time to correctly complete safety-related operator actions, provides data for objective evaluation of quantitative design criteria

  3. Scenario Development Workshop Synopsis. Integration Group for the Safety Case - June 2015

    International Nuclear Information System (INIS)

    Smith, Paul; Voinis, Sylvie; Griffault, Lise; De Meredieu, Jean; Kwong, Gloria; ); Van Luik, Abraham; Bailey, Lucy; Capouet, Manuel; Depaus, Christophe; Makino, Hitoshi; Leigh, Christi; Kirkes, Ross; Leino, Jaakko; Niemeyer, Matthias; Wolf, Jens; Watson, Sarah; Franke, Bettina; Ilett, Doug; Pastina, Barbara; Weetjens, Eef

    2016-03-01

    safety case that also includes a broad range of evidence and arguments that complement and support the reliability of the results of the quantitative analyses. Assessments typically describe and evaluate repository evolution and potential radiological and other consequences for a range of scenarios. The present report is based largely on the presentations and discussions at the second workshop, including the working group sessions, and on a review of the questionnaire responses. It is structured as follows: - Chapter 2 summarises the work of the NEA and other international organisations on scenario development and related topics. - Chapter 3 discusses regulatory perspectives on scenario development, including general regulatory principles, more specific guidance, the level of detail in regulatory guidance and the importance of dialogue and review. - Chapter 4 describes the roles of scenario development both in safety assessments and, more generally, in the management of uncertainty in repository programmes. Its role in promoting interdisciplinary communication is also discussed. - Chapter 5 describes the broad classes into which scenarios are generally divided, including what-if scenarios and the special case of human intrusion. - Chapter 6 reviews the approaches to scenario development followed by various national programmes, including their evolution, common features and differences between programmes, the main broad steps in scenario development and the tools that have been used to implement these and also the issues of comprehensiveness and sufficiency of the sets of scenarios that are derived. - Chapter 7 discusses the analysis of scenarios, including the development of models and their application in deterministic and probabilistic calculations. - Finally, Chapter 8 summarises the main findings of this report and draws some conclusions

  4. The role of safety analyses in site selection. Some personal observations based on the experience from the Swiss site selection process

    International Nuclear Information System (INIS)

    Zuidema, Piet

    2015-01-01

    geological barrier (host rock and confining units); long-term stability (erosion, differential movements, etc.); reliability of geological information (explorability; predictability); technical feasibility (sufficient space for allocating the disposal rooms; depth of repository; rock strength, etc.). For some of these issues, rather detailed quantitative analyses are made (e.g. for erosion). Besides long-term safety, also operational safety is considered. This is done to ensure that suitable sites are chosen for the surface infrastructure (waste acceptance facilities, entrance to access to underground). The main emphasis is on external events (e.g. very severe flooding) that need to be avoided. The involvement of society in the site selection process is also very important. This requires that the scientific information needed (and wanted) by society is delivered in a format understandable to them. This helps society develop an understanding of the question ''why here and not there'' in the siting decision; something that is considered essential to get the necessary support for the siting decision.

  5. Railway safety climate: a study on organizational development.

    Science.gov (United States)

    Cheng, Yung-Hsiang

    2017-09-07

    The safety climate of an organization is considered a leading indicator of potential risk for railway organizations. This study adopts the perceptual measurement-individual attribute approach to investigate the safety climate of a railway organization. The railway safety climate attributes are evaluated from the perspective of railway system staff. We identify four safety climate dimensions from exploratory factor analysis, namely safety communication, safety training, safety management and subjectively evaluated safety performance. Analytical results indicate that the safety climate differs at vertical and horizontal organizational levels. This study contributes to the literature by providing empirical evidence of the multilevel safety climate in a railway organization, presents possible causes of the differences under various cultural contexts and differentiates between safety climate scales for diverse workgroups within the railway organization. This information can be used to improve the safety sustainability of railway organizations and to conduct safety supervisions for the government.

  6. Development of Realistic Safety Analysis Technology for CANDU Reactors

    International Nuclear Information System (INIS)

    Park, Joo Hwan; Rhee, B. W.; Rho, G. H.

    2010-04-01

    The following 3 research items have been studied to develop and establish the realistic safety analysis and the associated technologies for a CANDU reactor. At the first, WIMS-CANDU which is physics cell code for a CANDU has been improved and validated against the physics criticality experiment data transferred through the international cooperation programs. Also an improved physics model to take into account the pressure tube creep was developed and utilized to assess the effects of the pressure tube creep of 0%, 2.5% and 5% diametral increase of pressure tube on core physics parameters. Secondly, the interfacing module between physics and thermal-hydraulics codes has been developed to provide the enhancement of reliability and convenience of the calculation results of the physics parameters such as power coefficient which was calculated by independent code systems. Finally, the important parameters related to the complex heat transfer mechanisms in the crept pressure tubes were identified to find how to improve the existing fuel channel models. One of the important parameters such as the oxidation model of Zr-steam reaction was identified, implemented and verified with the experimental data of the high pressure and temperature fuel channel and its model was utilized for CFD analysis of the crept pressure tube effect on the reactor safety. The results were also provided to validate the CATNENA models of the crept pressure tube and the effects of the pressure tube creep on the blowdown and post-blowdown phase during LOCA was assessed. The results of this study can be used to assess the uncertainty analysis of coolant void reactivity and the effects of the creep deformed pressure tubes on physics/TH/safety issues. Also, those results will be used to improve the current design and operational safety analysis codes, and to technically support the related issues to resolve their problems

  7. Safety culture indicators for NPP: international trends and development status in Korea

    International Nuclear Information System (INIS)

    Choi, Y. S.; Ko, J. D.; Choi, K. S.; Jung, Y. H.

    2004-01-01

    Safety culture has been recognized as important to achieve high level of nuclear safety, as several recent events that have occurred in advanced countries were found to have important implications for safety culture. Under the recognition, implementation-focused and practical methods to foster safety culture have become necessary. Development of safety culture indicators for assessing the level of safety culture and identifying some deficiencies is being conducted. This paper examines the regulatory positions of major nuclear power countries on licensee's safety culture, introduces the development status of Korean Safety Culture Indicators and presents its future direction

  8. Occupational safety and health criteria for responsible development of nanotechnology

    Science.gov (United States)

    Schulte, P. A.; Geraci, C. L.; Murashov, V.; Kuempel, E. D.; Zumwalde, R. D.; Castranova, V.; Hoover, M. D.; Hodson, L.; Martinez, K. F.

    2014-01-01

    Organizations around the world have called for the responsible development of nanotechnology. The goals of this approach are to emphasize the importance of considering and controlling the potential adverse impacts of nanotechnology in order to develop its capabilities and benefits. A primary area of concern is the potential adverse impact on workers, since they are the first people in society who are exposed to the potential hazards of nanotechnology. Occupational safety and health criteria for defining what constitutes responsible development of nanotechnology are needed. This article presents five criterion actions that should be practiced by decision-makers at the business and societal levels—if nanotechnology is to be developed responsibly. These include (1) anticipate, identify, and track potentially hazardous nanomaterials in the workplace; (2) assess workers' exposures to nanomaterials; (3) assess and communicate hazards and risks to workers; (4) manage occupational safety and health risks; and (5) foster the safe development of nanotechnology and realization of its societal and commercial benefits. All these criteria are necessary for responsible development to occur. Since it is early in the commercialization of nanotechnology, there are still many unknowns and concerns about nanomaterials. Therefore, it is prudent to treat them as potentially hazardous until sufficient toxicology, and exposure data are gathered for nanomaterial-specific hazard and risk assessments. In this emergent period, it is necessary to be clear about the extent of uncertainty and the need for prudent actions.

  9. Development of a draft of human factors safety review procedures for the Korean next generation reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jung Woon; Moon, B. S.; Park, J. C.; Lee, Y. H.; Oh, I. S.; Lee, H. C. [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2000-02-01

    In this study, a draft of human factors engineering (HFE) safety review procedures (SRP) was developed for the safety review of KNGR based on HFE Safety and Regulatory Requirements and Guidelines (SRRG). This draft includes acceptance criteria, review procedure, and evaluation findings for the areas of review including HFE Program Management, Human Factors Analyses, Human Factors Design, and HFE Verification and Validation, based on Section 15.1 'Human Factors Engineering Design Process' and 15.2 'Control Room Human Factors Engineering' of KNGR Specific Safety Requirements and Chapter 15 'Human Factors Engineering' of KNGR Safety Regulatory Guides. For the effective review, human factors concerns or issues related to advanced HSI design that have been reported so far should be extensively examined. In this study, a total of 384 human factors issues related to the advanced HSI design were collected through our review of a total of 145 documents. A summary of each issue was described and the issues were identified by specific features of HSI design. These results were implemented into a database system. 8 refs., 2 figs. (Author)

  10. Development of a draft of human factors safety review procedures for the Korean Next Generation Reactor

    International Nuclear Information System (INIS)

    Lee, Jung Woon; Moon, B. S.; Park, J. C.; Lee, Y. H.; Oh, I. S.; Lee, H. C.

    2000-02-01

    In this study, a draft of Human Factors Engineering (HFE) Safety Review Procedures (SRP) was developed for the safety review of KNGR based on HFE Safety and Regulatory Requirements and Guidelines (SRRG). This draft includes acceptance criteria, review procedure, and evaluation findings for the areas of review including HFE program management, human factors analyses, human factors design, and HFE verification and validation, based on section 15.1 'human factors engineering design process' and 15.2 'control room human factors engineering' of KNGR specific safety requirements and chapter 15 'human factors engineering' of KNGR safety regulatory guides. For the effective review, human factors concerns or issues related to advanced HSI design that have been reported so far should be extensively examined. In this study, a total of 384 human factors issues related to the advanced HSI design were collected through our review of a total of 145 documents. A summary of each issue was described and the issues were identified by specific features of HSI design. These results were implemented into a database system

  11. Development of a safety and regulation systems simulation program II

    International Nuclear Information System (INIS)

    1985-05-01

    This report describes the development of a safety and regulation systems simulation program under contract to the Atomic Energy Control Board of Canada. A systems logic interaction simulation (SLISIM) program was developed for the AECB's HP-1000 computer which operates in the interactive simulation (INSIM) program environment. The SLISIM program simulates the spatial neutron dynamics, the regulation of the reactor power and in this version the CANDU-PHW 600 MW(e) computerized shutdown systems' trip parameters. The modular concept and interactive capability of the INSIM environment provides the user with considerable flexibility of the setup and control of the simulation

  12. Communicating vaccine safety during the development and introduction of vaccines.

    Science.gov (United States)

    Kochhar, Sonali

    2015-01-01

    Vaccines are the best defense available against infectious diseases. Vaccine safety is of major focus for regulatory bodies, vaccine manufacturers, public health authorities, health care providers and the public as vaccines are often given to healthy children and adults as well as to pregnant woman. Safety assessment is critical at all stages of vaccine development. Effective, clear and consistent communication of the risks and benefits of vaccines and advocacy during all stages of clinical research (including the preparation, approvals, conduct of clinical trials through the post marketing phase) is critically important. This needs to be done for all major stakeholders (e.g. community members, Study Team, Health Care Providers, Ministry of Health, Regulators, Ethics Committee members, Public Health Authorities and Policy Makers). Improved stakeholder alignment would help to address some of the concerns that may affect the clinical research, licensing of vaccines and their wide-spread use in immunization programs around the world.

  13. Nuclear criticality safety program development using necessary and sufficient standards

    Energy Technology Data Exchange (ETDEWEB)

    Croucher, D.W.; Stachowiak, R.V. [Kaiser-Hill Co., LLC, Golden, CO (United States); Wilson, R.E. [Safe Sites of Colorado, Golden, CO (United States)

    1996-12-31

    The U.S. Department of Energy`s (DOE`s) Necessary and Sufficient Standards Closure Process has been used to develop a new criticality, safety program manual for the Rocky Flats Environmental Technology Site (RFETS). Standards define and communicate the expectations for performance of work. The purpose of the necessary and sufficient standards closure process is to apply standards determined to be necessary and sufficient for protecting the workers, the public, and the environment. This ensures that the applied standards add value to the performance of the activity; work effectiveness is increased. The purpose of this paper is to briefly describe the process and the results for the selection of national criticality safety standards for use at the Rocky Flats facilities.

  14. Analyses and computer code developments for accident-induced thermohydraulic transients in water-cooled nuclear reactor systems

    International Nuclear Information System (INIS)

    Wulff, W.

    1977-01-01

    A review is presented on the development of analyses and computer codes for the prediction of thermohydraulic transients in nuclear reactor systems. Models for the dynamics of two-phase mixtures are summarized. Principles of process, reactor component and reactor system modeling are presented, as well as the verification of these models by comparing predicted results with experimental data. Codes of major importance are described, which have recently been developed or are presently under development. The characteristics of these codes are presented in terms of governing equations, solution techniques and code structure. Current efforts and problems of code verification are discussed. A summary is presented of advances which are necessary for reducing the conservatism currently implied in reactor hydraulics codes for safety assessment

  15. Developing the radiation protection safety culture in the UK

    International Nuclear Information System (INIS)

    Cole, P; Marsh, A; Hallard, R; Broughton, J; Coates, R; Croft, J; Davies, K; Devine, I; Lewis, C; Marsden, P; McGeary, R; Riley, P; Rogers, A; Rycraft, H; Shaw, A

    2014-01-01

    In the UK, as elsewhere, there is potential to improve how radiological challenges are addressed through improvement in, or development of, a strong radiation protection (RP) safety culture. In preliminary work in the UK, two areas have been identified as having a strong influence on UK society: the healthcare and nuclear industry sectors. Each has specific challenges, but with many overlapping common factors. Other sectors will benefit from further consideration. In order to make meaningful comparisons between these two principal sectors, this paper is primarily concerned with cultural aspects of RP in the working environment and occupational exposures rather than patient doses. The healthcare sector delivers a large collective dose to patients each year, particularly for diagnostic purposes, which continues to increase. Although patient dose is not the focus, it must be recognised that collective patient dose is inevitably linked to collective occupational exposure, especially in interventional procedures. The nuclear industry faces major challenges as work moves from operations to decommissioning on many sites. This involves restarting work in the plants responsible for the much higher radiation doses of the 1960/70s, but also performing tasks that are considerably more difficult and hazardous than those original performed in these plants. Factors which influence RP safety culture in the workplace are examined, and proposals are considered for a series of actions that may lead to an improvement in RP culture with an associated reduction in dose in many work areas. These actions include methods to improve knowledge and awareness of radiation safety, plus ways to influence management and colleagues in the workplace. The exchange of knowledge about safety culture between the nuclear industry and medical areas may act to develop RP culture in both sectors, and have a wider impact in other sectors where exposures to ionising radiations can occur. (memorandum)

  16. Application of software engineering to development of reactor safety codes

    International Nuclear Information System (INIS)

    Wilburn, N.P.; Niccoli, L.G.

    1981-01-01

    Software Engineering, which is a systematic methodology by which a large scale software development project is partitioned into manageable pieces, has been applied to the development of LMFBR safety codes. The techniques have been applied extensively in the business and aerospace communities and have provided an answer to the drastically increasing cost of developing and maintaining software. The five phases of software engineering (Survey, Analysis, Design, Implementation, and Testing) were applied in turn to development of these codes, along with Walkthroughs (peer review) at each stage. The application of these techniques has resulted in SUPERIOR SOFTWARE which is well documented, thoroughly tested, easy to modify, easier to use and maintain. The development projects have resulted in lower overall cost. (orig.) [de

  17. Foundational development of an advanced nuclear reactor integrated safety code.

    Energy Technology Data Exchange (ETDEWEB)

    Clarno, Kevin (Oak Ridge National Laboratory, Oak Ridge, TN); Lorber, Alfred Abraham; Pryor, Richard J.; Spotz, William F.; Schmidt, Rodney Cannon; Belcourt, Kenneth (Ktech Corporation, Albuquerque, NM); Hooper, Russell Warren; Humphries, Larry LaRon

    2010-02-01

    This report describes the activities and results of a Sandia LDRD project whose objective was to develop and demonstrate foundational aspects of a next-generation nuclear reactor safety code that leverages advanced computational technology. The project scope was directed towards the systems-level modeling and simulation of an advanced, sodium cooled fast reactor, but the approach developed has a more general applicability. The major accomplishments of the LDRD are centered around the following two activities. (1) The development and testing of LIME, a Lightweight Integrating Multi-physics Environment for coupling codes that is designed to enable both 'legacy' and 'new' physics codes to be combined and strongly coupled using advanced nonlinear solution methods. (2) The development and initial demonstration of BRISC, a prototype next-generation nuclear reactor integrated safety code. BRISC leverages LIME to tightly couple the physics models in several different codes (written in a variety of languages) into one integrated package for simulating accident scenarios in a liquid sodium cooled 'burner' nuclear reactor. Other activities and accomplishments of the LDRD include (a) further development, application and demonstration of the 'non-linear elimination' strategy to enable physics codes that do not provide residuals to be incorporated into LIME, (b) significant extensions of the RIO CFD code capabilities, (c) complex 3D solid modeling and meshing of major fast reactor components and regions, and (d) an approach for multi-physics coupling across non-conformal mesh interfaces.

  18. Review and Analysis of Development of "Safety by Design" Requirements

    Energy Technology Data Exchange (ETDEWEB)

    Vance, Scott A.; Hockert, John

    2009-10-20

    This report, the deliverable for Task 4 of the NA-243 Safeguards by Design Work Plan for Fiscal Year 2009, develops the lessons to be learned for the institutionalization of Safeguards By Design (SBD) from the Department of Energy (DOE) experience developing and implementing DOE-STD-1189, Integration of Safety into the Design Process. This experience was selected for study because of the similarity of the challenges of integrating safety and safeguards into the design process. Development of DOE-STD-1189 began in January 2006 and the standard was issued for implementation in March 2008. The process was much more time consuming than originally anticipated and might not have come to fruition had senior DOE management been less committed to its success. Potentially valuable lessons can be learned from both the content and presentation of the integration approach in DOE-STD-1189 and from the DOE experience in developing and implementing DOE-STD-1189. These lessons are important because the instutionalization of SBD does not yet appear to have the level of senior management commitment afforded development and implementation of DOE-STD-1189.

  19. Analysing supercritical water reactor's (SCWR's) special safety systems using probabilistic tools

    International Nuclear Information System (INIS)

    Ituen, I.; Novog, D.R.

    2011-01-01

    The next generation of reactors, termed Generation IV, has very attractive features -- its superior safety characteristics, high thermal efficiency, and fuel cycle sustainability. A key element of the Generation IV designs is the improvement in safety, which in turn requires improvements in safety system performance and reliability, as well as a reduction in initiating event frequencies. This study compares the response of the systems important to safety in the CANDU-Supercritical Water Reactor to those of the generic CANDU under a main steamline break accident and loss of forced circulation events -- to quantify the improvements in safety for the pre-conceptual CANDU SCWR design. Probabilistic safety analysis is the tool used in this study to test the behavior of the pre- conceptual design during these events. (author)

  20. [Safety assessment of nanomaterials for development of nano-cosmetics].

    Science.gov (United States)

    Yoshida, Tokuyuki; Yoshioka, Yasuo; Tsutsumi, Yasuo

    2012-01-01

    A diverse array of nanomaterials (NMs), such as amorphous nanosilica (nSP), carbon nanotubes and titanium dioxide, has become widespread in use due to the development of nanotechnology. NMs are already being applied in universal fields because they have unique physicochemical properties. On the other hands, the safety of NMs has not been well assessed, because NMs have been considered as safe as common larger sized materials which are known not to be absorbed by the body. Because NMs have the potential to improve the quality of human life, it is essential to ensure the safety of NMs and provide information for designing safer NMs. In this regard, we studied the biological distribution and hazard identification of nSP following dermal administration, because nSP is used NMs in the cosmetics field. In the future, our study would help to set the no observed adverse effect level (NOAEL) and acceptable daily intake (ADI), and be useful information for the safety/hazard assessment and evaluation.

  1. 2005 dossier: granite. Tome: safety analysis of the geologic disposal; Dossier 2005: granite. Tome analyse de surete du stockage geologique

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2005-07-01

    This document makes a status of the researches carried out by the French national agency of radioactive wastes (ANDRA) about the safety aspects of the geologic disposal of high-level and long-lived (HLLL) radioactive wastes in granite formations. Content: 1 - safety approach: context and general goal, references, design approach by safety functions, safety approach during the construction-exploitation-observation-closure phase, safety analysis during the post-closure phase; 2 - general description: HLLL wastes, granitic environment, general structure of the architecture of a disposal facility; 3 - safety functions and disposal design: general context, safety functions of the long-term disposal, design dispositions retained to answer the functions; 4 - operational safety: people's protection, radiological risks during exploitation, risk analysis in accident situation; 5 - qualitative safety analysis: methodology, main results of the analysis of the features, events and processes (FEP) database; 6 - disposal efficiency evaluation during post-closure phase: calculation models, calculation tools used for the modeling of radionuclides transport, calculation results and main lessons. (J.S.)

  2. Developing Measures for Assessing the Causality of Safety Culture in a Petrochemical Industry

    International Nuclear Information System (INIS)

    Wu, T.-C.; Lin, C.-H.; Shiau, S.-Y.

    2009-01-01

    This paper discusses safety culture in the petrochemical sector and the causes and consequences of safety culture. A sample of 520 responses selected by simple random sampling completed questionnaires for this survey, the return rate was 86.75%. The research instrument comprises four sections: basic information, the safety leadership scale (SLS), the safety climate scale (SCS), and the safety performance scale (SPS). SPSS 12.0, a statistical software package, was used for item analysis, validity analysis, and reliability analysis. Exploratory factor analysis indicated that (1) SLS abstracted three factors such as safety caring, safety controlling, and safety coaching; (2) SCS comprised three factors such as emergency response, safety commitment, and risk perception; and (3) SPS was composed of accident investigation, safety training, safety inspections, and safety motivation. We conclude that the SLS, SCS, and SPS developed in this paper have good construct validity and internal consistency and can serve as the basis for future research.

  3. 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?

  4. Multilevel Safety Climate and Safety Performance in the Construction Industry: Development and Validation of a Top-Down Mechanism

    Directory of Open Access Journals (Sweden)

    Ran Gao

    2016-11-01

    Full Text Available The character of construction projects exposes front-line workers to dangers and accidents. Safety climate has been confirmed to be a predictor of safety performance in the construction industry. This study aims to explore the underlying mechanisms of the relationship between multilevel safety climate and safety performance. An integrated model was developed to study how particular safety climate factors of one level affect those of other levels, and then affect safety performance from the top down. A questionnaire survey was administered on six construction sites in Vietnam. A total of 1030 valid questionnaires were collected from this survey. Approximately half of the data were used to conduct exploratory factor analysis (EFA and the remaining data were submitted to structural equation modeling (SEM. Top management commitment (TMC and supervisors’ expectation (SE were identified as factors to represent organizational safety climate (OSC and supervisor safety climate (SSC, respectively, and coworkers’ caring and communication (CCC and coworkers’ role models (CRM were identified as factors to denote coworker safety climate (CSC. SEM results show that OSC factor is positively related to SSC factor and CSC factors significantly. SSC factor could partially mediate the relationship between OSC factor and CSC factors, as well as the relationship between OSC factor and safety performance. CSC factors partially mediate the relationship between OSC factor and safety performance, and the relationship between SSC factor and safety performance. The findings imply that a positive safety culture should be established both at the organizational level and the group level. Efforts from all top management, supervisors, and coworkers should be provided to improve safety performance in the construction industry.

  5. Multilevel Safety Climate and Safety Performance in the Construction Industry: Development and Validation of a Top-Down Mechanism.

    Science.gov (United States)

    Gao, Ran; Chan, Albert P C; Utama, Wahyudi P; Zahoor, Hafiz

    2016-11-08

    The character of construction projects exposes front-line workers to dangers and accidents. Safety climate has been confirmed to be a predictor of safety performance in the construction industry. This study aims to explore the underlying mechanisms of the relationship between multilevel safety climate and safety performance. An integrated model was developed to study how particular safety climate factors of one level affect those of other levels, and then affect safety performance from the top down. A questionnaire survey was administered on six construction sites in Vietnam. A total of 1030 valid questionnaires were collected from this survey. Approximately half of the data were used to conduct exploratory factor analysis (EFA) and the remaining data were submitted to structural equation modeling (SEM). Top management commitment (TMC) and supervisors' expectation (SE) were identified as factors to represent organizational safety climate (OSC) and supervisor safety climate (SSC), respectively, and coworkers' caring and communication (CCC) and coworkers' role models (CRM) were identified as factors to denote coworker safety climate (CSC). SEM results show that OSC factor is positively related to SSC factor and CSC factors significantly. SSC factor could partially mediate the relationship between OSC factor and CSC factors, as well as the relationship between OSC factor and safety performance. CSC factors partially mediate the relationship between OSC factor and safety performance, and the relationship between SSC factor and safety performance. The findings imply that a positive safety culture should be established both at the organizational level and the group level. Efforts from all top management, supervisors, and coworkers should be provided to improve safety performance in the construction industry.

  6. Managers’ Perceptions of Organizational Safety: Implication for the Development of Safety Culture

    OpenAIRE

    Jarvis, Marina; Virovere, Anu; Tint, Piia

    2014-01-01

    A key feature of a company’s safety culture is shared perceptions between managers and employees concerning the importance of safety. The purpose of the study was to assess senior managers’ perceptions of the current safety culture in Estonian SMEs from different branches of industry, based on a sample of 463 senior managers. The relationships of different aspects of safety culture and safety management systems, senior managers’ and employees’ perceptions and attitudes to safety were explored...

  7. Development and application of the San Onofre safety monitor

    International Nuclear Information System (INIS)

    Hook, Thomas G.; Lee, Roger J.; Morgan, Thomas A.

    2004-01-01

    Halliburton NUS Corporation (NUS) has developed a risk-based configuration management software tool for use at Southern California Edison's San Onofre Nuclear Generating Station. The software, called the Safety Monitor, calculates an estimate of current plant core damage risk based upon the plant's current operating configuration (e.g., equipment operability, system operating alignments). All data is entered and displayed in a format easily understood by plant personnel. The plant hopes to use this tool to ensure that risk is minimized during plant operations and to identify situations in which current Technical Specifications can be optimized. Plant configuration data and out-of-service time data is also automatically collected. (author)

  8. Development of IAEA description of passive safety and subsequent thoughts

    International Nuclear Information System (INIS)

    Lang, P.M.

    1996-01-01

    The description of passive components and systems published by the IAEA in its TECDOC-626 was developed in the course of a Technical Committee Meeting held in Sweden and two subsequent Consultants Meetings held in Vienna. This description is reviewed and discussed in terms of the philosophies behind it, alternatives considered, problems encountered, and conclusions drawn. Also discussed is an Appendix to the TECDOC, which illustrates the spectrum of possibilities from passive to active by describing four typical categories of passivity. Subsequent thoughts on passive safety include a discussion of its advantages and disadvantages, concluding with a summary of current views and problems with it. (author). 8 refs

  9. Future generations of CANDU: advantages and development with passive safety

    International Nuclear Information System (INIS)

    Duffey, R. B.

    2006-01-01

    Atomic Energy of Canada Limited (AECL) advances water reactor and CANDLT technology using an evolutionary development strategy. This strategy ensures that innovations are based firmly on current experience and keeps our development programs focused on one reactor concept, reducing risks, development costs, and product development cycle times. It also assures our customers that our products will never become obsolete or unsupported, and the continuous line of water reactor development is secure and supported into the future. Using the channel reactor advantage of modularity, the subdivided core has the advantage of passive safety by heat removal to the low- pressure moderator. With continuous improvements, the Advanced CANDU Reactor TM (ACR-1000TM) concept will likely remain highly competitive for a number of years and leads naturally to the next phase of CANDU development, namely the Generation IV CANDU -SCWR concept. This is conventional water technology, since supercritical boilers and turbines have been operating for some time in coal-fired power plants. Significant cost, safety, and performance advantages would result from the CANDU-SCWR concept, plus the flexibility of a range of plant sizes suitable for both small and large electric grids, and the ability for co-generation of electric power, process heat, and hydrogen. In CANDU-SCWR, novel developments are included in the primary circuit layout and channel design. The R and D in Canada is integrated with the Generation IV international Forum (GIF) plans, and has started on examining replaceable insulating liners that would ensure channel life, and on providing completely passive reactor decay heat removal directly to the moderator heat sink without forced cooling. In the interests of sustainability, hydrogen production by a CANDU- SCWR is also be included as part of the system requirements, where the methods for hydrogen production will depend on the outlet temperature of the reactor

  10. Role of management in the development of safety culture at the operating organization

    International Nuclear Information System (INIS)

    Zhong, W.

    1997-01-01

    Role of management in the development of safety culture at the operating organization to offer practical suggestions to assist in the development or improvement of a progressive safety culture. 2 figs

  11. SAFETY

    CERN Document Server

    Niels Dupont

    2013-01-01

    CERN Safety rules and Radiation Protection at CMS The CERN Safety rules are defined by the Occupational Health & Safety and Environmental Protection Unit (HSE Unit), CERN’s institutional authority and central Safety organ attached to the Director General. In particular the Radiation Protection group (DGS-RP1) ensures that personnel on the CERN sites and the public are protected from potentially harmful effects of ionising radiation linked to CERN activities. The RP Group fulfils its mandate in collaboration with the CERN departments owning or operating sources of ionising radiation and having the responsibility for Radiation Safety of these sources. The specific responsibilities concerning "Radiation Safety" and "Radiation Protection" are delegated as follows: Radiation Safety is the responsibility of every CERN Department owning radiation sources or using radiation sources put at its disposition. These Departments are in charge of implementing the requi...

  12. A Server-Client-Based Graphical Development Environment for Physics Analyses (VISPA)

    Science.gov (United States)

    Bretz, H.-P.; Erdmann, M.; Fischer, R.; Hinzmann, A.; Klingebiel, D.; Komm, M.; Müller, G.; Rieger, M.; Steffens, J.; Steggemann, J.; and, M. Urban; Winchen, T.

    2012-12-01

    The Visual Physics Analysis (VISPA) project provides a graphical development environment for data analysis. It addresses the typical development cycle of (re-)designing, executing, and verifying an analysis. We present the new server-client-based web application of the VISPA project to perform physics analyses via a standard internet browser. This enables individual scientists to work with a large variety of devices including touch screens, and teams of scientists to share, develop, and execute analyses on a server via the web interface.

  13. A Server-Client-Based Graphical Development Environment for Physics Analyses (VISPA)

    International Nuclear Information System (INIS)

    Bretz, H-P; Erdmann, M; Fischer, R; Hinzmann, A; Klingebiel, D; Komm, M; Müller, G; Rieger, M; Steffens, J; Steggemann, J; Urban, M; Winchen, T

    2012-01-01

    The Visual Physics Analysis (VISPA) project provides a graphical development environment for data analysis. It addresses the typical development cycle of (re-)designing, executing, and verifying an analysis. We present the new server-client-based web application of the VISPA project to perform physics analyses via a standard internet browser. This enables individual scientists to work with a large variety of devices including touch screens, and teams of scientists to share, develop, and execute analyses on a server via the web interface.

  14. HTGR Dust Safety Issues and Needs for Research and Development

    Energy Technology Data Exchange (ETDEWEB)

    Paul W. Humrickhouse

    2011-06-01

    This report presents a summary of high temperature gas-cooled reactor dust safety issues. It draws upon a literature review and the proceedings of the Very High Temperature Reactor Dust Assessment Meeting held in Rockville, MD in March 2011 to identify and prioritize the phenomena and issues that characterize the effect of carbonaceous dust on high temperature reactor safety. It reflects the work and input of approximately 40 participants from the U.S. Department of Energy and its National Labs, the U.S. Nuclear Regulatory Commission, industry, academia, and international nuclear research organizations on the topics of dust generation and characterization, transport, fission product interactions, and chemical reactions. The meeting was organized by the Idaho National Laboratory under the auspices of the Next Generation Nuclear Plant Project, with support from the U.S. Nuclear Regulatory Commission. Information gleaned from the report and related meetings will be used to enhance the fuel, graphite, and methods technical program plans that guide research and development under the Next Generation Nuclear Plant Project. Based on meeting discussions and presentations, major research and development needs include: generating adsorption isotherms for fission products that display an affinity for dust, investigating the formation and properties of carbonaceous crust on the inside of high temperature reactor coolant pipes, and confirming the predominant source of dust as abrasion between fuel spheres and the fuel handling system.

  15. Development of a safety analysis code for molten salt reactors

    International Nuclear Information System (INIS)

    Zhang Dalin; Qiu Suizheng; Su Guanghui

    2009-01-01

    The molten salt reactor (MSR) well suited to fulfill the criteria defined by the Generation IV International Forum (GIF) is presently revisited all around the world because of different attractive features of current renewed relevance. The MSRs are characterized by using the fluid-fuel, so that their technologies are fundamentally different from those used in the conventional solid-fuel reactors. In this work, in particular, the attention is focused on the safety characteristic analysis of the MSRs, in which a point kinetic model considering the flow effects of the fuel salt is established for the MSRs and calculated by developing a microcomputer code coupling with a simplified heat transfer model in the core. The founded models and developed code are applied to analyze the safety characteristics of the molten salt actinide recycler and transmuter system (MOSART) by simulating three types of basic transient conditions including the unprotected loss of flow, unprotected overcooling accident and unprotected transient overpower. Some reasonable results are obtained for the MOSART, which show that the MOSART conceptual design is an inherently stable reactor design. The present study provides some valuable information for the research and design of the new generation MSRs.

  16. Development of safety culture - A Chinese traditional cultural perspective

    International Nuclear Information System (INIS)

    Zhou Weihong . E-mail zhouwh@lanps.com

    2002-01-01

    Living in a social community, the culture of an enterprise is certainly under the influence of that society. Safety culture of nuclear utilities is the core of the enterprise culture. As a formal expression as defined in INSAG 3 and 4 by IAEA, it as a matter of fact originated from the summing up of the experiences of western nuclear industry, particularly after such epoch-making accidents of Three Miles Island and Chernobyl. In view of the geographical culture theory, whether or not this conception of western industrial culture will be absorbed and assimilated by Chinese Nuclear Industry is a challenging issue. This is because, on the one hand, Nuclear Power is comparatively speaking a newly developing industry in China and, on the other hand, China has enjoyed an uninterrupted history of traditional culture over five thousand years. In other words, whether the new and alien values will conflict with or be constructively assimilated by our traditional mindset is a critical question to be answered in any development program of safety culture. (author)

  17. On groundwater flow modelling in safety analyses of spent fuel disposal. A comparative study with emphasis on boundary conditions

    Energy Technology Data Exchange (ETDEWEB)

    Jussila, P

    1999-11-01

    Modelling groundwater flow is an essential part of the safety assessment of spent fuel disposal because moving groundwater makes a physical connection between a geological repository and the biosphere. Some of the common approaches to model groundwater flow in bedrock are equivalent porous continuum (EC), stochastic continuum and various fracture network concepts. The actual flow system is complex and measuring data are limited. Multiple distinct approaches and models, alternative scenarios as well as calibration and sensitivity analyses are used to give confidence on the results of the calculations. The correctness and orders of magnitude of results of such complex research can be assessed by comparing them to the results of simplified and robust approaches. The first part of this study is a survey of the objects, contents and methods of the groundwater flow modelling performed in the safety assessment of the spent fuel disposal in Finland and Sweden. The most apparent difference of the Swedish studies compared to the Finnish ones is the approach of using more different models, which is enabled by the more resources available in Sweden. The results of more comprehensive approaches provided by international co-operation are very useful to give perspective to the results obtained in Finland. In the second part of this study, the influence of boundary conditions on the flow fields of a simple 2D model is examined. The assumptions and simplifications in this approach include e.g. the following: (1) the EC model is used, in which the 2-dimensional domain is considered a continuum of equivalent properties without fractures present, (2) the calculations are done for stationary fields, without sources or sinks present in the domain and with a constant density of the groundwater, (3) the repository is represented by an isotropic plate, the hydraulic conductivity of which is given fictitious values, (4) the hydraulic conductivity of rock is supposed to have an exponential

  18. Use and development of probabilistic safety assessment - CSNI WGRISK

    International Nuclear Information System (INIS)

    Siu, Nathan; Monninger, John; Gomez-Cobo, Ana; Kao, Tsu-Mu; Schoen, Gerhard; Gunsell, Lars; Nyman, Ralph; Jelinek, Tomas; Hultquist, Goeran; Rapp, Anders; Eriksson, Stefan; Lantaron, Alfredo; Vojnovic, Djordje; Husarcek, Jan; Kovacs, Zoltan; Versteeg, M.F.; Lopez Morones, Ramon; Lee, Chang-Ju; Fukuda, Mamoru; Burgazzi, Luciano; Caporali, Rino; RoeWEKAMP, Marina; MACSUGA, Geza; Bareith, Attila; Lanore, J.M.; Sorel, Vincent; Virolainen, Reino; Patrik, Milan; Mlady, Ondrej; Raducu, Gheorghe; De Gelder, Pieter; Hendrickx, Isabelle; Lanore, Jeanne-Marie; Murphy, Joseph A.; Shepherd, Charles; Pyy, Pekka T.; Mauny, Elisabeth

    2007-01-01

    The CSNI WGRISK produced a report in July 2002 on 'The Use and Development of Probabilistic Safety Assessment in NEA Member Countries'. This provides a description of the PSA programmes in the member countries at the time that the report was produced. However, there have been significant developments in PSA since 2002. Consequently, a decision was made at the WGRISK meeting in October 2005 to produce an updated version of the report. The aim was to produce an updated, stand alone version of the report that presents an analysis of the position on the use and development of PSA in the WGRISK member countries as of spring 2006. A detailed questionnaire was circulated to WGRISK members and to the IAEA to ascertain the state of the art in PSA use and development at the end of 2006. Detailed responses were prepared by 20 countries totalling several hundred pages of information. After first compilation of information, an updating round was organized by showing to the countries all the answers and the summary made of them by a small group of experts. The process led to some clarifications and more consistency in the report. The collected information was finally analyzed and summarized to reach the conclusions presented in this report. The set of section headings in the report is as follows: Executive summary. 1. Introduction. 2. PSA Framework and Environment. 3. Numerical Safety Criteria. 4. PSA Standards and Guidance. 5. Status and Scope of PSA Programmes. 6. PSA Methodology and Data. 7. PSA Applications. 8. Results and Insights from the PSAs. 9. Future Developments. Appendix A: Overview of the Status of PSA Programmes. Appendix B: Contact information. Appendix C: Questionnaire and Guidance to authors

  19. Status of Physics and Safety Analyses for the Liquid-Salt-Cooled Very High-Temperature Reactor (LS-VHTR)

    Energy Technology Data Exchange (ETDEWEB)

    Ingersoll, DT

    2005-12-15

    A study has been completed to develop a new baseline core design for the liquid-salt-cooled very high-temperature reactor (LS-VHTR) that is better optimized for liquid coolant and that satisfies the top-level operational and safety targets, including strong passive safety performance, acceptable fuel cycle parameters, and favorable core reactivity response to coolant voiding. Three organizations participated in the study: Oak Ridge National Laboratory (ORNL), Idaho National Laboratory (INL), and Argonne National Laboratory (ANL). Although the intent was to generate a new reference LS-VHTR core design, the emphasis was on performing parametric studies of the many variables that constitute a design. The results of the parametric studies not only provide the basis for choosing the optimum balance of design options, they also provide a valuable understanding of the fundamental behavior of the core, which will be the basis of future design trade-off studies. A new 2400-MW(t) baseline design was established that consists of a cylindrical, nonannular core cooled by liquid {sup 7}Li{sub 2}BeF{sub 4} (Flibe) salt. The inlet and outlet coolant temperatures were decreased by 50 C, and the coolant channel diameter was increased to help lower the maximum fuel and vessel temperatures. An 18-month fuel cycle length with 156 GWD/t burnup was achieved with a two-batch shuffling scheme, while maintaining a core power density of 10 MW/m{sup 3} using graphite-coated uranium oxicarbide particle fuel enriched to 15% {sup 235}U and assuming a 25 vol-% packing of the coated particles in the fuel compacts. The revised design appears to have excellent steady-state and transient performance. The previous concern regarding the core's response to coolant voiding has been resolved for the case of Flibe coolant by increasing the coolant channel diameter and the fuel loading. Also, the LSVHTR has a strong decay heat removal performance and appears capable of surviving a loss of forced

  20. Experience of RIA safety analyses performance for NPP Temelin core arranged with TVSA-T fuel assemblies

    International Nuclear Information System (INIS)

    Kryukov, S.A.; Lizorkin, M.P.

    2010-01-01

    The contents of the presentation are as follows: 1. Definition of categories for initiating events; 2. Acceptance criteria for safety assessment; 3. Main aspects of safety assessment methodology; 4. Main stages of calculation analysis; 5. Interface with other parts of the core design; 6. Codes used for calculation; 6.1 Main performances of code package TIGR-1; 6.2 Main performances of code BIPR-7A; 7. TIGR-1 accounting of design margins in calculation of fuel rod powers; 8. Peculiar features of Instrumentation and Control System for Temelin NPP; 9. Calculations; 10. Checklist of margin data important for reload safety assessment. (P.A.)

  1. Safety

    CERN Multimedia

    2003-01-01

    Please note that the safety codes A9, A10 AND A11 (ex annexes of SAPOCO/42) entitled respectively "Safety responsibilities in the divisions" "The safety policy committee (SAPOCO) and safety officers' committees" and "Administrative procedure following a serious accident or incident" are available on the web at the following URLs: Code A9: http://edms.cern.ch/document/337016/LAST_RELEASED Code A10: http://edms.cern.ch/document/337019/LAST_RELEASED Code A11: http://edms.cern.ch/document/337026/LAST_RELEASED Paper copies can also be obtained from the TIS divisional secretariat, e-mail: tis.secretariat@cern.ch. TIS Secretariat

  2. Analysis of international approaches which are used at development of theoperational safety performance indicators

    International Nuclear Information System (INIS)

    Lyigots'kij, O.Yi.; Nosovs'kij, A.V.; Chemeris, Yi.O.

    2009-01-01

    Description of international approaches and experience of the use of theoperational safety performance indicators system is provided for estimationof current status and making a decision on corrections in the operationpractice. The state of development of the operational safety performanceindicators system by the operating organization is overviewed. Thepossibility of application of international approaches during development ofthe integral safety performance indicators system is analyzed. Aims and tasksof future researches are formulated in relation to development of theintegral safety performance indicators system.

  3. The safety regulation of small-scale coal mines in China: Analysing the interests and influences of stakeholders

    International Nuclear Information System (INIS)

    Song, Xiaoqian; Mu, Xiaoyi

    2013-01-01

    Small scale coal mines (SCMs) have played an important role in China’s energy supply. At the same time, they also suffer from many social, economic, environmental, and safety problems. The Chinese government has made considerable efforts to strengthen the safety regulation of the coal mining industry. Yet, few of these efforts have proven to be very effective. This paper analyzes the interests and influences of key stakeholders in the safety regulation of SCMs, which includes the safety regulator, the local government, the mine owner, and mineworkers. We argue that the effective regulation of coal mine safety must both engage and empower mineworkers. - Highlights: ► Small scale coal mines have played an important role in China's energy supply. ► We analyze the interests and influences of key stakeholders in the safety regulation of small coal mines. ► The mineworkers have the strongest interest but least influence. ► An effective regulation must engage the mineworkers, organize, and empower them.

  4. Assessing the general safety and tolerability of vildagliptin: value of pooled analyses from a large safety database versus evaluation of individual studies

    Directory of Open Access Journals (Sweden)

    Schweizer A

    2011-02-01

    Full Text Available Anja Schweizer1, Sylvie Dejager2, James E Foley3, Wolfgang Kothny31Novartis Pharma AG, Basel, Switzerland; 2Novartis Pharma SAS, Rueil-Malmaison, France; 3Novartis Pharmaceuticals Corporation, East Hanover, NJ, USAAim: Analyzing safety aspects of a drug from individual studies can lead to difficult-to-interpret results. The aim of this paper is therefore to assess the general safety and tolerability, including incidences of the most common adverse events (AEs, of vildagliptin based on a large pooled database of Phase II and III clinical trials.Methods: Safety data were pooled from 38 studies of ≥12 to ≥104 weeks' duration. AE profiles of vildagliptin (50 mg bid; N = 6116 were evaluated relative to a pool of comparators (placebo and active comparators; N = 6210. Absolute incidence rates were calculated for all AEs, serious AEs (SAEs, discontinuations due to AEs, and deaths.Results: Overall AEs, SAEs, discontinuations due to AEs, and deaths were all reported with a similar frequency in patients receiving vildagliptin (69.1%, 8.9%, 5.7%, and 0.4%, respectively and patients receiving comparators (69.0%, 9.0%, 6.4%, and 0.4%, respectively, whereas drug-related AEs were seen with a lower frequency in vildagliptin-treated patients (15.7% vs 21.7% with comparators. The incidences of the most commonly reported specific AEs were also similar between vildagliptin and comparators, except for increased incidences of hypoglycemia, tremor, and hyperhidrosis in the comparator group related to the use of sulfonylureas.Conclusions: The present pooled analysis shows that vildagliptin was overall well tolerated in clinical trials of up to >2 years in duration. The data further emphasize the value of a pooled analysis from a large safety database versus assessing safety and tolerability from individual studies.Keywords: type 2 diabetes, dipeptidyl peptidase-4, edema, safety, vildagliptin

  5. Developing Expert Teams with a Strong Safety Culture

    Science.gov (United States)

    Rogers, David G.

    2010-01-01

    Would you like to lead a world renowned team that draws out all the talents and expertise of its members and consistently out performs all others in the industry? Ever wonder why so many organizations fail to truly learn from past mistakes only to repeat the same ones at a later date? Are you a program/project manager or team member in a high-risk organization where the decisions made often carry the highest of consequences? Leadership, communication, team building, critical decision-making and continuous team improvement skills and behaviors are mere talking points without the attitudes, commitment and strategies necessary to make them the very fabric of a team. Developing Expert Teams with a Strong Safety Culture, will provide you with proven knowledge and strategies to take your team soaring to heights you may have not thought possible. A myriad of teams have applied these strategies and techniques within their organization team environments: military and commercial aviation, astronaut flight crews, Shuttle flight controllers, members of the Space Shuttle Program Mission Management Team, air traffic controllers, nuclear power control teams, surgical teams, and the fire service report having spectacular success. Many industry leaders are beginning to realize that although the circumstances and environments of these teams may differ greatly to their own, the core elements, governing principles and dynamics involved in managing and building a stellar safety conscious team remain identical.

  6. Modeling the Relationship between Safety Climate and Safety Performance in a Developing Construction Industry: A Cross-Cultural Validation Study.

    Science.gov (United States)

    Zahoor, Hafiz; Chan, Albert P C; Utama, Wahyudi P; Gao, Ran; Zafar, Irfan

    2017-03-28

    This study attempts to validate a safety performance (SP) measurement model in the cross-cultural setting of a developing country. In addition, it highlights the variations in investigating the relationship between safety climate (SC) factors and SP indicators. The data were collected from forty under-construction multi-storey building projects in Pakistan. Based on the results of exploratory factor analysis, a SP measurement model was hypothesized. It was tested and validated by conducting confirmatory factor analysis on calibration and validation sub-samples respectively. The study confirmed the significant positive impact of SC on safety compliance and safety participation , and negative impact on number of self-reported accidents/injuries . However, number of near-misses could not be retained in the final SP model because it attained a lower standardized path coefficient value. Moreover, instead of safety participation , safety compliance established a stronger impact on SP. The study uncovered safety enforcement and promotion as a novel SC factor, whereas safety rules and work practices was identified as the most neglected factor. The study contributed to the body of knowledge by unveiling the deviations in existing dimensions of SC and SP. The refined model is expected to concisely measure the SP in the Pakistani construction industry, however, caution must be exercised while generalizing the study results to other developing countries.

  7. Development of a safety decision-making scenario to measure worker safety in agriculture.

    Science.gov (United States)

    Mosher, G A; Keren, N; Freeman, S A; Hurburgh, C R

    2014-04-01

    Human factors play an important role in the management of occupational safety, especially in high-hazard workplaces such as commercial grain-handling facilities. Employee decision-making patterns represent an essential component of the safety system within a work environment. This research describes the process used to create a safety decision-making scenario to measure the process that grain-handling employees used to make choices in a safety-related work task. A sample of 160 employees completed safety decision-making simulations based on a hypothetical but realistic scenario in a grain-handling environment. Their choices and the information they used to make their choices were recorded. Although the employees emphasized safety information in their decision-making process, not all of their choices were safe choices. Factors influencing their choices are discussed, and implications for industry, management, and workers are shared.

  8. Technical co-operation for nuclear safety in developing countries

    International Nuclear Information System (INIS)

    Flakus, F.N.; Giuliani, P.

    1984-01-01

    The Agency's programme on technical co-operation for nuclear safety is, largely, responsive in character and the Agency's response is tailored to needs identified by developing countries. However, the Agency's assistance alone is not sufficient: technical co-operation can only be successful and is most effective when there is also a strong input from the counterpart body participating in a particular project. The commitment of national governments is fundamental to success. Technical co-operation is most fruitful if the Agency's assistance capabilities and the recipient country's co-operation capabilities match. Co-operation activities mostly take the form of single projects hosted by individual institutions within a single country; regional and inter-regional projects are also important

  9. Development, application and licensing of FPGA based safety systems

    Energy Technology Data Exchange (ETDEWEB)

    Tuite, T.C.; Carvajal, J.V., E-mail: tuitetc@westinghouse.com, E-mail: carvajjv@westinghouse.com [Westinghouse Electric Company, Pennsylvania, (United States)

    2015-07-01

    Westinghouse has developed the Advanced Logic System (ALS) platform. The ALS platform was recently approved by the US NRC. In addition, ALS was successfully installed and declared operational as a Thermocouple/Core Cooling Monitor upgrade at the Wolf Creek Generating Station. The ALS has also been installed at the AP1000 Sanmen and Haiyang unit sites as the Diverse Actuation System. The ALS platform is based on FPGA technology. FPGA safety system designs are simpler than comparable CPU based system designs in that they do not require an Operating System or instruction set. The ALS platform provides inherent diversity in redundant cores and diverse core designs. In addition, the ALS provides extensive self-testing and diagnostics which allows for extension of plant surveillance intervals. (author)

  10. MRI evaluation and safety in the developing brain.

    Science.gov (United States)

    Tocchio, Shannon; Kline-Fath, Beth; Kanal, Emanuel; Schmithorst, Vincent J; Panigrahy, Ashok

    2015-03-01

    Magnetic resonance imaging (MRI) evaluation of the developing brain has dramatically increased over the last decade. Faster acquisitions and the development of advanced MRI sequences, such as magnetic resonance spectroscopy (MRS), diffusion tensor imaging (DTI), perfusion imaging, functional MR imaging (fMRI), and susceptibility-weighted imaging (SWI), as well as the use of higher magnetic field strengths has made MRI an invaluable tool for detailed evaluation of the developing brain. This article will provide an overview of the use and challenges associated with 1.5-T and 3-T static magnetic fields for evaluation of the developing brain. This review will also summarize the advantages, clinical challenges, and safety concerns specifically related to MRI in the fetus and newborn, including the implications of increased magnetic field strength, logistics related to transporting and monitoring of neonates during scanning, and sedation considerations, and a discussion of current technologies such as MRI conditional neonatal incubators and dedicated small-foot print neonatal intensive care unit (NICU) scanners. Copyright © 2015 Elsevier Inc. All rights reserved.

  11. MRI Evaluation and Safety in the Developing Brain

    Science.gov (United States)

    Tocchio, Shannon; Kline-Fath, Beth; Kanal, Emanuel; Schmithorst, Vincent J.; Panigrahy, Ashok

    2015-01-01

    Magnetic resonance imaging (MRI) evaluation of the developing brain has dramatically increased over the last decade. Faster acquisitions and the development of advanced MRI sequences such as magnetic resonance spectroscopy (MRS), diffusion tensor imaging (DTI), perfusion imaging, functional MR imaging (fMRI), and susceptibility weighted imaging (SWI), as well as the use of higher magnetic field strengths has made MRI an invaluable tool for detailed evaluation of the developing brain. This article will provide an overview of the use and challenges associated with 1.5T and 3T static magnetic fields for evaluation of the developing brain. This review will also summarize the advantages, clinical challenges and safety concerns specifically related to MRI in the fetus and newborn, including the implications of increased magnetic field strength, logistics related to transporting and monitoring of neonates during scanning, sedation considerations and a discussion of current technologies such as MRI-conditional neonatal incubators and dedicated small-foot print neonatal intensive care unit (NICU) scanners. PMID:25743582

  12. WP-Cave - assessment of feasibility, safety and development potential

    International Nuclear Information System (INIS)

    1989-09-01

    According to SKB R and D-programme 1986, alternative disposal methods will be investigated to provide a basis for selecting a site and a repository system for the Swedish spent nuclear fuel. The present report is a comparison between the WP-Cave and the reference concept KBS-3. The comparison has resulted in the following conclusions: - Both concepts are judged to be able to provide adequate safety. - A utilization of the potential of the WP-Cave requires, however, extensive development in areas where the current state of knowledge and available data are incomplete. - The higher temperatures in the WP-Cave lead to greater uncertainty as to long-term performance. Reducing this uncertainty would require many yaers of research and substantial resources. - Both repositories, including the barriers they incorporate, could be built with a normal adaption of available technology. -It is not possible to say today whether it would be simpler to find suitable sites for one design or the other. - The WP-Cave is considerably more expensive. A future research direction based on a concentrated emplacement of spent fuel along the lines of the WP-Cave is therefore judged to entail greater uncertainty as regards the possibilities of achieving acceptable safety and to require greater resources for research and development, at the same time as the costs of building the repository would be higher. The studies of the WP-Cave as an integral system should therfore be discontinued. Certain barrier designs in the WP-Cave could also be utulized in repository designs with lower temperature, for example the reduction potential of the steel canisters and the hydraulic cage's diversion of groundwater. Studies within these areas are being conducted within SKB and should continue

  13. Web-Based Assessment and Test Analyses (WATA) System: Development and Evaluation

    Science.gov (United States)

    Wang, T. H.; Wang, K. H.; Wang, W. L.; Huang, S. C.; Chen, S. Y.

    2004-01-01

    This study introduces the development of a Web-based assessment system, the Web-based Assessment and Test Analyses (WATA) system, and examines its impacts on teacher education. The WATA system is a follow-on system, which applies the Triple-A Model (assembling, administering, and appraising). Its functions include (1) an engine for teachers to…

  14. Neutronic analyses and tools development efforts in the European DEMO programme

    Energy Technology Data Exchange (ETDEWEB)

    Fischer, U., E-mail: ulrich.fischer@kit.edu [Association KIT-Euratom, Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Bachmann, C. [European Fusion Development Agreement (EFDA), Garching (Germany); Bienkowska, B. [Association IPPLM-Euratom, IPPLM Warsaw/INP Krakow (Poland); Catalan, J.P. [Universidad Nacional de Educación a Distancia (UNED), Madrid (Spain); Drozdowicz, K.; Dworak, D. [Association IPPLM-Euratom, IPPLM Warsaw/INP Krakow (Poland); Leichtle, D. [Association KIT-Euratom, Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Fusion for Energy (F4E), Barcelona (Spain); Lengar, I. [MESCS-JSI, Ljubljana (Slovenia); Jaboulay, J.-C. [CEA, DEN, Saclay, DM2S, SERMA, F-91191 Gif-sur-Yvette (France); Lu, L. [Association KIT-Euratom, Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Moro, F. [Associazione ENEA-Euratom, ENEA Fusion Division, Frascati (Italy); Mota, F. [Centro de Investigaciones Energéticas, Medioambientales y Tecnológicas (CIEMAT), Madrid (Spain); Sanz, J. [Universidad Nacional de Educación a Distancia (UNED), Madrid (Spain); Szieberth, M. [Budapest University of Technology and Economics (BME), Budapest (Hungary); Palermo, I. [Centro de Investigaciones Energéticas, Medioambientales y Tecnológicas (CIEMAT), Madrid (Spain); Pampin, R. [Fusion for Energy (F4E), Barcelona (Spain); Porton, M. [Euratom/CCFE Fusion Association, Culham Science Centre for Fusion Energy (CCFE), Culham (United Kingdom); Pereslavtsev, P. [Association KIT-Euratom, Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Ogando, F. [Universidad Nacional de Educación a Distancia (UNED), Madrid (Spain); Rovni, I. [Budapest University of Technology and Economics (BME), Budapest (Hungary); and others

    2014-10-15

    Highlights: •Evaluation of neutronic tools for application to DEMO nuclear analyses. •Generation of a DEMO model for nuclear analyses based on MC calculations. •Nuclear analyses of the DEMO reactor equipped with a HCLL-type blanket. -- Abstract: The European Fusion Development Agreement (EFDA) recently launched a programme on Power Plant Physics and Technology (PPPT) with the aim to develop a conceptual design of a fusion demonstration reactor (DEMO) addressing key technology and physics issues. A dedicated part of the PPPT programme is devoted to the neutronics which, among others, has to define and verify requirements and boundary conditions for the DEMO systems. The quality of the provided data depends on the capabilities and the reliability of the computational tools. Accordingly, the PPPT activities in the area of neutronics include both DEMO nuclear analyses and development efforts on neutronic tools including their verification and validation. This paper reports on first neutronics studies performed for DEMO, and on the evaluation and further development of neutronic tools.

  15. Spatial analyses of the urban village development process in Shenzhen, China

    NARCIS (Netherlands)

    Hao, Pu; Geertman, S.; Hooimeijer, P.; Sliuzas, R.V.

    2013-01-01

    Urban villages are widespread in many Chinese cities, providing affordable and accessible housing for rural migrants. These urban villages develop rapidly over time to create more housing units and accommodate increasing numbers of residents. This article provides systematic analyses of urban

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-07-01

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

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

    International Nuclear Information System (INIS)

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

    1997-07-01

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

  18. The development of safety cases for healthcare services: Practical experiences, opportunities and challenges

    International Nuclear Information System (INIS)

    Sujan, Mark; Spurgeon, Peter; Cooke, Matthew; Weale, Andy; Debenham, Philip; Cross, Steve

    2015-01-01

    There has been growing interest in the concept of safety cases for medical devices and health information technology, but questions remain about how safety cases can be developed and used meaningfully in the safety management of healthcare services and processes. The paper presents two examples of the development and use of safety cases at a service level in healthcare. These first practical experiences at the service level suggest that safety cases might be a useful tool to support service improvement and communication of safety in healthcare. The paper argues that safety cases might be helpful in supporting healthcare organisations with the adoption of proactive and rigorous safety management practices. However, it is also important to consider the different level of maturity of safety management and regulatory oversight in healthcare. Adaptations to the purpose and use of safety cases might be required, complemented by the provision of education to both practitioners and regulators. - Highlights: • Empirical description of safety case development at service level in healthcare. • Safety cases can support adoption of proactive and rigorous safety management. • Adaptation to purpose and use of safety cases might be required in healthcare. • Education should be provided to practitioners and regulators

  19. Development of a gated optical multichannel analyser for laser-plasma spectroscopy

    OpenAIRE

    Corcoran, Richard

    1990-01-01

    An Optical Multichannel Analyser (OMA) has been developed for the detection of radiation from laser-produced plasmas (LPPs). The system is based on a gated image - intensified photodiode array (PDA) Software for the control of, and data acquisition from, the OMA system has been developed. A high resolution (10ns) delay generator was also designed and constructed to permit timeresolved. optical spectroscopy. The system has been tested and operated with a laser plasma source m...

  20. Accident and safety analyses for the HTR-modul. Partial project 1: Computer codes for system behaviour calculation. Final report. Pt. 2

    International Nuclear Information System (INIS)

    Lohnert, G.; Becker, D.; Dilcher, L.; Doerner, G.; Feltes, W.; Gysler, G.; Haque, H.; Kindt, T.; Kohtz, N.; Lange, L.; Ragoss, H.

    1993-08-01

    The project encompasses the following project tasks and problems: (1) Studies relating to complete failure of the main heat transfer system; (2) Pebble flow; (3) Development of computer codes for detailed calculation of hypothetical accidents; (a) the THERMIX/RZKRIT temperature buildup code (covering a.o. a variation to include exothermal heat sources); (b) the REACT/THERMIX corrosion code (variation taking into account extremely severe air ingress into the primary loop); (c) the GRECO corrosion code (variation for treating extremely severe water ingress into the primary loop); (d) the KIND transients code (for treating extremely fast transients during reactivity incidents. (4) Limiting devices for safety-relevant quantities. (5) Analyses relating to hypothetical accidents. (a) hypothetical air ingress; (b) effects on the fuel particles induced by fast transients. The problems of the various tasks are defined in detail and the main results obtained are explained. The contributions reporting the various project tasks and activities have been prepared for separate retrieval from the database. (orig./HP) [de

  1. Accident and safety analyses for the HTR-modul. Partial project 1: Computer codes for system behaviour calculation. Final report. Pt. 1

    International Nuclear Information System (INIS)

    Lohnert, G.; Becker, D.; Dilcher, L.; Doerner, G.; Feltes, W.; Gysler, G.; Haque, H.; Kindt, T.; Kohtz, N.; Lange, L.; Ragoss, H.

    1993-08-01

    The project encompasses the following project tasks and problems: (1) Studies relating to complete failure of the main heat transfer system; (2) Pebble flow; (3) Development of computer codes for detailed calculation of hypothetical accidents; (a) the THERMIX/RZKRIT temperature buildup code (covering a.o. a variation to include exothermal heat sources); (b) the REACT/THERMIX corrosion code (variation taking into account extremely severe air ingress into the primary loop); (c) the GRECO corrosion code (variation for treating extremely severe water ingress into the primary loop); (d) the KIND transients code (for treating extremely fast transients during reactivity incidents. (4) Limiting devices for safety-relevant quantities. (5) Analyses relating to hypothetical accidents. (a) hypothetical air ingress; (b) effects on the fuel particles induced by fast transients. The problems of the various tasks are defined in detail and the main results obtained are explained. The contributions reporting the various project tasks and activities have been prepared for separate retrieval from the database. (orig./HP) [de

  2. Development of the computer code system for the analyses of PWR core

    International Nuclear Information System (INIS)

    Tsujimoto, Iwao; Naito, Yoshitaka.

    1992-11-01

    This report is one of the materials for the work titled 'Development of the computer code system for the analyses of PWR core phenomena', which is performed under contracts between Shikoku Electric Power Company and JAERI. In this report, the numerical method adopted in our computer code system are described, that is, 'The basic course and the summary of the analysing method', 'Numerical method for solving the Boltzmann equation', 'Numerical method for solving the thermo-hydraulic equations' and 'Description on the computer code system'. (author)

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

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  4. The development of technologies of safety analysis for LMR ('03)

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Y. B.; Suk, S. D.; Chang, W. P.; Kwon, Y. M.; Jeong, H. Y.; Ha, K. W.; Heo, S

    2004-03-01

    The developmental objectives of the project, 'The development of safety analysis techniques in LMR', are the code development for the subchannel blockage analysis, the code development for the system transient analysis, the code development for the HCDA(Hypothetical Core Disruptive Accident) analysis, the preliminary safety analysis for KALIMER-600 equipped with the components of new concepts, and the establishment of data base. The purpose of the analysis for subchannel blockage in the subassembly of LMR is to represent quantitatively that the maximum damage due to the accident is within the safety criteria. The computational program should be developed to simulate the thermal hydraulic phenomena and to verify the safety of LMR for the accident. For the purpose, the hybrid scheme has been implemented into the MATRA-LMR code based on the upwind scheme to analyze the various flow fields occurred in the subchannel blockage accident. The turbulent mixing models using the CFX code were assessed to compute more precisely the heat transfer between subchannels. Through this assessment, empirical correction factors of 1.7 for the heat conduction, 0.006 for the turbulent mixing coefficient were obtained. The distributed resistance model instead of wire forcing function has been developed to represent the more exact flow field due to wire-wrap. Other models, such as heat conductor model and various turbulent mixing model, have been implemented into the MATRA-LMR. The ORNL THORS 19-Pin FFM-5B tests have been assessed to validate above new models using the improved MATRA-LMR. The results using MATRA-LMR were well agreed with the experimental data. The subchannel blockage accidents which assumed to be occurred at the three locations for the conceptual plant of KALIMER-600 have been analysed according to blockage size using the MATRA-LMR code. The results of calculations for the design basis events which 6 subchannels were blocked showed the margins of the 290 7.dog. C

  5. REGIONAL DEVELOPMENT THEORIES AND MODELS, A COMPARATIVE ANALYSE.CHALLENGE OF REGIONAL DEVELOPMENT IN ALBANIA

    Directory of Open Access Journals (Sweden)

    Eva\tDHIMITRI

    2015-12-01

    Full Text Available Local governance is a broad concept and is defined as the formulation and execution of collective action at the local level. The purpose of local government is to ensure effective and efficient use of public resources and service delivery at the level closest to citizens. Regional development is a new concept that aims to stimulate and diversify the economic activity of a country (region, to encourage investment in the private sector, to create a new jobs vacancy and improves living standards of the country. Regional development policies are a number of measures designed and promoted by the central and local administration, but the cooperation undertaken at the actors are in a different one, which included the private sector and civil society. At the center of these regional policies or practices is the use of efficient potential of each region, being particularly focused on business, means promoting the development of the new enterprises, promoting labor market and investment, improve the quality of environment, health , education and culture. Traditional objective of regional development policies is the reduction of territorial disparities for achieving a relative balance between economic and social levels of development in different areas in the national territory. Regional development is the actual task of local government units in Albania, and is one of the tasks and challenges of the future. Currently it takes a special importance in the context of European Union integration. Reforms have begun to change the system in 1990 in order to implement local democracy and decentralization principles that are present today. Inequalities that exist within the region and between them indicate that in some regions the economic potential is not being fully utilized, and that it reduces the overall performance in national level.

  6. Development of Broadband Nuclear Safety Data Network (B-NSDN)

    International Nuclear Information System (INIS)

    Son, Gwang Seop; Kim, Dong Hoon; Park, Gi Yong

    2011-01-01

    Recently as introducing digital safety system in nuclear power plant, more data transmission capacity is required. Bandwidth of existing communication network is about a few Mbps. Thus data transmission quantity in recently digital safety system is beyond existing communication network's capacity. In this paper, new protocol that is suitable for safety system communication network is designed. FPGA based communication system is implemented. As result of test, effective bandwidth of B-NSDN is about 20Mbps

  7. Nursing students' perspectives and suggestions on patient safety--implications for developing the nursing education curriculum in Iran.

    Science.gov (United States)

    Vaismoradi, Mojtaba; Bondas, Terese; Jasper, Melanie; Turunen, Hannele

    2014-02-01

    Nursing students' close involvement in knowledge development about patient safety will enhance the integrity of the current content of nursing education and pave the way towards developing a nursing curriculum that facilitates achieving a safer health-care system. This study explores nursing students' perspectives and suggestions on developing patient safety aspects of the nursing curriculum in the context of Iranian culture. A qualitative methodology involving three focus groups with a purposive sample of 18 nursing students from a large Iranian nursing school, utilising directed semi-structured interviews generated data, which was analysed using the content analysis process. Two main themes emerged from content analysis: (1) "involving students fully in patient care" with subthemes 'building a trusting relationship between education and practice', and 'promoting inter-dependence between health-care providers', and (2) "structuring patient safety education" with subthemes 'transforming nursing routines to evidence-based care', and 'connecting care to patient safety issues'. The extent of students' involvement in clinical practice and clinical nurses' roles in student education in practice requires clarification. The curriculum needs to incorporate patient safety aspects throughout, and include interdisciplinary education to ensure compliance with patient safety policies. Moreover, successful implementation of such a curriculum necessitates cooperation from nursing practice and instructors to meet nursing students' expectations. Copyright © 2014 Elsevier Ltd. All rights reserved.

  8. LABOUR PROTECTION AND INDUSTRIAL SAFETY IN UKRAINE: PROBLEMS OF TRANSITION PERIOD AND PERSPECTIVE WAYS OF DEVELOPMENT

    Directory of Open Access Journals (Sweden)

    A. P. BOCHKOVSKY

    2016-12-01

    1.9 fold greater than the corresponding figure in Ukraine, and the number of subjects with regard to the issues of the labour protection and industrial safety, which are taught to students in fulfilling the work programmes at nonspecialised Polish higher educational establishments is greater than that in Ukrainian several fold. The statistical data regarding the dynamics of the accident number increase in Ukraine and their causes within a period of  “Зернові продукти і комбікорми”, 201643 http://www.grain-mixedfodders.com Зернові продукти і комбікорми Vol.64, I.4/ 2016 2015 and 2016 are presented and analysed in the context of recent negative changes including the reduction of class hours for students learning the disciplines of "Sectoral Labour Protection", "Basics of Labour Protection", "Foundations of Life Activity Safety" and "Civil protection", merging such subjects, and cancellation of the graduation project relevant sections in most HEI of Ukraine On the grounds of the research, priority directions for developing the labour protection and industrial safety in Ukraine on the stage of European integration are proposed.Based on comparative analysis of the industrial accident causes in Ukraine and EU countries this article establishes that the main accident reasons are organizational ones (50 to 70% of the total number of cases, however such indicators as the registered in Ukraine fatal cases frequency coefficient (per 1 thousand of employees and the fatal accidents-total accidents number ratio are greater than the similar indicators in Europe by about 2- and 100-fold, respectively. It is noted that the issues of improving the work safety in Ukraine towards the association with the European Union should be considered in the context of two main planes, which are associated with changes in the legislative and educational systems. Within this article, the authors analyse the main inter-sectoral and sectoral

  9. Safety and effective developing nuclear power to realize green and low-carbon development

    Directory of Open Access Journals (Sweden)

    Qi-Zhen Ye

    2016-03-01

    Full Text Available This paper analyzes the role of nuclear power of China's energy structure and industry system. Comparing with other renewable energy the nuclear power chain has very low greenhouse gas emission, so it will play more important role in China's low-carbon economy. The paper also discussed the necessity of nuclear power development to achieve emission reduction, energy structure adjustment, nuclear power safety, environmental protection, enhancement of nuclear power technology, nuclear waste treatment, and disposal, as well as nuclear power plant decommissioning. Based on the safety record and situation of the existing power plants in China, the current status of the development of world nuclear power technology, and the features of the independently designed advanced power plants in China, this paper aims to demonstrate the safety of nuclear power. A nuclear power plant will not cause harm either to the environment and nor to the public according to the real data of radioactivity release, which are obtained from an operational nuclear plant. The development of nuclear power technology can enhance the safety of nuclear power. Further, this paper discusses issues related to the nuclear fuel cycle, the treatment, and disposal strategies of nuclear waste, and the decommissioning of a nuclear power plant, all of which are issues of public concern.

  10. Development of 'health and environmental safety assessment network system (HESANS)'

    International Nuclear Information System (INIS)

    Nakamura, Yuji

    1994-01-01

    With the recent advance of the utilization of nuclear energy in a large scale, social interest is being focussed in the potential risk which the nuclear technology will accompany. Especially after the accidents in Chernobyl and other nuclear facilities, serious anxiety to the utilization of nuclear energy is prevailing among the general public. In order to meet the anxiety and distrust of the population in the use of the nuclear power, the health effect or risk which radioactive materials released into the environment will bring about should be comprehensively and properly evaluated, and then should be widely reported to the population. The development of HESANS code system (Health and Environmental Safety Assessment Network System) was planned to set up such a comprehensive computer code that covers a whole pathway of radioactive material from its release to estimates of derived health effects in the population, including the countermeasures for intervention as well. Though the whole system is not totally completed yet so far, the framework of the system has been concreted together with many sub-systems which compose the main part of the code. This report puts main stress on the objective of the development project and the main frame or the structure of the code system. (author)

  11. Strategic niche management for biofuels: Analysing past experiments for developing new biofuel policies

    International Nuclear Information System (INIS)

    Laak, W.W.M. van der; Raven, R.P.J.M.; Verbong, G.P.J.

    2007-01-01

    Biofuels have gained a lot of attention since the implementation of the 2003 European Directive on biofuels. In the Netherlands the contribution of biofuels is still very limited despite several experiments in the past. This article aims to contribute to the development of successful policies for stimulating biofuels by analysing three experiments in depth. The approach of strategic niche management (SNM) is used to explain success and failure of these projects. Based on the analysis as well as recent innovation literature we develop a list of guidelines that is important to consider when developing biofuel policies

  12. Development and validation of safety climate scales for mobile remote workers using utility/electrical workers as exemplar.

    Science.gov (United States)

    Huang, Yueng-Hsiang; Zohar, Dov; Robertson, Michelle M; Garabet, Angela; Murphy, Lauren A; Lee, Jin

    2013-10-01

    The objective of this study was to develop and test the reliability and validity of a new scale designed for measuring safety climate among mobile remote workers, using utility/electrical workers as exemplar. The new scale employs perceived safety priority as the metric of safety climate and a multi-level framework, separating the measurement of organization- and group-level safety climate items into two sub-scales. The question of the emergence of shared perceptions among remote workers was also examined. For the initial survey development, several items were adopted from a generic safety climate scale and new industry-specific items were generated based on an extensive literature review, expert judgment, 15-day field observations, and 38 in-depth individual interviews with subject matter experts (i.e., utility industry electrical workers, trainers and supervisors of electrical workers). The items were revised after 45 cognitive interviews and a pre-test with 139 additional utility/electrical workers. The revised scale was subsequently implemented with a total of 2421 workers at two large US electric utility companies (1560 participants for the pilot company and 861 for the second company). Both exploratory (EFA) and confirmatory factor analyses (CFA) were adopted to finalize the items and to ensure construct validity. Reliability of the scale was tested based on Cronbach's α. Homogeneity tests examined whether utility/electrical workers' safety climate perceptions were shared within the same supervisor group. This was followed by an analysis of the criterion-related validity, which linked the safety climate scores to self-reports of safety behavior and injury outcomes (i.e., recordable incidents, missing days due to work-related injuries, vehicle accidents, and near misses). Six dimensions (Safety pro-activity, General training, Trucks and equipment, Field orientation, Financial Investment, and Schedule flexibility) with 29 items were extracted from the EFA to

  13. New developments in South African health and safety legislation ,

    African Journals Online (AJOL)

    unions and employee representatives. BOSH unfonu- nately compounds the confusion as to what and whom the safety representatives represent. Do' they represent health and safety? Or do they represent employees exposed to hazards at work (as opposed to manage- ment)? Either way the interface between health and.

  14. Software safety analysis techniques for developing safety critical software in the digital protection system of the LMR

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jang Soo; Cheon, Se Woo; Kim, Chang Hoi; Sim, Yun Sub

    2001-02-01

    This report has described the software safety analysis techniques and the engineering guidelines for developing safety critical software to identify the state of the art in this field and to give the software safety engineer a trail map between the code and standards layer and the design methodology and documents layer. We have surveyed the management aspects of software safety activities during the software lifecycle in order to improve the safety. After identifying the conventional safety analysis techniques for systems, we have surveyed in details the software safety analysis techniques, software FMEA(Failure Mode and Effects Analysis), software HAZOP(Hazard and Operability Analysis), and software FTA(Fault Tree Analysis). We have also surveyed the state of the art in the software reliability assessment techniques. The most important results from the reliability techniques are not the specific probability numbers generated, but the insights into the risk importance of software features. To defend against potential common-mode failures, high quality, defense-in-depth, and diversity are considered to be key elements in digital I and C system design. To minimize the possibility of CMFs and thus increase the plant reliability, we have provided D-in-D and D analysis guidelines.

  15. Software safety analysis techniques for developing safety critical software in the digital protection system of the LMR

    International Nuclear Information System (INIS)

    Lee, Jang Soo; Cheon, Se Woo; Kim, Chang Hoi; Sim, Yun Sub

    2001-02-01

    This report has described the software safety analysis techniques and the engineering guidelines for developing safety critical software to identify the state of the art in this field and to give the software safety engineer a trail map between the code and standards layer and the design methodology and documents layer. We have surveyed the management aspects of software safety activities during the software lifecycle in order to improve the safety. After identifying the conventional safety analysis techniques for systems, we have surveyed in details the software safety analysis techniques, software FMEA(Failure Mode and Effects Analysis), software HAZOP(Hazard and Operability Analysis), and software FTA(Fault Tree Analysis). We have also surveyed the state of the art in the software reliability assessment techniques. The most important results from the reliability techniques are not the specific probability numbers generated, but the insights into the risk importance of software features. To defend against potential common-mode failures, high quality, defense-in-depth, and diversity are considered to be key elements in digital I and C system design. To minimize the possibility of CMFs and thus increase the plant reliability, we have provided D-in-D and D analysis guidelines

  16. 75 FR 20009 - Development of NRC's Safety Culture Policy Statement: Cancellation of Public Workshops Scheduled...

    Science.gov (United States)

    2010-04-16

    ... COMMISSION Development of NRC's Safety Culture Policy Statement: Cancellation of Public Workshops Scheduled... apply to all licensees/certificate holders; and (3) receive comments on the draft safety culture policy... forging a consensus around the objectives, strategies, activities and measures that enhance safety culture...

  17. Preliminary analysing of experimental data for the development of high Cr Alloy Creep damage Constitutive Equations

    OpenAIRE

    An, Lili; Xu, Qiang; Xu, Donglai; Lu, Zhongyu

    2012-01-01

    This conference paper presents the current research of preliminary analysing of experimental data for the development of high Cr Alloy Creep damage Constitutive Equations (such as P91 alloy). Firstly, it briefly introduces the background of general creep deformation, rupture and continuum damage mechanics. Secondly, it illustrates the constitutive equations used for P91 alloy or its weldment, especially of the form and deficiencies of two kinds of most widely used typical creep damage constit...

  18. Status, experience and future prospects for the development of probabilistic safety criteria

    International Nuclear Information System (INIS)

    1989-09-01

    During 27-31 January 1986 the IAEA held a Technical Committee Meeting on ''Status, Experience, and Future Prospects for the Development of Probabilistic Safety Criteria''. Participation included representation of essentially all countries with major developments in the area as well as the Nuclear Energy Agency of the OECD and CEC. Though it has to be recognized that in such a short time period it is impossible to resolve or even analyse all aspects of this complex issue, the present situation, the main problems and the directions for future work clearly emerged. This report was prepared by the members of the Technical Committee based on the opinions expressed and on the information available at the time of the meeting. The report also contains 20 papers presented at the meeting by participants. A separate abstract was prepared for each of these 20 papers. Refs, figs and tabs

  19. Development and application of neutron transport methods and uncertainty analyses for reactor core calculations. Technical report; Entwicklung und Einsatz von Neutronentransportmethoden und Unsicherheitsanalysen fuer Reaktorkernberechnungen. Technischer Bericht

    Energy Technology Data Exchange (ETDEWEB)

    Zwermann, W.; Aures, A.; Bernnat, W.; and others

    2013-06-15

    This report documents the status of the research and development goals reached within the reactor safety research project RS1503 ''Development and Application of Neutron Transport Methods and Uncertainty Analyses for Reactor Core Calculations'' as of the 1{sup st} quarter of 2013. The superordinate goal of the project is the development, validation, and application of neutron transport methods and uncertainty analyses for reactor core calculations. These calculation methods will mainly be applied to problems related to the core behaviour of light water reactors and innovative reactor concepts. The contributions of this project towards achieving this goal are the further development, validation, and application of deterministic and stochastic calculation programmes and of methods for uncertainty and sensitivity analyses, as well as the assessment of artificial neutral networks, for providing a complete nuclear calculation chain. This comprises processing nuclear basis data, creating multi-group data for diffusion and transport codes, obtaining reference solutions for stationary states with Monte Carlo codes, performing coupled 3D full core analyses in diffusion approximation and with other deterministic and also Monte Carlo transport codes, and implementing uncertainty and sensitivity analyses with the aim of propagating uncertainties through the whole calculation chain from fuel assembly, spectral and depletion calculations to coupled transient analyses. This calculation chain shall be applicable to light water reactors and also to innovative reactor concepts, and therefore has to be extensively validated with the help of benchmarks and critical experiments.

  20. Development of electro-optical instrumentation for reactor safety studies

    International Nuclear Information System (INIS)

    Turko, B.T.; Kolbe, W.F.; Leskovar, B.; Sun, R.K.

    1980-11-01

    The development of new electro-optical instrumentation for reactor safety studies is described. The system measures the thickness of the water film and droplet size and velocity distributions which would be encountered in the annular two-phase flow in a reactor cooling system. The water film thickness is measured by a specially designed capacitance system with a short time constant. Water droplet size and velocity are measured by a subsystem consisting of a continuously pulsed laser light source, a vidicon camera, a video recorder, and an automatic image analyzer. An endoscope system attached to the video camera is used to image the droplets. Each frame is strobed with two accurately spaced uv light pulses, from two sequentially fired nitrogen lasers. The images are stored in the video disk recorder. The modified automatic image analyzer is programmed to digitize the droplet size and velocity distributions. Many special optical, mechanical and electronic system components were designed and fabricated. They are described in detail, together with calibration charts and experimental results

  1. Development of a heat transfer correlation for the HPLWR fuel assembly by means of CFD analyses

    Energy Technology Data Exchange (ETDEWEB)

    Lycklama a Nijeholt, J.A.; Visser, D.C. [Nuclear Research and Consultancy Group (NRG), Petten (Netherlands); Laurien, E. [Univ. of Stuttgart, Stuttgart (Germany); Anglart, H. [Royal Inst. of Tech., Stockholm (Sweden); Chandra, L. [Indian Inst. of Tech., Rajasthan (India)

    2011-07-01

    The High Performance Light Water Reactor (HPLWR) has been under development in the HPLWR phase-2 project funded by the European Union. The HPLWR project started September 2006 and ended February 2010. Work package 5 within this project involves the improved understanding of heat transfer, CFD model development and validation, and the prediction of the heat transfer rate in a HPLWR fuel assembly. USTUTT, KTH, NRG and FZK contributed to this work package. The overall objective of work package 5 was the development of a heat transfer correlation for the prediction of the heat transfer rate in the HPLWR fuel assembly by means of CFD analyses. In the HPLWR fuel assembly, a helical wire has been selected as spacer and mixing device. This wire-wrap imposed a significant challenge in the development of the geometrical models for the CFD analyses. Due to the wire-wrap it was not possible to model a full fuel assembly consisting of 40 rods. Therefore, an alternative procedure has been adopted to develop a heat transfer correlation for the HPLWR fuel assembly. This procedure involved the definition of correction factors accounting for the effect of the rod bundle geometry and the wire-wrap spacer with respect to a smooth circular tube with super-critical water. The present paper describes the procedure followed in work package 5 of the HPLWR phase-2 project for the development of a heat transfer correlation for the HPLWR fuel assembly design and presents the derivation of the applied correction factors from a large set of CFD analyses for different representative geometries like an annulus, a single sub-channel and a 4 rod-bundle, all with and without inclusion of the wire wrap. (author)

  2. Development of a methodology for the safety assessment of near surface disposal facilities for radioactive waste

    International Nuclear Information System (INIS)

    Simon, I.; Cancio, D.; Alonso, L.F.; Agueero, A.; Lopez de la Higuera, J.; Gil, E.; Garcia, E.

    2000-01-01

    The Project on the Environmental Radiological Impact in CIEMAT is developing, for the Spanish regulatory body Consejo de Seguridad Nuclear (CSN), a methodology for the Safety Assessment of near surface disposal facilities. This method has been developed incorporating some elements developed through the participation in the IAEA's ISAM Programme (Improving Long Term Safety Assessment Methodologies for Near Surface Radioactive Waste Disposal Facilities). The first step of the approach is the consideration of the assessment context, including the purpose of the assessment, the end-Points, philosophy, disposal system, source term and temporal scales as well as the hypothesis about the critical group. Once the context has been established, and considering the peculiarities of the system, an specific list of features, events and processes (FEPs) is produced. These will be incorporated into the assessment scenarios. The set of scenarios will be represented in the conceptual and mathematical models. By the use of mathematical codes, calculations are performed to obtain results (i.e. in terms of doses) to be analysed and compared against the criteria. The methodology is being tested by the application to an hypothetical engineered disposal system based on an exercise within the ISAM Programme, and will finally be applied to the Spanish case. (author)

  3. Development of Non-LOCA Safety Analysis Methodology with RETRAN-3D and VIPRE-01/K

    International Nuclear Information System (INIS)

    Kim, Yo-Han; Cheong, Ae-Ju; Yang, Chang-Keun

    2004-01-01

    Korea Electric Power Research Institute has launched a project to develop an in-house non-loss-of-coolant-accident analysis methodology to overcome the hardships caused by the narrow analytical scopes of existing methodologies. Prior to the development, some safety analysis codes were reviewed, and RETRAN-3D and VIPRE-01 were chosen as the base codes. The codes have been modified to improve the analytical capabilities required to analyze the nuclear power plants in Korea. The methodologies of the vendors and the Electric Power Research Institute have been reviewed, and some documents of foreign utilities have been used to compensate for the insufficiencies. For the next step, a draft methodology for pressurized water reactors has been developed and modified to apply to Westinghouse-type plants in Korea. To verify the feasibility of the methodology, some events of Yonggwang Units 1 and 2 have been analyzed from the standpoints of reactor coolant system pressure and the departure from nucleate boiling ratio. The results of the analyses show trends similar to those of the Final Safety Analysis Report

  4. Development of safety assessment model based on TRU-2 report using GoldSim

    International Nuclear Information System (INIS)

    Ebina, Takanori; Inagaki, Manabu; Kato, Tomoko

    2011-03-01

    The safety assessment model at 'Second Progress Report on Research and Development for TRU Waste Disposal in Japan'(TRU-2 report) was designed using the numerical code TIGER, that allows the physical and chemical properties within the system to vary with time. In the future, at the examination to optimize nuclear fuel cycle for geological disposal, it is expected that the analysis that has many cases like sensitivity analysis and uncertainty analysis are in demand. The numerical code TIGER is a calculation code that analyze engineered barrier system and geological barrier system, and its numerical model is verified with nuclide migration code for engineered barrier system MESHNOTE, and nuclide migration code for geosphere MATRICS. At the analysis using TIGER, the migration (i.e. Engineered barrier system, Host rock and Fault) have to be analysed independently at each region, consequently the huge number of complicated parameter setting have been required. On the other hand, by using numerical code GoldSim, all regions are analyzed synchronously and parameters can be defined at same model. So it makes quality control of parameters easier. Furthermore, analysis time by GoldSim is shorter than TIGER and GoldSim can calculate many number of Monte Carlo simulations among multiple computers. In future, Safety Analyses of TRU waste package disposal will be carried out according as study of an optimization of nuclear fuel cycle. Therefor, safety assessment model for TRU waste disposal using GoldSim was designed, and calculation results were verified by comparing with the result of TRU-2 report. (author)

  5. Provision and Updating of Estimates of Reliability Parameters for Use in Reliability Analyses of Safety-Instrumented Systems

    OpenAIRE

    Bjartnes, Magnus Woll

    2012-01-01

    Safety-instrumented systems are implemented in the industry to prevent accidents to occur and escalate. The blowdown system on a oil production ship is one example of such system. If a fire breaks out on the ship, the blowdown system's role is to remove the flammable gases from the current production lines on the ship. This is done by opening of the blowdown valves, that are installed on the different production lines. In this thesis, the blowdown system on a new Teekay ship, and espesically ...

  6. Development and Application of Level 2 Probabilistic Safety Assessment for Nuclear Power Plants. Specific Safety Guide

    International Nuclear Information System (INIS)

    2010-01-01

    The objective of this Safety Guide is to provide recommendations for meeting the IAEA safety requirements in performing or managing a level 2 probabilistic safety assessment (PSA) project for a nuclear power plant; thus it complements the Safety Guide on level 1 PSA. One of the aims of this Safety Guide is to promote a standard framework, standard terms and a standard set of documents for level 2 PSAs to facilitate regulatory and external peer review of their results. It describes all elements of the level 2 PSA that need to be carried out if the starting point is a fully comprehensive level 1 PSA. Contents: 1. Introduction; 2. PSA project management and organization; 3. Identification of design aspects important to severe accidents and acquisition of information; 4. Interface with level 1 PSA: Grouping of sequences; 5. Accident progression and containment analysis; 6. Source terms for severe accidents; 7. Documentation of the analysis: Presentation and interpretation of results; 8. Use and applications of the PSA; Annex I: Example of a typical schedule for a level 2 PSA; Annex II: Computer codes for simulation of severe accidents; Annex III: Sample outline of documentation for a level 2 PSA study.

  7. Transit safety retrofit package development : TRP concept of operations.

    Science.gov (United States)

    2014-05-01

    This document describes the Concept of Operations (ConOps) for the Transit Safety Retrofit Package (TRP). The : ConOps describes the current state of operations with respect to the integration of Connected Vehicle technology in : transit buses, estab...

  8. Insulin analogues versus human insulin in type 1 diabetes: direct and indirect meta-analyses of efficacy and safety

    Directory of Open Access Journals (Sweden)

    Andréia Cristina Conegero Sanches

    2013-09-01

    Full Text Available All patients with Diabetes Mellitus (DM receive insulin therapy. In this study, we evaluated the efficacy, safety and tolerability of human insulin and insulin analogues. We performed a systematic review of the literature and a meta-analysis according to the Cochrane Collaboration methodology. In the absence of clinical studies comparing insulins, we performed a mixed treatment comparison to establish the differences between the active treatments. We included studies published from 1995 to 2010. HbA1c results, episodes of hypoglycemia and nocturnal hypoglycemia data were extracted and analyzed. Thirty-five randomized clinical trials were selected after examining the abstract and a full text review. These studies included 4,206 patients who received long-acting insulin analogues and 5,733 patients who received short-acting insulin analogues. Pooled data regarding efficacy indicated no significant differences in HbA1c values between glargine or detemir (once daily and NPH insulin. However, a twice-daily dose of detemir produced differences in HbA1c values that favored detemir (-0.14% [95% CI: -0.21 to -0.08]; p<0.0001; I²=0%. Direct and indirect comparisons are consistent and show that there were no significant differences between human insulin and insulin analogues in efficacy or safety. Our results indicate that long- and short-acting insulin analogues offer few clinical advantages over conventional human insulin.

  9. Development of the KINS Safety Culture Maturity Model for Self and Independent Assessment

    International Nuclear Information System (INIS)

    Sheen, C.; Choi, Y.S.

    2016-01-01

    Safety culture of an organization is cultivated and affected not only by societal and regulatory environment of the organization, but by its philosophies, policies, events and activities experienced in the process of accomplishing its mission. The safety culture would be continuously changed by the interactions between its members along with time as an organic entity. In order to perform a systematic self- or independent assessment of safety culture, a safety culture assessment model (SCAM) properly reflecting cultural characteristics should be necessary. In addition, a SCAM should be helpful not only to establish correct directions, goals, and strategies for safety culture development, but should anticipating obstacles against safety culture development in the implementation process derived from the assessment. In practical terms, a SCAM should be useful for deriving effective guidelines and implementing of corrective action programs for the evaluated organization. Korea Institute of Nuclear Safety (KINS) performed a research project for six years to develop a SCAM satisfying the above prerequisites for self- and independent assessment. The KINS SCAM was developed based on the five stage safety culture maturity model proposed by Professor Patrick Hudson and was modified into four stages to reflect existing safety culture assessment experiences at Korean nuclear power plants. In order to define the change mechanism of safety culture for development and reversion, the change model proposed by Prochaska and DiClemente was introduced into KINS SCAM and developed into the Spiral Change Model.

  10. DEVELOPING SAFETY INDICATORS ON THE BASIS OF THE ICAO RECOMMENDATIONS

    Directory of Open Access Journals (Sweden)

    V. D. Sharov

    2014-01-01

    Full Text Available The article offers direct use of the recommendations of SMM ICAO Doc.9859, 3rd ed. 2013, for calculation the target and alert levels of safety indicators. Examples of calculation based on data of 2011 and monitoring of the current indicators during 2012 are presented. Safety indicators for airlines in terms of “numbers of incidents per 1000 flight hours” could be calculated on the basis of the state values through the «coefficient of conformity».

  11. Development of safety factors to be used for evaluation of cracked nuclear components

    International Nuclear Information System (INIS)

    Brickstad, B.; Bergman, M.

    1996-10-01

    A modified concept for safety evaluation is introduced which separately accounts for the failure mechanisms fracture and plastic collapse. For application on nuclear components a set of safety factors are also proposed that retain the safety margins expressed in ASME, section III and XI. By performing comparative studies of the acceptance levels for surface cracks in pipes and a pressure vessel, it is shown that some of the anomalies connected with the old safety procedures are removed. It is the authors belief that the outlined safety evaluation procedure has the capability of treating cracks in a consistent way and that the procedure together with the proposed safety factors fulfill the basic safety requirements for nuclear components. Hopefully, it is possible in the near future to develop a probabilistic safety assessment procedure in Sweden, which enables a systematic treatment of uncertainties in the involved data. 14 refs

  12. Development of a quality assurance safety assessment database for near surface radioactive waste disposal

    International Nuclear Information System (INIS)

    Park, J. W.; Kim, C. L.; Park, J. B.; Lee, E. Y.; Lee, Y. M.; Kang, C. H.; Zhou, W.; Kozak, M. W.

    2003-01-01

    A quality assurance safety assessment database, called QUARK (QUality Assurance program for Radioactive waste management in Korea), has been developed to manage both analysis information and parameter database for safety assessment of Low- and Intermediate-Level radioactive Waste (LILW) disposal facility in Korea. QUARK is such a tool that serves QA purposes for managing safety assessment information properly and securely. In QUARK, the information is organized and linked to maximize the integrity of information and traceability. QUARK provides guidance to conduct safety assessment analysis, from scenario generation to result analysis, and provides a window to inspect and trace previous safety assessment analysis and parameter values. QUARK also provides default database for safety assessment staff who construct input data files using SAGE(Safety Assessment Groundwater Evaluation), a safety assessment computer code

  13. Safety and effective developing nuclear power to realize green and low-carbon development

    OpenAIRE

    Ye, Qi-Zhen

    2016-01-01

    This paper analyzes the role of nuclear power of China's energy structure and industry system. Comparing with other renewable energy the nuclear power chain has very low greenhouse gas emission, so it will play more important role in China's low-carbon economy. The paper also discussed the necessity of nuclear power development to achieve emission reduction, energy structure adjustment, nuclear power safety, environmental protection, enhancement of nuclear power technology, nuclear waste trea...

  14. Method to develop data supporting consequence analyses of transporting nuclear materials in the United States

    International Nuclear Information System (INIS)

    Reese, R.T.; Sandoval, R.P.

    1980-01-01

    The Transportation System Safety Evaluation (TSSE) program at Sandia National Laboratories' Transportation Technology Center was initiated to provide the necessary information on source terms for nuclear materials subjected to extreme environments. The techniques for derivation of source terms for the fuel alone has been described as well as the outline for package response. An additional facet of this problem is the development of analytical methods to describe the transport of the released radionuclides from the fuel rods to possible release points. This work is also covered in the TSSE program. Not all the work required will be performed or funded by Sandia; rather existing work will be sought out and ongoing work will be utilized in an attempt to unify the presentation of data and thus increase its usefulness

  15. Factor Analysis and Framework Development for Incorporating Public Trust on Nuclear Safety issues

    International Nuclear Information System (INIS)

    Cho, Seongkyung; Lee, Gyebong; Lee, Gihyung; Lee, Gyehwi; Jeong, Jina

    2014-01-01

    The Korea Institute of Nuclear Safety (KINS), a regulatory expert organization in charge of nuclear safety in Korea, realized that a more fundamental and systematic analysis of activities is needed to actively meet the greater variety of concerns people have and increase the reliability of the results of regulation. Nuclear safety, a highly specialized field, has previously been discussed primarily from the viewpoint of the engineers who deal with the technology, but now 'public trust in nuclear safety' has to be viewed from the standpoint of the general public and from the socio-cultural perspective. Specific measures must be taken to examine which factors affect public trust and how we can secure and reproduce those factors to gain it. Also, an efficient system for incorporating public trust in nuclear safety must be established. In this study, various case studies were examined to identify the factors that affect public trust in nuclear safety. First, nuclear safety laws and information disclosure systems of major countries were examined by investigating data and conducting in-depth interviews. To explore a public framework concerning nuclear safety, big data of social media were analyzed. Also, Q methodology was used to analyze the risk schemata of the opinion leaders living in areas near nuclear power plants. Several surveys were conducted to analyze the amount of trust the public had in nuclear safety as well as their awareness of nuclear safety issues. Based on these analyses, factors affecting public trust in nuclear safety were extracted, and measures to build systems incorporating public trust in nuclear safety were proposed. This study addresses the public trust in nuclear safety on condition that the safety is ensured technically and mechanically

  16. State of the art in establishing computed models of adsorption processes to serve as a basis of radionuclide migration assessment for safety analyses

    International Nuclear Information System (INIS)

    Koss, V.

    1991-01-01

    An important point in safety analysis of an underground repository is adsorption of radionuclides in the overlying cover. Adsorption may be judged according to experimental results or to model calculations. Because of the reliability aspired in safety analyses, it is necessary to strengthen experimental results by theoretical calculations. At the time, there is no single thermodynamic model of adsorption to be agreed on. Therefore, this work reviews existing equilibrium models of adsorption. Limitations of the K d -concept and of adsorption-isotherms according to Freundlich and Langmuir are mentioned. The surface ionisation and complexation edl model is explained in full as is the criticism of this model. The application is stressed of simple surface complexation models to adsorption experiments in natural systems as is experimental and modelling work according to systems from Gorleben. Hints are given how to deal with modelling of adsorption related to Gorleben systems in the future. (orig.) [de

  17. Lessons learned - development of the tritium facilities 5480.23 safety analysis report and technical safety requirements

    International Nuclear Information System (INIS)

    Cappucci, A.J. Jr.; Bowman, M.E.; Goff, L.

    1997-01-01

    A review was performed which identified open-quotes Lessons Learnedclose quotes from the development of the 5480.23 Tritium Safety Analysis Report (SAR) and the Technical Safety Requirements (TSR) for the Tritium Facilities (TF). The open-quotes Lessons Learnedclose quotes were based on an evaluation of the use of the SRS procedures, processes, and work practices which contributed to the success or lack thereof. This review also identified recommendations and suggestions for improving the development of SARs and TSRs at SRS. The 5480.23 SAR describes the site for the TF, the various process systems in the process buildings, a complete hazards and accident analysis of the most significant hazards affecting the nearby offsite population, and the selection of safety systems, structures, and components to protect both the public and site workers. It also provides descriptions of important programs and processes which add defense in depth to public and worker protection

  18. The business case for safety and health at work : cost-benefit analyses of interventions in small and medium-sized enterprises

    NARCIS (Netherlands)

    Targoutzidis, A.; Koukoulaki, T.; Schmitz-Felten, E.; Kuhl, K.; Oude Hengel, K.M.; Rijken, E.; Broek K. van den; Kluser, R.

    2014-01-01

    This report examines the economic aspects of occupational safety and health (OSH) interventions in small and medium-sized businesses (SMEs). First, case studies in the existing literature were identified and examined. Second, 13 new case studies on OSH initiatives in European SMEs were developed,

  19. Safety report content and development for test loop facility on MARIA reactor

    International Nuclear Information System (INIS)

    Konechko, A.; Shumskij, A.M.; Mikul'ahin, V.E.

    1982-01-01

    A 600 kW test loop facility for investigatin.o safety problems is realized on MARIA reactor in Poland together with USSR organizations. Safety reports have been developed in two steps at the designstage. The 1st report being essentially a preliminary safety analysis was developed within the scope of the feasibility study. At the engineering design stage the preliminary test loop facility safety report had been prepared considering measures excluding the possibility of the MARIA reactor damage. The test loop facility safety report is fulfilled for normal, transient and emergency operation regimes. Separate safety basing for each group of experiments will be prepared. The report presents the test loop facility safety criteria coordinated by the nuclear safety comission. They contains the preliminary reports on the test loop facility safety. At the final stage of construction and at thecommitioning stage the start-up safety report will be developed which after required correction and adding up the putting into operation data will turn into operation safety report [ru

  20. DEVELOPMENT OF RISK-BASED AND TECHNOLOGY-INDEPENDENT SAFETY CRITERIA FOR GENERATION IV SYSTEMS

    International Nuclear Information System (INIS)

    Kastenberg, William E.; Blandford, Edward; Kim, Lance

    2009-01-01

    This project has developed quantitative safety goals for Generation IV (Gen IV) nuclear energy systems. These safety goals are risk based and technology independent. The foundations for a new approach to risk analysis has been developed, along with a new operational definition of risk. This project has furthered the current state-of-the-art by developing quantitative safety goals for both Gen IV reactors and for the overall Gen IV nuclear fuel cycle. The risk analysis approach developed will quantify performance measures, characterize uncertainty, and address a more comprehensive view of safety as it relates to the overall system. Appropriate safety criteria are necessary to manage risk in a prudent and cost-effective manner. This study is also important for government agencies responsible for managing, reviewing, and for approving advanced reactor systems because they are charged with assuring the health and safety of the public

  1. DEVELOPMENT OF RISK-BASED AND TECHNOLOGY-INDEPENDENT SAFETY CRITERIA FOR GENERATION IV SYSTEMS

    Energy Technology Data Exchange (ETDEWEB)

    William E. Kastenberg; Edward Blandford; Lance Kim

    2009-03-31

    This project has developed quantitative safety goals for Generation IV (Gen IV) nuclear energy systems. These safety goals are risk based and technology independent. The foundations for a new approach to risk analysis has been developed, along with a new operational definition of risk. This project has furthered the current state-of-the-art by developing quantitative safety goals for both Gen IV reactors and for the overall Gen IV nuclear fuel cycle. The risk analysis approach developed will quantify performance measures, characterize uncertainty, and address a more comprehensive view of safety as it relates to the overall system. Appropriate safety criteria are necessary to manage risk in a prudent and cost-effective manner. This study is also important for government agencies responsible for managing, reviewing, and for approving advanced reactor systems because they are charged with assuring the health and safety of the public.

  2. Development of Flight Safety Prediction Methodology for U. S. Naval Safety Center. Revision 1

    Science.gov (United States)

    1970-02-01

    POWER DISTRIBUTION K Fbl FAH FBA FAAAAAAAAA K FOG FBB FAAAAAAAAA K FRG F8C FAkAAAAAAA K FOG FBO FAAAAAAAAA K FOG FBE FAAAAAAAAA K FBO FoF FAAAAAAAAA FOG F...safety-prediction math model. *Department of the Navy, Naval Aviation Maintenance anid Material Mangement Manual, Publication 0618-200-0100. iE-15 TABLE

  3. The development of safety assessment technology for the radwaste disposal

    International Nuclear Information System (INIS)

    Han, Kyong Won; Cho, Won Jin; Lee, Han Soo; Lee, Jai Wan; Park, Chung Kyun; Lee, Myun Joo; Cho, Young Hwan; Choi, Heui Joo; Lee, Youn Myoung; Park, Hee Sung

    1989-02-01

    This report is composed of three parts. In part I, an improved radwaste disposal safety assessment code named SADROCM, was developed by upgrading the existing SADROC. A numerical approach was adopted for the simulation of diffusion into rock pore and advection in the fracture. Also quantification of resaturation time in repository was obtained by introducing theoretical resaturation model and groundwater flow model. In radiation dose model the calculated dose by existing model was compared with that by LIMCAL code. Part II resulted in several findings. Regarding the leaching of radionuclide, two steps were observed. It was identified that in the initial step, radionuclide leaching occurred at the surface of solidified waste, and then leaching proceeded by dissolution and diffusion of radionuclide from inside of the solidified waste.The results of the sorption experiments on Cs, Co, Sr for the rock samples from four regions, showed that Kd value of F-site quartz-porphery was the lowest, and that of B-site Rhyolitic tuff the highest. E-site granite, M-site granite gneiss showed medium sorption capacities. The Kd values of each radionuclide were in the following order: Cs > Co > Sr . For the radionuclide migration through fractured rock, examination of hydraulic dispersion model and channeling model was carried out to simulate the dispersion phenomena through the fractured rock. In part III, optimum model was established which allows mathematical prediction of the container corrosion rate and leach rate.Based on this model, the equations were obtained to predict corrosion rate and radionuclide leach rate. A good agreement of these equations and existing experimental data was considered to be an indication that it can be used effectively for the estimation of long-term leaching. (Author)

  4. Safety assessment of historical masonry churches based on pre-assigned kinematic limit analysis, FE limit and pushover analyses

    Energy Technology Data Exchange (ETDEWEB)

    Milani, Gabriele, E-mail: milani@stru.polimi.it; Valente, Marco, E-mail: milani@stru.polimi.it [Department of Architecture, Built Environment and Construction Engineering (ABC), Politecnico di Milano, Piazza Leonardo da Vinci 32, 20133 Milan (Italy)

    2014-10-06

    This study presents some results of a comprehensive numerical analysis on three masonry churches damaged by the recent Emilia-Romagna (Italy) seismic events occurred in May 2012. The numerical study comprises: (a) pushover analyses conducted with a commercial code, standard nonlinear material models and two different horizontal load distributions; (b) FE kinematic limit analyses performed using a non-commercial software based on a preliminary homogenization of the masonry materials and a subsequent limit analysis with triangular elements and interfaces; (c) kinematic limit analyses conducted in agreement with the Italian code and based on the a-priori assumption of preassigned failure mechanisms, where the masonry material is considered unable to withstand tensile stresses. All models are capable of giving information on the active failure mechanism and the base shear at failure, which, if properly made non-dimensional with the weight of the structure, gives also an indication of the horizontal peak ground acceleration causing the collapse of the church. The results obtained from all three models indicate that the collapse is usually due to the activation of partial mechanisms (apse, façade, lateral walls, etc.). Moreover the horizontal peak ground acceleration associated to the collapse is largely lower than that required in that seismic zone by the Italian code for ordinary buildings. These outcomes highlight that structural upgrading interventions would be extremely beneficial for the considerable reduction of the seismic vulnerability of such kind of historical structures.

  5. Safety assessment of historical masonry churches based on pre-assigned kinematic limit analysis, FE limit and pushover analyses

    International Nuclear Information System (INIS)

    Milani, Gabriele; Valente, Marco

    2014-01-01

    This study presents some results of a comprehensive numerical analysis on three masonry churches damaged by the recent Emilia-Romagna (Italy) seismic events occurred in May 2012. The numerical study comprises: (a) pushover analyses conducted with a commercial code, standard nonlinear material models and two different horizontal load distributions; (b) FE kinematic limit analyses performed using a non-commercial software based on a preliminary homogenization of the masonry materials and a subsequent limit analysis with triangular elements and interfaces; (c) kinematic limit analyses conducted in agreement with the Italian code and based on the a-priori assumption of preassigned failure mechanisms, where the masonry material is considered unable to withstand tensile stresses. All models are capable of giving information on the active failure mechanism and the base shear at failure, which, if properly made non-dimensional with the weight of the structure, gives also an indication of the horizontal peak ground acceleration causing the collapse of the church. The results obtained from all three models indicate that the collapse is usually due to the activation of partial mechanisms (apse, façade, lateral walls, etc.). Moreover the horizontal peak ground acceleration associated to the collapse is largely lower than that required in that seismic zone by the Italian code for ordinary buildings. These outcomes highlight that structural upgrading interventions would be extremely beneficial for the considerable reduction of the seismic vulnerability of such kind of historical structures

  6. Development of a hybrid safety system: Actuation of the secondary automatic depressurization system at an early stage

    International Nuclear Information System (INIS)

    Nishimoto, Masae; Umezawa, Shigemitsu; Okabe, Kazuharu; Matsuoka, Tsuyoshi

    1996-01-01

    A Hybrid Safety System, which is an optimum combination of active and passive safety systems, has been developed in order to improve the safety, reliability and economic features of the next generation of PWRs. The passive safety systems include Automatic primary Depressurization System (ADS), Secondary Automatic Depressurization System (SADS), advanced accumulators, gravity injection system and so on. In this study the authors have improved the actuation logic of the passive safety systems. The original logic in the previous study actuates ADS at an early stage of an event such as a Loss-of-Coolant Accident (LOCA), and this is followed by the actuation of SADS. In this study they divide SADS into two systems. The first, small SADS, uses small valves corresponding to the relief valves of the conventional PWR plants. The second, large SADS, corresponds to the original SADS using multiple valves of large capacity. With the new logic, the passive systems are actuated during a typical small LOCA. Small LOCA analyses using several break areas were performed for a 1,400 MWe PWR plant with a Hybrid Safety System. The results predict that core uncovery does not occur in the case of a relatively small break area and that core heat removal during a small LOCA is improved in comparison with the analyses for conventional PWR plants, where the secondary pressure remains higher during the event. The results also predict that this new logic make it possible to reduce the ADS valve size and the actuation pressure setpoint of the passive safety systems

  7. A Systematic Analysis of Functional Safety Certification Practices in Industrial Robot Software Development

    Directory of Open Access Journals (Sweden)

    Tong Xie

    2017-01-01

    Full Text Available For decades, industry robotics have delivered on the promise of speed, efficiency and productivity. The last several years have seen a sharp resurgence in the orders of industrial robots in China, and the areas addressed within industrial robotics has extended into safety-critical domains. However, safety standards have not yet been implemented widely in academia and engineering applications, particularly in robot software development. This paper presents a systematic analysis of functional safety certification practices in software development for the safety-critical software of industrial robots, to identify the safety certification practices used for the development of industrial robots in China and how these practices comply with the safety standard requirements. Reviewing from Chinese academic papers, our research shows that safety standards are barely used in software development of industrial robot. The majority of the papers propose various solutions to achieve safety, but only about two thirds of the papers refer to non-standardized approaches that mainly address the systematic level rather than the software development level. In addition, our research shows that with the development of artificial intelligent, an emerging field is still on the quest for standardized and suitable approaches to develop safety-critical software.

  8. Geostatistical analyses of communication routes in a geo-strategic and regional development perspective

    Directory of Open Access Journals (Sweden)

    Alexandru-Ionuţ Petrişor

    2017-12-01

    Full Text Available Accessibility is a key concept in regional development, with numerous ties to territorial cohesion and polycentricity. Moreover, it also exhibits a geo-strategic function, anchored in the international relationships between countries and continents. The article reviews several case studies, placing analyses of the Romanian accessibility in a broader context. The results show that regional development, overall EU connectivity and possible transit fluxes are prevented by the configuration or lack of communication routes. Increasing the accessibility of regions must be a priority of governments, regardless of political opinions. It is expected that the transition of economy to post-carbon era or other models – green economy, knowledge-based economy etc. – to result into the emergence of new poles and axes of development, and ensure transport sustainability.

  9. Quantitative Evaluation of the Effect on System Safety Engineer Training Course for the Aerospace Development

    Science.gov (United States)

    Sekita, Ryuichi; Yamada, Shu

    The system safety has been being applied in Japan Aerospace Exploration Agency (JAXA) ‧s launch vehicle and satellite development projects. The engineering state of system safety has some room for improvement. Therefore, JAXA is continuously working for system safety improvement. The system safety engineer training course is the top priority for the improvement. This paper represents the practical training evaluation way using Kirkpatrick‧s 4-level approach and the actual results in JAXA system safety engineer training course. Also this paper represents the importance of the engineer training evaluation as a part of PDCA cycle in the industry field.

  10. Progress in the development of methodology for fusion safety systems studies

    International Nuclear Information System (INIS)

    Ho, S.K.; Cambi, G.; Ciattaglia, S.; Fujii-e, Y.; Seki, Y.

    1994-01-01

    The development of fusion safety systems-study methodology, including the aspects of schematic classification of overall fusion safety system, qualitative assessment of fusion system for identification of critical accident scenarios, quantitative analysis of accident consequences and risk for safety design evaluation, and system-level analysis of accident consequences and risk for design optimization, by a consortium of international efforts is presented. The potential application of this methodology into reactor design studies will facilitate the systematic assessment of safety performance of reactor designs and enhance the impacts of safety considerations on the selection of design configurations

  11. 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.

  12. ADAPTER: Analysing and developing adaptability and performance in teams to enhance resilience

    International Nuclear Information System (INIS)

    Beek, Dolf van der; Schraagen, Jan Maarten

    2015-01-01

    In the current study, the concept of team resilience was operationalized by developing a first version of a questionnaire (ADAPTER) driven by the four essential abilities of resilience (Hollnagel E, 2011, Resilience engineering in practice: a guidebook, p. 275–96) and expanded with more relation-oriented abilities of leadership and cooperation. The development and administration of ADAPTER took place within two companies. Factor analyses using data of 91 participants largely supported the hypothesized 6-dimension taxonomy. Support was found for Team responding behavior, Shared Leadership and Cooperation with other teams/departments. Anticipation showed considerable overlap with the monitoring scale, possibly due to the fact that monitoring items dealt with prospective situations. Using ADAPTER questionnaire results as a starting point for further in-depth discussion among the different teams in the pilot companies proved very useful. Suggestions for future research include contextualizing the questionnaire by embedding it in actual cases or having it filled in after specific incidents. Also, support of organization should be included as a separate dimension in ADAPTER. - Highlights: • Development of a team resilience questionnaire (ADAPTER). • Driven by Hollnagel's resilience abilities plus shared leadership and cooperation. • Pilot testing of ADAPTER took place within two companies. • Factor analyses (N=91) largely supported the hypothesized 6-dimension taxonomy. • Results provide a useful starting point for further in-depth discussions

  13. Integrated Process Model Development and Systems Analyses for the LIFE Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Meier, W R; Anklam, T; Abbott, R; Erlandson, A; Halsey, W; Miles, R; Simon, A J

    2009-07-15

    We have developed an integrated process model (IPM) for a Laser Inertial Fusion-Fission Energy (LIFE) power plant. The model includes cost and performance algorithms for the major subsystems of the plant, including the laser, fusion target fabrication and injection, fusion-fission chamber (including the tritium and fission fuel blankets), heat transfer and power conversion systems, and other balance of plant systems. The model has been developed in Visual Basic with an Excel spreadsheet user interface in order to allow experts in various aspects of the design to easily integrate their individual modules and provide a convenient, widely accessible platform for conducting the system studies. Subsystem modules vary in level of complexity; some are based on top-down scaling from fission power plant costs (for example, electric plant equipment), while others are bottom-up models based on conceptual designs being developed by LLNL (for example, the fusion-fission chamber and laser systems). The IPM is being used to evaluate design trade-offs, do design optimization, and conduct sensitivity analyses to identify high-leverage areas for R&D. We describe key aspects of the IPM and report on the results of our systems analyses. Designs are compared and evaluated as a function of key design variables such as fusion target yield and pulse repetition rate.

  14. The quantifying of road safety developments. Paper presented at the International Conference 'Road Safety in Europe', Birmingham, September 9-11, 1996.

    NARCIS (Netherlands)

    Koornstra, M.J.

    1996-01-01

    The evaluation of the effectiveness of road safety policies and measures must be based on quantative information in road safety developments and the relevant variables that influence that development. However, the concept of road safety itself is not well defined theoretically and quantatively. This

  15. The safety analysis and thermohydraulic methodologies for the power updating analyses in Spanish PWR plants; Methodologias de diseno termohidraulico y de analisis de seguridad en los aumentos de potencia de centrales PWR

    Energy Technology Data Exchange (ETDEWEB)

    Salesa, F.

    2014-02-01

    This article describes the Safety Analysis and Thermohydraulic methodologies used by ENUSA for the Power Updating analyses in Spanish PWR plants of Westinghouse design: Design tools have been developed over the first cycles resulting new correlations of DNB, fitted to the new fuel assemblies, new DNBR calculation methodology and other improvements in the design areas. Using these methodologies, the available margins between design and limit values are wider. These new margins have allowed to accomplish the design criteria under the new power updating operational conditions. (Author)

  16. Development of the Continued Improvement System for Nuclear Safety Culture

    International Nuclear Information System (INIS)

    Park, H. C.; Park, H. G.; Park, Y. W.; Park, J. Y.

    2016-01-01

    It has been found that almost 80 % of the incidents and accidents occurred recently, such as the Fukushima Daiichi disaster and Domestic SBO accident etc. were analyzed to be caused from human errors. (IAEA NES NG-G-2.1) Which strongly claims the importance of the safety culture system. Accordingly, it should be away from a cursory approach like one-off field survey or Snap shop which were being conducted at present for the continued improvement of safety culture. This study introduces an analytical methodology which approaches the generic form of the safety both consciously and unconsciously expressed with behavior, thoughts, and attitude etc. This study was implemented only for open materials such as Inspection report, incidents and accidents reports, QA documents because of the limitation in accessibility to data. More effective use with securing operational data will be possible in future

  17. Integrated safety case development for deep geological repositories

    International Nuclear Information System (INIS)

    Kawamura, Hideki; McKinley, Ian G.

    2008-01-01

    The paper will illustrate an 'integrated safety case', which involves combining both pre-closure and post-closure safety arguments from the point of view of a repository implementer, who must also ensure that projects are practical, acceptable and economic. The post-closure safety case is based on the performance of a number of barriers, which are established during construction, operation and closure. Such barriers must be confirmed using quality assured methods, supported, as required, by inspection and monitoring. The requirement for integrated assessment means that even the final process to end institutional control and transfer any liabilities from the implementer needs to be considered at present, even though this will undoubtedly be refined and tailored to the site characteristics over the many decades that will pass before this occurs. To illustrate the practical application of this approach, assessment of variants for remote-handled emplacement of the EBS for disposal of HLW in Japan will be discussed. (author)

  18. Fuel cycle safety research. Role and development in future

    International Nuclear Information System (INIS)

    Matsumoto, Shiro

    2005-01-01

    The report stresses important roles played by risk informed methodology in chemical process safety problems as encountered in fuel cycle facility such as a reprocessing plant. Abnormal situation management is a safety issue - The lesson learned from the pesticide manufacturing facility in Bhopal, India. Little attention has been given to understanding the issues regarding performance during normal versus abnormal situation. The first step is in abnormal situation management is to define what really is abnormal. The second step is to ensure that everyone understands the difference between normal and abnormal, and the root causes of abnormal events. The third step is to be aware of current practices that support abnormal situation management, and the procedures, practices, and techniques used to respond to abnormal conditions. Dynamic simulation will become to useful techniques for abnormal situation management as well as probabilistic safety assessment for process facilities including nuclear fuel cycle. (S. Ohno)

  19. Development of an instrument to measure seniors' patient safety health beliefs: the Seniors Empowerment and Advocacy in Patient Safety (SEAPS) survey.

    Science.gov (United States)

    Elder, Nancy C; Regan, Saundra L; Pallerla, Harini; Levin, Linda; Post, Douglas; Cegela, Donald J

    2007-12-01

    To develop a survey to measure seniors' embracement of ambulatory patient safety self-advocacy behaviors, the Senior Empowerment and Advocacy in Patient Safety (SEAPS) survey. Content was developed by review of published recommendations combined with interviews and focus groups with community members; items were generated for subscales based on the health belief model (HBM). Psychometric characteristics were assessed by cluster and correlation analyses on a pilot test of 143 community dwelling seniors; the ability of the subscales and demographic variables to predict reported behavior was investigated by multiple regression. The four subscales of the SEAPS were outcome efficacy (OE), attitudes (ATT), self-efficacy (SE) and behaviors (BEH). Cronbach alphas were 0.74 for ATT, 0.79 for BEH, and 0.91 for OE and SE. Analysis of variance showed that there were no differences in any subscale score by race, education level or frequency of doctor visits, but women were noted to have significantly higher scores (pprograms aimed at improving seniors' self-advocacy skills. Effective interventions may improve the involvement of patients in their own safety in the clinical setting.

  20. Cultivating and Development — 30 Years Practice of Safety Culture in China

    International Nuclear Information System (INIS)

    Hu, L.; Zhang, Y.; Zhang, W.; Xu, G.

    2016-01-01

    The safety culture has been cultivated and promoted in China since its very beginning by IAEA. The 1st stage—stage of start and exploration—was from 1984 to 2007, in which the international concept of safety culture was imported and studied, with the process of combination and convergence with the positive elements of Chinese traditional culture. The basic ideas, such as the principles and directing ideas for the nuclear safety, were established in China. The 2nd stage — stage of practice and growing — was from 2007 to 2014, where safety culture was promoted by the Government, and the regulatory body NNSA established its basic supervision value based on the safety culture. The Chinese nuclear industry was encouraged to develop their of safety culture in a vivid form of presenting. The 3rd stage — stage of fast development — is from 2014 to now. The Chinese president Xi announce the Chinese Nuclear Safety View in The Hague in March 2014, showing the states position regarding the nuclear safety and safety culture. The policy declaration was issued and the nuclear safety promotion special action was carried out by NNSA. Safety culture is widely accepted and acknowledged by the nuclear and radioactivity relevant industry. (author)

  1. Developing an industry-oriented safety curriculum using the Delphi technique.

    Science.gov (United States)

    Chen, Der-Fa; Wu, Tsung-Chih; Chen, Chi-Hsiang; Chang, Shu-Hsuan; Yao, Kai-Chao; Liao, Chin-Wen

    2016-09-01

    In this study, we examined the development of industry-oriented safety degree curricula at a college level. Based on a review of literature on the practices and study of the development of safety curricula, we classified occupational safety and health curricula into the following three domains: safety engineering, health engineering, and safety and health management. We invited 44 safety professionals to complete a four-round survey that was designed using a modified Delphi technique. We used Chi-square statistics to test the panel experts' consensus on the significance of the items in the three domains and employed descriptive statistics to rank the participants' rating of each item. The results showed that the top three items for each of the three domains were Risk Assessment, Dangerous Machinery and Equipment, and Fire and Explosion Prevention for safety engineering; Ergonomics, Industrial Toxicology, and Health Risk Assessment for health engineering; and Industrial Safety and Health Regulations, Accident Investigation and Analysis, and Emergency Response for safety and health management. Only graduates from safety programmes who possess practical industry-oriented abilities can satisfy industry demands and provide value to the existence of college safety programmes.

  2. Analysing the nexus of sustainable development and climate change. An overview

    International Nuclear Information System (INIS)

    Munasinghe, M.

    2003-01-01

    This document is an output from the OECD Development and Climate Change project, an activity being jointly overseen by the (Environment Policy Committee) Working Party on Global and Structural Policies (WPGSP), and the DAC (Development Assistance Committee) Working Party on Development Co-operation and Environment (WPENV). The overall objective of the project is to provide guidance on how to mainstream responses to climate change within economic development planning and assistance policies, with natural resource management as an overarching theme. This paper sketches out a broad framework to address the nexus of sustainable development and climate change. It also draws out some implications for the preparation of future case studies aimed at exploring the dynamics of climate change vulnerability and adaptation - especially when one goes beyond simple win-win outcomes, and confronts difficult trade-off situations among conflicting objectives. Section 2 introduces the concept of sustainable development; Section 3 links that concept to climate change. In section 4, tools and methods of integrating and analysing the social, economic, and environmental dimensions of this nexus are briefly presented. These ideas are illustrated in section 5, by applying them to specific examples involving climate-related problems across the full range of spatial scales - at the global, national-economy-wide, sub-national-sectoral, and local-project levels. Section 6 contains some concluding thoughts and a discussion of implications for case studies

  3. Modelling software failures of digital I and C in probabilistic safety analyses based on the TELEPERM registered XS operating experience

    International Nuclear Information System (INIS)

    Jockenhoevel-Barttfeld, Mariana; Taurines Andre; Baeckstroem, Ola; Holmberg, Jan-Erik; Porthin, Markus; Tyrvaeinen, Tero

    2015-01-01

    Digital instrumentation and control (I and C) systems appear as upgrades in existing nuclear power plants (NPPs) and in new plant designs. In order to assess the impact of digital system failures, quantifiable reliability models are needed along with data for digital systems that are compatible with existing probabilistic safety assessments (PSA). The paper focuses on the modelling of software failures of digital I and C systems in probabilistic assessments. An analysis of software faults, failures and effects is presented to derive relevant failure modes of system and application software for the PSA. The estimations of software failure probabilities are based on an analysis of the operating experience of TELEPERM registered XS (TXS). For the assessment of application software failures the analysis combines the use of the TXS operating experience at an application function level combined with conservative engineering judgments. Failure probabilities to actuate on demand and of spurious actuation of typical reactor protection application are estimated. Moreover, the paper gives guidelines for the modelling of software failures in the PSA. The strategy presented in this paper is generic and can be applied to different software platforms and their applications.

  4. Development of safety culture at Kozloduy NPP: The role of the regulatory body

    International Nuclear Information System (INIS)

    Miliovsky, Ventzislav

    2002-01-01

    An evolution is made upon external factor that accompany the development of safety culture at Kozloduy NPP (KNPP). The two basic stages and the major results achieved independently and with external assistance were considered in the process of development of safety culture. The areas in which the regulatory body insists on having active participation were defined. (author)

  5. Noise exposure during pregnancy, birth outcomes and fetal development: meta-analyses using quality effects model.

    Science.gov (United States)

    Dzhambov, Angel M; Dimitrova, Donka D; Dimitrakova, Elena D

    2014-01-01

    Many women are exposed daily to high levels of occupational and residential noise, so the effect of noise exposure on pregnancy should be considered because noise affects both the fetus and the mother herself. However, there is a controversy in the literature regarding the adverse effects of occupational and residential noise on pregnant women and their fetuses. The aim of this study was to conduct systematic review of previously analyzed studies, to add additional information omitted in previous reviews and to perform meta-analyses on the effects of noise exposure on pregnancy, birth outcomes and fetal development. Previous reviews and meta-analyses on the topic were consulted. Additionally, a systematic search in MEDLINE, EMBASE and Internet was carried out. Twenty nine studies were included in the meta-analyses. Quality effects meta-analytical model was applied. Women exposed to high noise levels (in most of the studies ≥ 80 dB) during pregnancy are at a significantly higher risk for having small-for-gestational-age newborn (RR = 1.19, 95% CI: 1.03, 1.38), gestational hypertension (RR = 1.27, 95% CI: 1.03, 1.58) and infant with congenital malformations (RR = 1.47, 95% CI: 1.21, 1.79). The effect was not significant for preeclampsia, perinatal death, spontaneous abortion and preterm birth. The results are consistent with previous findings regarding a higher risk for small-for-gestational-age. They also highlight the significance of residential and occupational noise exposure for developing gestational hypertension and especially congenital malformations.

  6. Vaccine safety monitoring systems in developing countries: an example of the Vietnam model.

    Science.gov (United States)

    Ali, Mohammad; Rath, Barbara; Thiem, Vu Dinh

    2015-01-01

    Only few health intervention programs have been as successful as vaccination programs with respect to preventing morbidity and mortality in developing countries. However, the success of a vaccination program is threatened by rumors and misunderstanding about the risks of vaccines. It is short-sighted to plan the introduction of vaccines into developing countries unless effective vaccine safety monitoring systems are in place. Such systems that track adverse events following immunization (AEFI) is currently lacking in most developing countries. Therefore, any rumor may affect the entire vaccination program. Public health authorities should implement the safety monitoring system of vaccines, and disseminate safety issues in a proactive mode. Effective safety surveillance systems should allow for the conduct of both traditional and alternative epidemiologic studies through the use of prospective data sets. The vaccine safety data link implemented in Vietnam in mid-2002 indicates that it is feasible to establish a vaccine safety monitoring system for the communication of vaccine safety in developing countries. The data link provided the investigators an opportunity to evaluate AEFI related to measles vaccine. Implementing such vaccine safety monitoring system is useful in all developing countries. The system should be able to make objective and clear communication regarding safety issues of vaccines, and the data should be reported to the public on a regular basis for maintaining their confidence in vaccination programs.

  7. Developing and analysing a curriculum map in Occupational- and Environmental Medicine.

    Science.gov (United States)

    Hege, Inga; Nowak, Dennis; Kolb, Stefanie; Fischer, Martin R; Radon, Katja

    2010-09-14

    During the last 5 years a fundamental curriculum reform was realized at the medical school of the Ludwig-Maximilians-University. In contrast to those efforts, the learning objectives were not defined consistently for the curriculum and important questions concerning the curriculum could not be answered. This also applied to Occupational and Environmental Medicine where teachers of both courses were faced with additional problems such as the low number of students attending the lectures.The aims of the study were to develop and analyse a curriculum map for Occupational and Environmental Medicine based on learning objectives using a web-based database.Furthermore we aimed to evaluate student perception about the curricular structure. Using a web-based learning objectives database, a curriculum map for Occupational and Environmental Medicine was developed and analysed. Additionally online evaluations of students for each course were conducted. The results show a discrepancy between the taught and the assessed curriculum. For both curricula, we identified that several learning objectives were not covered in the curriculum. There were overlaps with other content domains and redundancies within both curricula. 53% of the students in Occupational Medicine and 43% in Environmental Medicine stated that there is a lack of information regarding the learning objectives of the curriculum. The results of the curriculum mapping and the poor evaluation results for the courses suggest a need for re-structuring both curricula.

  8. Developing and analysing a curriculum map in Occupational- and Environmental Medicine

    Directory of Open Access Journals (Sweden)

    Hege Inga

    2010-09-01

    Full Text Available Abstract Background During the last 5 years a fundamental curriculum reform was realized at the medical school of the Ludwig-Maximilians-University. In contrast to those efforts, the learning objectives were not defined consistently for the curriculum and important questions concerning the curriculum could not be answered. This also applied to Occupational and Environmental Medicine where teachers of both courses were faced with additional problems such as the low number of students attending the lectures. The aims of the study were to develop and analyse a curriculum map for Occupational and Environmental Medicine based on learning objectives using a web-based database. Furthermore we aimed to evaluate student perception about the curricular structure. Methods Using a web-based learning objectives database, a curriculum map for Occupational and Environmental Medicine was developed and analysed. Additionally online evaluations of students for each course were conducted. Results The results show a discrepancy between the taught and the assessed curriculum. For both curricula, we identified that several learning objectives were not covered in the curriculum. There were overlaps with other content domains and redundancies within both curricula. 53% of the students in Occupational Medicine and 43% in Environmental Medicine stated that there is a lack of information regarding the learning objectives of the curriculum. Conclusions The results of the curriculum mapping and the poor evaluation results for the courses suggest a need for re-structuring both curricula.

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

    African Journals Online (AJOL)

    2012-12-01

    Dec 1, 2012 ... great vacuum of Communication Bridge between the daily road users and the. Federal Road Safety Corps (FRSC). There is an urgent need therefore to fill this gap. The inability for fast, efficient and accurate decision making through provision of adequate traffic situation information to road users (motorist ...

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

    African Journals Online (AJOL)

    2012-12-01

    Dec 1, 2012 ... 1.62 Indian hemp. The role of poor sight in accident causation has also been well documented. With the high rate of road mishaps and road congestion, there lies a great vacuum of Communication Bridge between the daily road users and the. Federal Road Safety Corps (FRSC). There is an urgent need ...

  11. Developing Probabilistic Safety Performance Margins for Unknown and Underappreciated Risks

    Science.gov (United States)

    Benjamin, Allan; Dezfuli, Homayoon; Everett, Chris

    2015-01-01

    Probabilistic safety requirements currently formulated or proposed for space systems, nuclear reactor systems, nuclear weapon systems, and other types of systems that have a low-probability potential for high-consequence accidents depend on showing that the probability of such accidents is below a specified safety threshold or goal. Verification of compliance depends heavily upon synthetic modeling techniques such as PRA. To determine whether or not a system meets its probabilistic requirements, it is necessary to consider whether there are significant risks that are not fully considered in the PRA either because they are not known at the time or because their importance is not fully understood. The ultimate objective is to establish a reasonable margin to account for the difference between known risks and actual risks in attempting to validate compliance with a probabilistic safety threshold or goal. In this paper, we examine data accumulated over the past 60 years from the space program, from nuclear reactor experience, from aircraft systems, and from human reliability experience to formulate guidelines for estimating probabilistic margins to account for risks that are initially unknown or underappreciated. The formulation includes a review of the safety literature to identify the principal causes of such risks.

  12. SAFETY

    CERN Multimedia

    M. Plagge, C. Schaefer and N. Dupont

    2013-01-01

    Fire Safety – Essential for a particle detector The CMS detector is a marvel of high technology, one of the most precise particle measurement devices we have built until now. Of course it has to be protected from external and internal incidents like the ones that can occur from fires. Due to the fire load, the permanent availability of oxygen and the presence of various ignition sources mostly based on electricity this has to be addressed. Starting from the beam pipe towards the magnet coil, the detector is protected by flooding it with pure gaseous nitrogen during operation. The outer shell of CMS, namely the yoke and the muon chambers are then covered by an emergency inertion system also based on nitrogen. To ensure maximum fire safety, all materials used comply with the CERN regulations IS 23 and IS 41 with only a few exceptions. Every piece of the 30-tonne polyethylene shielding is high-density material, borated, boxed within steel and coated with intumescent (a paint that creates a thick co...

  13. Hybrid and Parallel Domain-Decomposition Methods Development to Enable Monte Carlo for Reactor Analyses

    International Nuclear Information System (INIS)

    Wagner, John C.; Mosher, Scott W.; Evans, Thomas M.; Peplow, Douglas E.; Turner, John A.

    2010-01-01

    This paper describes code and methods development at the Oak Ridge National Laboratory focused on enabling high-fidelity, large-scale reactor analyses with Monte Carlo (MC). Current state-of-the-art tools and methods used to perform real commercial reactor analyses have several undesirable features, the most significant of which is the non-rigorous spatial decomposition scheme. Monte Carlo methods, which allow detailed and accurate modeling of the full geometry and are considered the gold standard for radiation transport solutions, are playing an ever-increasing role in correcting and/or verifying the deterministic, multi-level spatial decomposition methodology in current practice. However, the prohibitive computational requirements associated with obtaining fully converged, system-wide solutions restrict the role of MC to benchmarking deterministic results at a limited number of state-points for a limited number of relevant quantities. The goal of this research is to change this paradigm by enabling direct use of MC for full-core reactor analyses. The most significant of the many technical challenges that must be overcome are the slow, non-uniform convergence of system-wide MC estimates and the memory requirements associated with detailed solutions throughout a reactor (problems involving hundreds of millions of different material and tally regions due to fuel irradiation, temperature distributions, and the needs associated with multi-physics code coupling). To address these challenges, our research has focused on the development and implementation of (1) a novel hybrid deterministic/MC method for determining high-precision fluxes throughout the problem space in k-eigenvalue problems and (2) an efficient MC domain-decomposition (DD) algorithm that partitions the problem phase space onto multiple processors for massively parallel systems, with statistical uncertainty estimation. The hybrid method development is based on an extension of the FW-CADIS method, which

  14. Hybrid and parallel domain-decomposition methods development to enable Monte Carlo for reactor analyses

    International Nuclear Information System (INIS)

    Wagner, J.C.; Mosher, S.W.; Evans, T.M.; Peplow, D.E.; Turner, J.A.

    2010-01-01

    This paper describes code and methods development at the Oak Ridge National Laboratory focused on enabling high-fidelity, large-scale reactor analyses with Monte Carlo (MC). Current state-of-the-art tools and methods used to perform 'real' commercial reactor analyses have several undesirable features, the most significant of which is the non-rigorous spatial decomposition scheme. Monte Carlo methods, which allow detailed and accurate modeling of the full geometry and are considered the 'gold standard' for radiation transport solutions, are playing an ever-increasing role in correcting and/or verifying the deterministic, multi-level spatial decomposition methodology in current practice. However, the prohibitive computational requirements associated with obtaining fully converged, system-wide solutions restrict the role of MC to benchmarking deterministic results at a limited number of state-points for a limited number of relevant quantities. The goal of this research is to change this paradigm by enabling direct use of MC for full-core reactor analyses. The most significant of the many technical challenges that must be overcome are the slow, non-uniform convergence of system-wide MC estimates and the memory requirements associated with detailed solutions throughout a reactor (problems involving hundreds of millions of different material and tally regions due to fuel irradiation, temperature distributions, and the needs associated with multi-physics code coupling). To address these challenges, our research has focused on the development and implementation of (1) a novel hybrid deterministic/MC method for determining high-precision fluxes throughout the problem space in k-eigenvalue problems and (2) an efficient MC domain-decomposition (DD) algorithm that partitions the problem phase space onto multiple processors for massively parallel systems, with statistical uncertainty estimation. The hybrid method development is based on an extension of the FW-CADIS method

  15. Development, implementation and evaluation of a pedestrian safety zone for elderly pedestrians

    Science.gov (United States)

    1998-02-01

    The objectives of this study were to develop and apply procedures for defining pedestrian safety zones for the older (age 65+) adult and to develop, implement and evaluate a countermeasure program in the defined zones. Zone definition procedures were...

  16. 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.

  17. Genome-wide analyses reveal a role for peptide hormones in planarian germline development.

    Directory of Open Access Journals (Sweden)

    James J Collins

    Full Text Available Bioactive peptides (i.e., neuropeptides or peptide hormones represent the largest class of cell-cell signaling molecules in metazoans and are potent regulators of neural and physiological function. In vertebrates, peptide hormones play an integral role in endocrine signaling between the brain and the gonads that controls reproductive development, yet few of these molecules have been shown to influence reproductive development in invertebrates. Here, we define a role for peptide hormones in controlling reproductive physiology of the model flatworm, the planarian Schmidtea mediterranea. Based on our observation that defective neuropeptide processing results in defects in reproductive system development, we employed peptidomic and functional genomic approaches to characterize the planarian peptide hormone complement, identifying 51 prohormone genes and validating 142 peptides biochemically. Comprehensive in situ hybridization analyses of prohormone gene expression revealed the unanticipated complexity of the flatworm nervous system and identified a prohormone specifically expressed in the nervous system of sexually reproducing planarians. We show that this member of the neuropeptide Y superfamily is required for the maintenance of mature reproductive organs and differentiated germ cells in the testes. Additionally, comparative analyses of our biochemically validated prohormones with the genomes of the parasitic flatworms Schistosoma mansoni and Schistosoma japonicum identified new schistosome prohormones and validated half of all predicted peptide-encoding genes in these parasites. These studies describe the peptide hormone complement of a flatworm on a genome-wide scale and reveal a previously uncharacterized role for peptide hormones in flatworm reproduction. Furthermore, they suggest new opportunities for using planarians as free-living models for understanding the reproductive biology of flatworm parasites.

  18. Numerical investigation on turbulent natural convection in partially connected cylindrical enclosures for analysing SFR safety under core meltdown scenario

    International Nuclear Information System (INIS)

    David, Dijo K.; Mangarjuna Rao, P.; Nashine, B.K.; Selvaraj, P.

    2015-01-01

    Under the unlikely event of severe core meltdown accident in pool type SFR, the molten core materials may rupture the grid plate which supports the fuel subassemblies and it can get relocated in to the lower pool. These debris may eventually settle on the debris collector (i.e., core catcher) installed above the bottom wall of the lower pool. The bed thus formed generates heat due to radioactive decay which has to be passively removed for maintaining the structural integrity of main vessel. By means of natural convection, the heat generated in the debris bed will be transferred to the top pool where the heat sink (i.e., Decay heat exchanger (DHX)) is installed. Heat transfer to the DHX (which is a part of safety grade decay heat removal system) can take place through the opening created in the grid plate which connects the two liquid pools (i.e., the top pool and the lower pool). Heat transfer can also take place through the lateral wall of the lower cylindrical pool to the side pool and eventually to the top pool, and thus to the DHX. This study numerically investigates the effectiveness of heat transfer between lower pool and top pool during PARR by considering them as partially connected cylindrical enclosures. The governing equations have been numerically solved using finite volume method in cylindrical co-ordinates using SIMPLE algorithm. Turbulence has been modeled using k-ω model and the model is validated against benchmark problems of natural convection found in literature. The effect of parameters such as the heat generation rate in the bed and the size of the grid plate opening are evaluated. Also PAHR in SFR pool is modeled using an axi-symmetric model to fund out the influence of grid plate opening on heat removal from core catcher. The results obtained are useful for improving the cooling capability of in-vessel tray type core catcher for handling the whole core meltdown scenarios in SFR. (author)

  19. The patient safety climate in healthcare organizations (PSCHO) survey: Short-form development.

    Science.gov (United States)

    Benzer, Justin K; Meterko, Mark; Singer, Sara J

    2017-08-01

    Measures of safety climate are increasingly used to guide safety improvement initiatives. However, cost and respondent burden may limit the use of safety climate surveys. The purpose of this study was to develop a 15- to 20-item safety climate survey based on the Patient Safety Climate in Healthcare Organizations survey, a well-validated 38-item measure of safety climate. The Patient Safety Climate in Healthcare Organizations was administered to all senior managers, all physicians, and a 10% random sample of all other hospital personnel in 69 private sector hospitals and 30 Veterans Health Administration hospitals. Both samples were randomly divided into a derivation sample to identify a short-form subset and a confirmation sample to assess the psychometric properties of the proposed short form. The short form consists of 15 items represented 3 overarching domains in the long-form scale-organization, work unit, and interpersonal. The proposed short form efficiently captures 3 important sources of variance in safety climate: organizational, work-unit, and interpersonal. The short-form development process was a practical method that can be applied to other safety climate surveys. This safety climate short form may increase response rates in studies that involve busy clinicians or repeated measures. Published 2017. This article is a U.S. Government work and is in the public domain in the USA.

  20. Key Factors Affecting Construction Safety Performance in Developing Countries: Evidence from Cambodia

    OpenAIRE

    Durdyev, Serdar; Mohamed, Sherif; Lay, Meng Leang; Ismail, Syuhaida

    2017-01-01

    Although proper safety management in construction is of utmost importance; anecdotal evidence suggests that safety is not adequately considered in many developing countries. This paper considers the key variables affecting construction safety performance in Cambodia. Using an empirical questionnaire survey targeting local construction professionals, respondents were invited to rate the level of importance of 30 variables identified from the seminal literature. The data set was subjected to f...

  1. Safety aspects of MOVPE in research and development: An example

    Science.gov (United States)

    Kaufmann, L. M. F.; Heuken, M.; Tilders, R.; Heime, K.; Jürgensen, H.; Heyen, M.

    1988-12-01

    In this paper active and passive safety aspects of the new MOVPE laboratory of the University of Duisburg are discussed. The laboratory is in use since march 1987. During normal operation a spatial separation of operators and MOVPE equipment ensures a maximum of safety. In case of emergency the MOVPE equipment is automatically switched to a safe set of operational conditions and alarms are activated. The gas supervision system interrupts the supply of the toxic gases near the gas tanks stored outside the building and the MOVPE equipment is purged by nitrogen. A wet-chemical scrubbing system removes the toxic gases from the process exhaust to a residual content below the detection limit (handling, use and measuring of toxic materials in semiconductor processes are discussed to avoid chronical poisoning.

  2. Developing of the System Safety Documentation for RSG-GAS

    International Nuclear Information System (INIS)

    Mariatmo; Korua, Johny A.

    2004-01-01

    P2TRR performs periodical evaluations on SAR documents to assess their conformance to the current field conditions. In case of non-conformance between the field conditions and SAR directly associated with the reactor safety, SAR documents will he immediately amended. Any changes in SAR documents will cause all documents associated with the reactor safety and operation be revised. Besides evaluation of field conditions, BAPETEN and IAEA also perform inspection. The results of BAPETEN and IAEA's inspection need to be seriously followed up because it is related to the reactor operation permit covered in the BAPETEN's authority. The RSG-GAS's SAR document needs to be revised based on the above considerations in addition to the inputs and suggestions from the issuer of reactor operation permit and other parties. (author)

  3. Development of Large-Scale Spacecraft Fire Safety Experiments

    DEFF Research Database (Denmark)

    Ruff, Gary A.; Urban, David L.; Fernandez-Pello, A. Carlos

    2013-01-01

    exploration missions outside of low-earth orbit and accordingly, more complex in terms of operations, logistics, and safety. This will increase the challenge of ensuring a fire-safe environment for the crew throughout the mission. Based on our fundamental uncertainty of the behavior of fires in low...... Corporation Cygnus vehicle after it has deberthed from the ISS. Although the experiment will need to meet rigorous safety requirements to ensure the carrier vehicle does not sustain damage, the absence of a crew removes the need for strict containment of combustion products. The tests will be fully automated...... with the data downlinked at the conclusion of the test before the Cygnus vehicle reenters the atmosphere. Several computer modeling and ground-based experiment efforts will complement the flight experiment effort. The international topical team is collaborating with the NASA team in the definition...

  4. Evaluation of methods and tools to develop safety concepts and to demonstrate safety for an HLW repository in salt. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Bollingerfehr, W.; Buhmann, D.; Doerr, S.; and others

    2017-03-15

    Salt formations have been the preferred option as host rocks for the disposal of high level radioactive waste in Germany for more than 40 years. During this period comprehensive geological investigations have been carried out together with a broad spectrum of concept and safety related R and D work. The behaviour of an HLW repository in salt formations, particularly in salt domes, has been analysed in terms of assessment of the total system performance. This was first carried out for concepts of generic waste repositories in salt and, since 1998, for a repository concept with specific boundary conditions, taking the geology of the Gorleben salt dome as an example. Suitable repository concepts and designs were developed, the technical feasibility has been proven and operational and long-term safety evaluated. Numerical modelling is an important input into the development of a comprehensive safety case for a waste repository. Significant progress in the development of numerical tools and their application for long-term safe ty assessment has been made in the last two decades. An integrated approach has been used in which the repository concept and relevant scientific and engineering data are combined with the results from iterative safety assessments to increase the clarity and the traceability of the evaluation. A safety concept that takes full credit of the favourable properties of salt formations was developed in the course of the R and D project ISIBEL, which started in 2005. This concept is based on the safe containment of radioactive waste in a specific part of the host rock formation, termed the containment providing rock zone, which comprises the geological barrier, the geotechnical barriers and the compacted backfill. The future evolution of the repository system will be analysed using a catalogue of Features, Events and Processes (FEP), scenario development and numerical analysis, all of which are adapted to suit the safety concept. Key elements of the

  5. Evaluation of methods and tools to develop safety concepts and to demonstrate safety for an HLW repository in salt. Final report

    International Nuclear Information System (INIS)

    Bollingerfehr, W.; Buhmann, D.; Doerr, S.

    2017-03-01

    Salt formations have been the preferred option as host rocks for the disposal of high level radioactive waste in Germany for more than 40 years. During this period comprehensive geological investigations have been carried out together with a broad spectrum of concept and safety related R and D work. The behaviour of an HLW repository in salt formations, particularly in salt domes, has been analysed in terms of assessment of the total system performance. This was first carried out for concepts of generic waste repositories in salt and, since 1998, for a repository concept with specific boundary conditions, taking the geology of the Gorleben salt dome as an example. Suitable repository concepts and designs were developed, the technical feasibility has been proven and operational and long-term safety evaluated. Numerical modelling is an important input into the development of a comprehensive safety case for a waste repository. Significant progress in the development of numerical tools and their application for long-term safe ty assessment has been made in the last two decades. An integrated approach has been used in which the repository concept and relevant scientific and engineering data are combined with the results from iterative safety assessments to increase the clarity and the traceability of the evaluation. A safety concept that takes full credit of the favourable properties of salt formations was developed in the course of the R and D project ISIBEL, which started in 2005. This concept is based on the safe containment of radioactive waste in a specific part of the host rock formation, termed the containment providing rock zone, which comprises the geological barrier, the geotechnical barriers and the compacted backfill. The future evolution of the repository system will be analysed using a catalogue of Features, Events and Processes (FEP), scenario development and numerical analysis, all of which are adapted to suit the safety concept. Key elements of the

  6. Support analysis for safety analysis development for CANDU nuclear power plant

    International Nuclear Information System (INIS)

    Bedreaga, L.; Florescu, Gh.; Apostol, M.; Nitoi, M.

    2004-01-01

    Probabilistic Safety Assessment analysis (PSA) is a technique used to assess the safety of a nuclear power plant. Assessments of the nuclear plant systems/components from safety point of view consist in accomplishment of a lot of support analyses that are the base for the main analysis, in order to evaluate the impact of occurrences of abnormal states for these systems. Evaluation of initiating events frequency and components failure rate is based on underlying probabilistic theory and mathematic statistics. Some of these analyses are detailed analyses and are known very well in PSA. There are also some analyses, named support analyses for PSA, which are very important but less applicable because they involve a huge human effort and hardware facilities to accomplish. The usual methods applicable in PSA such as input data extracted from the specific documentation (operation procedures, testing procedures, maintenance procedures and so on) or conservative evaluation provide a high level of uncertainty for both input and output data. The paper describes support analysis required to improve the certainty level in evaluation of reliability parameters and also in the final results (either risk, reliability or safety assessment). (author)

  7. Intelligence and handedness: Meta-analyses of studies on intellectually disabled, typically developing, and gifted individuals.

    Science.gov (United States)

    Papadatou-Pastou, Marietta; Tomprou, Dimitra-Maria

    2015-09-01

    Understanding the relationship between cerebral laterality and intelligence is important in elucidating the neurological underpinnings of individual differences in cognitive abilities. A widely used, behavioral indicator for cerebral laterality, mainly of language, is handedness. A number of studies have compared cognitive abilities between groups of left- and right-handers, while others have investigated the handedness prevalence between groups of different cognitive abilities. The present study comprises five meta-analyses of studies that have assessed the handedness prevalence in (a) individuals with intellectual disability (ID) of unknown/idiopathic nature compared to typically developing (TD) individuals, and (b) individuals with intellectual giftedness (IG) compared to TD individuals. Nineteen data sets totaling 16,076 participants (5795 ID, 8312 TD, and 1969 IG) were included in the analyses. Elevated levels of atypical handedness were found to be robust only for the ID to TD comparison. Findings constrain the range of acceptable theories on the handedness distribution for different intelligence levels. Copyright © 2015 Elsevier Ltd. All rights reserved.

  8. The Development of the Code Safety Valve Test Facility

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Chang Hyun; Kim, Young Ae; Park, Jong Woon [Korea Hydro and Nuclear Power Co., Ltd., Daejeon (Korea, Republic of)

    2007-07-01

    The Pressurizer Safety Valves (PSV) in Pressurized Water Reactors (PWRs) are required to provide the overpressure protection for the Reactor Coolant System (RCS) during the overpressure transients. According to the ASME OM code, all safety valves should be tested every 5 years with the acceptance tolerance of 1%. If one valve failed to meet this criterion, other two additional valves should be tested and if these valves don't meet the requirement, all valves should be tested. These frequent tests may make the valves decrepit and become a cause of leak. Therefore, increase of the acceptance tolerance is vital for the safe operation of the plant. In the United States, the acceptance tolerances are enlarged up to about 3% in most plants. This requires re-analysis of relevant accidents in FSAR. Also, the technical background data for the valve pop-up characteristics and the loop seal dynamics (if the plant has the loop seal in the upstream of PSV) are needed for the new safety analysis. Korea Hydro and Nuclear Power Company (KHNP) plans to build the PSV test facility for the purpose of providing the background data. This paper describes the preliminary design of the facility and studies on the system dynamics using GOTHIC-7.2a code to verify the pressure vessel capacities and to find the best operating condition.

  9. In vitro marker gene expression analyses in human peripheral blood mononuclear cells: A tool to assess safety of influenza vaccines in humans.

    Science.gov (United States)

    Sasaki, Eita; Momose, Haruka; Hiradate, Yuki; Ishii, Ken J; Mizukami, Takuo; Hamaguchi, Isao

    2018-12-01

    Vaccines are inoculated in healthy individuals from children to the elderly, and thus high levels of safety and consistency of vaccine quality in each lot must meet the required specifications by using preclinical and lot release testing. Because vaccines are inoculated into humans, recapitulation of biological reactions in humans should be considered for test methods. We have developed a new method to evaluate the safety of influenza vaccines using biomarker gene expression in mouse and rat models. Some biomarker genes are already known to be expressed in human lymphocytes, macrophages and dendritic cells; therefore, we considered some of these genes might be common biomarkers for human and mice to evaluate influenza vaccine safety. In this study, we used human peripheral blood mononuclear cells (PBMC) as a primary assessment tool to confirm the usefulness of potential marker genes in humans. Analysis of marker gene expression in PBMC revealed biomarker gene expressions were dose-relatedly increased in toxic reference influenza vaccine (RE)-stimulated PBMC. Although some marker genes showed increased expression in hemagglutinin split vaccine-stimulated PBMC, their expression levels were lower than that of RE in PBMC from two different donors. Many marker gene expressions correlated with chemokine production. Marker genes such as IRF7 were associated with other Type 1 interferon (IFN)-associated signals and were highly expressed in the CD304 + plasmacytoid dendritic cell (pDC) population. These results suggest PBMC and their marker genes may be useful for vaccine safety evaluation in humans.

  10. Developing a system dynamics model to analyse environmental problem in construction site

    Science.gov (United States)

    Haron, Fatin Fasehah; Hawari, Nurul Nazihah

    2017-11-01

    This study aims to develop a system dynamics model at a construction site to analyse the impact of environmental problem. Construction sites may cause damages to the environment, and interference in the daily lives of residents. A proper environmental management system must be used to reduce pollution, enhance bio-diversity, conserve water, respect people and their local environment, measure performance and set targets for the environment and sustainability. This study investigates the damaging impact normally occur during the construction stage. Environmental problem will cause costly mistake in project implementation, either because of the environmental damages that are likely to arise during project implementation, or because of modification that may be required subsequently in order to make the action environmentally acceptable. Thus, findings from this study has helped in significantly reducing the damaging impact towards environment, and improve the environmental management system performance at construction site.

  11. Development of the calibration transfer procedures for the XRF analyser AF-20

    CERN Document Server

    Kowalska, E

    2001-01-01

    Two standardization techniques; Direct (DS) and Picewise Direct Standardization (PDS) were applied to multivariate models developed using x-ray spectra. The data were obtained for the determination of Ca and fe in flying ash from brown coal using low resolution XRF analysis. The analyser AF-20 was used with an argon proportional counter and a Cd-109 radioisotope source. Number and comparison of samples from the calibration set were selected. Root mean square error of prediction (RMSEPr) was used as a criterion for assessment of the standardization quality. The best performance was obtained for the PDS method using 5 samples chosen with the leverage method. The RMSEPr value was 4 times lower for Ca determination and 2 times lower for Fe compared with the results without standardization. Software package was designed for fast computation of the standardization transform. The program can be used when the instrument parameters have changed as well as for calibration of the new instrument.

  12. Reading Ability Development from Kindergarten to Junior Secondary: Latent Transition Analyses with Growth Mixture Modeling

    Directory of Open Access Journals (Sweden)

    Yuan Liu

    2016-10-01

    Full Text Available The present study examined the reading ability development of children in the large scale Early Childhood Longitudinal Study (Kindergarten Class of 1998-99 data; Tourangeau, Nord, Lê, Pollack, & Atkins-Burnett, 2006 under the dynamic systems. To depict children's growth pattern, we extended the measurement part of latent transition analysis to the growth mixture model and found that the new model fitted the data well. Results also revealed that most of the children stayed in the same ability group with few cross-level changes in their classes. After adding the environmental factors as predictors, analyses showed that children receiving higher teachers' ratings, with higher socioeconomic status, and of above average poverty status, would have higher probability to transit into the higher ability group.

  13. A report on developing a checklist to assess company plans focused on improving safety awareness, safe behaviour and safety culture: final report

    NARCIS (Netherlands)

    Steijger, N.; Starren, H.; Keus, M.; Gort, J.; Vervoort, M.

    2003-01-01

    This report describes the process of developing a checklist to asses company plans focused on improving safety awareness, safe behaviour and safety culture. These plans are part of a programme initiated by the Ministry of Social Affairs and Employment aiming at improving the safety performance of

  14. Functional Hazard Analysis for Railway Safety

    OpenAIRE

    RAFRAFI, M; EL-KOURSI, EM

    2007-01-01

    The apportionment of railway safety targets is a key issue to develop a common safety management in the European railway system. In this paper, we develop a generic approach based on the Functional Hazard Analysis (FHA), to analyse the safety of railway systems for a unified European network and to comply with the Common Safety Targets (CSTs) required by the European railway safety directive. We suggest to combine the FHA technique with the functional railway architecture, developed by the AE...

  15. Testing a dual-systems model of adolescent brain development using resting-state connectivity analyses.

    Science.gov (United States)

    van Duijvenvoorde, A C K; Achterberg, M; Braams, B R; Peters, S; Crone, E A

    2016-01-01

    The current study aimed to test a dual-systems model of adolescent brain development by studying changes in intrinsic functional connectivity within and across networks typically associated with cognitive-control and affective-motivational processes. To this end, resting-state and task-related fMRI data were collected of 269 participants (ages 8-25). Resting-state analyses focused on seeds derived from task-related neural activation in the same participants: the dorsal lateral prefrontal cortex (dlPFC) from a cognitive rule-learning paradigm and the nucleus accumbens (NAcc) from a reward-paradigm. Whole-brain seed-based resting-state analyses showed an age-related increase in dlPFC connectivity with the caudate and thalamus, and an age-related decrease in connectivity with the (pre)motor cortex. nAcc connectivity showed a strengthening of connectivity with the dorsal anterior cingulate cortex (ACC) and subcortical structures such as the hippocampus, and a specific age-related decrease in connectivity with the ventral medial PFC (vmPFC). Behavioral measures from both functional paradigms correlated with resting-state connectivity strength with their respective seed. That is, age-related change in learning performance was mediated by connectivity between the dlPFC and thalamus, and age-related change in winning pleasure was mediated by connectivity between the nAcc and vmPFC. These patterns indicate (i) strengthening of connectivity between regions that support control and learning, (ii) more independent functioning of regions that support motor and control networks, and (iii) more independent functioning of regions that support motivation and valuation networks with age. These results are interpreted vis-à-vis a dual-systems model of adolescent brain development. Copyright © 2015. Published by Elsevier Inc.

  16. Development of a Culturally Informed Child Safety Curriculum for American Indian Families.

    Science.gov (United States)

    Berns, Ryan M; Tomayko, Emily J; Cronin, Kate A; Prince, Ronald J; Parker, Tassy; Adams, Alexandra K

    2017-04-01

    American Indian (AI) children are disproportionately affected by unintentional injuries, with injury mortality rates approximately 2.3 times higher than the combined rates for all children in the United States. Although multiple risk factors are known to contribute to these increased rates, a comprehensive, culturally informed curriculum that emphasizes child safety is lacking for this population. In response to this need, academic and tribal researchers, tribal community members, tribal wellness staff, and national child safety experts collaborated to develop a novel child safety curriculum. This paper describes its development and community delivery. We developed the safety curriculum as part of a larger randomized controlled trial known as Healthy Children, Strong Families 2 (HCSF2), a family-based intervention targeting obesity prevention in early childhood (2-5 years). During the development of the HCSF2 intervention, participating tribal communities expressed concern about randomizing enrolled families to a control group who would not receive an intervention. To address this concern and the significant disparities in injuries and unintentional death rates among AI children, we added an active control group (Safety Journey) that would utilize our safety curriculum. Satisfaction surveys administered at the 12-month time point of the intervention indicate 94% of participants (N = 196) were either satisfied or very satisfied with the child safety curriculum. The majority of participants (69%) reported spending more than 15 min with the curriculum materials each month, and 83% thought the child safety newsletters were either helpful or very helpful in making changes to improve their family's safety. These findings indicate these child safety materials have been well received by HCSF2 participants. The use of community-engaged approaches to develop this curriculum represents a model that could be adapted for other at-risk populations and serves as an initial

  17. Proposal for the establishment of an international project to analyse the safety of radioactive waste disposal into geological formations

    International Nuclear Information System (INIS)

    Anon.

    1977-01-01

    The discussion of the basic principles underlying waste disposal and the analysis of associated long-term risks appears to be areas where international co-operation offers great promise and meets with almost universal interest. An authoritative analysis and the development of an internationally agreed philosophy about waste disposal might even go a long way towards easing the apprehension expressed by part of the public in regard to nuclear energy

  18. Development of the switch requirements and architecture of a safety data communication system

    International Nuclear Information System (INIS)

    Jeong, K.I.; Lee, J.K.; Park, H.Y.; Koo, I.S.

    2004-12-01

    In accordance with digitalising the Instrumentation and Control(I and C) systems in the integral reactor, a communication network is required for effective information exchanges between the different equipment, an enhancement of the design flexibility, a simple installation and cost reduction. Generally, a communication network consists of a topology, the protocol, a communication medium, an interconnection device, etc. In this report, the development methods of switch and the architecture of a Safety Data Communication System(SDCS) are investigated and analyzed. In this report, the design requirements for switch are presented, which are the essential requirements to develop the switch in a SDCS of the SMART-P. To establish these requirements, the evaluation and analysis of the design and implementation method of the COTS switches, the architecture of SDCS and the design requirements of a SDCS were performed. At the detail design stage, these requirements will be used for the top-tier requirements, especially the design target and design basis. To develop the detail design requirements in the future, more quantitative and qualitative analyses are required. In the case of selecting the COTS switch and developing the switch, these requirements will also be used for the evaluation guide

  19. Combining ANOVA-PCA with POCHEMON to analyse micro-organism development in a polymicrobial environment.

    Science.gov (United States)

    Geurts, Brigitte P; Neerincx, Anne H; Bertrand, Samuel; Leemans, Manja A A P; Postma, Geert J; Wolfender, Jean-Luc; Cristescu, Simona M; Buydens, Lutgarde M C; Jansen, Jeroen J

    2017-04-22

    Revealing the biochemistry associated to micro-organismal interspecies interactions is highly relevant for many purposes. Each pathogen has a characteristic metabolic fingerprint that allows identification based on their unique multivariate biochemistry. When pathogen species come into mutual contact, their co-culture will display a chemistry that may be attributed both to mixing of the characteristic chemistries of the mono-cultures and to competition between the pathogens. Therefore, investigating pathogen development in a polymicrobial environment requires dedicated chemometric methods to untangle and focus upon these sources of variation. The multivariate data analysis method Projected Orthogonalised Chemical Encounter Monitoring (POCHEMON) is dedicated to highlight metabolites characteristic for the interaction of two micro-organisms in co-culture. However, this approach is currently limited to a single time-point, while development of polymicrobial interactions may be highly dynamic. A well-known multivariate implementation of Analysis of Variance (ANOVA) uses Principal Component Analysis (ANOVA-PCA). This allows the overall dynamics to be separated from the pathogen-specific chemistry to analyse the contributions of both aspects separately. For this reason, we propose to integrate ANOVA-PCA with the POCHEMON approach to disentangle the pathogen dynamics and the specific biochemistry in interspecies interactions. Two complementary case studies show great potential for both liquid and gas chromatography - mass spectrometry to reveal novel information on chemistry specific to interspecies interaction during pathogen development. Copyright © 2017 The Author(s). Published by Elsevier B.V. All rights reserved.

  20. SWOT analyses of the national energy sector for sustainable energy development

    International Nuclear Information System (INIS)

    Markovska, N.; Taseska, V.; Pop-Jordanov, J.

    2009-01-01

    A holistic perspective of various energy stakeholders regarding the Strengths, Weaknesses, Opportunities and Threats (SWOTs) of the energy sector in Macedonia is utilized as baseline to diagnose the current state and to sketch future action lines towards sustainable energy development. The resulting SWOT analyses pointed to the progressive adoption of European Union (EU) standards in energy policy and regulation as the most important achievement in the energy sector. The most important problems the national energy sector faces are scarce domestic resources and unfavorable energy mix, low electricity prices, a high degree of inefficiency in energy production and use, as well as insufficient institutional and human capacities. The formulated portfolio of actions towards enabling sustainable energy development urges the adoption of a comprehensive energy strategy built upon sustainability principles, intensified utilization of the natural gas, economic prices of electricity, structural changes in industry, promotion of energy efficiency and renewables, including Clean Development Mechanism (CDM) projects, enforcement of EU environmental standards and meeting the environmental requirements, as well as institutional and human capacity building.

  1. Comparative transcriptomic analyses of vegetable and grain pea (Pisum sativum L. seed development

    Directory of Open Access Journals (Sweden)

    Na eLiu

    2015-11-01

    Full Text Available Understanding the molecular mechanisms regulating pea seed developmental process is extremely important for pea breeding. In this study, we used high-throughput RNA-Seq and bioinformatics analyses to examine the changes in gene expression during seed development in vegetable pea and grain pea, and compare the gene expression profiles of these two pea types. RNA-Seq generated 18.7 G of raw data, which were then de novo assembled into 77,273 unigenes with a mean length of 930 bp. Our results illustrate that transcriptional control during pea seed development is a highly coordinated process. There were 459 and 801 genes differentially expressed at early and late seed maturation stages between vegetable pea and grain pea, respectively. Soluble sugar and starch metabolism related genes were significantly activated during the development of pea seeds coinciding with the onset of accumulation of sugar and starch in the seeds. A comparative analysis of genes involved in sugar and starch biosynthesis in vegetable pea (high seed soluble sugar and low starch and grain pea (high seed starch and low soluble sugar revealed that differential expression of related genes at late development stages results in a negative correlation between soluble sugar and starch biosynthetic flux in vegetable and grain pea seeds. RNA-Seq data was validated by using real-time quantitative RT-PCR analysis for 30 randomly selected genes. To our knowledge, this work represents the first report of seed development transcriptomics in pea. The obtained results provide a foundation to support future efforts to unravel the underlying mechanisms that control the developmental biology of pea seeds, and serve as a valuable resource for improving pea breeding.

  2. Safety culture giving impetus to the development of nuclear power enterprise

    International Nuclear Information System (INIS)

    Zhang Ying

    2011-01-01

    Jiangsu Nuclear Power Corporation (JNPC) have been continuously assimilating excellent nuclear safety culture at home and abroad and improving the plant safety operation and internal management level of corporation since the successful construction of Phase I project and the gradual success of the expansion project. Implemented the 'top management 8 expectations', executed the '3 into 1' (quality, environment and occupational health safety) management system. The culture of 'zero tolerance' has been deeply rooted. The safety culture brings people's heart closer, which is not only accepted by the employees, but also climbs up to a higher level and adds momentum to the scientific development of Tianwan Nuclear Power Station Base. (author)

  3. The Development of Severe Accident Codes at IRSN and Their Application to Support the Safety Assessment of EPR

    International Nuclear Information System (INIS)

    Caroli, Cataldo; Bleyer, Alexandre; Bentaib, Ahmed; Chatelard, Patrick; Cranga, Michel; Van Dorsselaere, Jean-Pierre

    2006-01-01

    IRSN uses a two-tier approach for development of codes analysing the course of a hypothetical severe accident (SA) in a Pressurized Water Reactor (PWR): on one hand, the integral code ASTEC, jointly developed by IRSN and GRS, for fast-running and complete analysis of a sequence; on the other hand, detailed codes for best-estimate analysis of some phenomena such as ICARE/CATHARE, MC3D (for steam explosion), CROCO and TONUS. They have been extensively used to support the level 2 Probabilistic Safety Assessment of the 900 MWe PWR and, in general, for the safety analysis of the French PWR. In particular the codes ICARE/CATHARE, CROCO, MEDICIS (module of ASTEC) and TONUS are used to support the safety assessment of the European Pressurized Reactor (EPR). The ICARE/CATHARE code system has been developed for the detailed evaluation of SA consequences in a PWR primary system. It is composed of the coupling of the core degradation IRSN code ICARE2 and of the thermal-hydraulics French code CATHARE2. The CFD code CROCO describes the corium flow in the spreading compartment. Heat transfer to the surrounding atmosphere and to the basemat, leading to the possible formation of an upper and lower crust, basemat ablation and gas sparging through the flow are modelled. CROCO has been validated against a wide experimental basis, including the CORINE, KATS and VULCANO programs. MEDICIS simulates MCCI (Molten-Corium-Concrete-Interaction) using a lumped-parameter approach. Its models are being continuously improved through the interpretation of most MCCI experiments (OECD-CCI, ACE...). The TONUS code has been developed by IRSN in collaboration with CEA for the analysis of the hydrogen risk (both distribution and combustion) in the reactor containment. The analyses carried out to support the EPR safety assessment are based on a CFD formulation. At this purpose a low-Mach number multi-component Navier-Stokes solver is used to analyse the hydrogen distribution. Presence of air, steam and

  4. Development, Dissemination, and Assessment of a Food Safety Systems Management Curriculum for Agribusiness Students in Armenia

    Science.gov (United States)

    Pokharel, Siroj; Marcy, Joseph E.; Neilan, Angela M.; Cutter, Catherine N.

    2017-01-01

    This study addresses the development, dissemination, and assessment of a Food Safety System Management (FSSM) curriculum offered to college-aged, agribusiness students in Yerevan, Armenia. Prior to beginning the program, demographic data were collected and a paper-based pretest was administered to access the food safety knowledge, behavior, and…

  5. What can nanosafety learn from drug development? The feasibility of “safety by design”

    DEFF Research Database (Denmark)

    Hjorth, Rune; van Hove, Lilian; Wickson, Fern

    2017-01-01

    Safety by design” (SbD) is an intuitively appealing concept that is on the rise within nanotoxicology and nanosafety research, as well as within nanotechnology research policy. It leans on principles established within drug discovery and development (DDD) and seeks to address safety early, as we...

  6. Quality and safety standards in the food industry, developments and challenges

    NARCIS (Netherlands)

    Trienekens, J.H.; Zuurbier, P.J.P.

    2008-01-01

    Consumer concerns related to food safety scandals and globalization of food production have resulted in a global and interconnected system for the production and distribution of food. In the last decade many public and private standards on food safety and quality have been developed as a result of

  7. Development of Strategic Goals of Road Safety Management: A Case Study of Montenegro

    Directory of Open Access Journals (Sweden)

    Milenko Čabarkapa

    2016-12-01

    Full Text Available Evaluation of the results of road safety management at the national level is carried out with a number of predefined indicators. These, predefined indicators should be measurable objectives of road safety management. They are set by national road safety strategies. This paper presents the control implementation of the Strategy to improve road safety in Montenegro for the period 2010-2019. The research showed that the five-year objectives of the National Strategy were achieved in the first years and significantly surmounted. This efficiency is achieved for two main reasons: the development of road safety management, and setting an unambitious, easily attainable goal. These findings are indicators that generally and globally set goals of reducing traffic fatalities cannot comprise at the same time national objectives in all countries. In this context, the methodological improvements of setting national strategic objectives established by the evidences on the national traffic safety issue are proposed.

  8. The effect of innovative activity in firm performance and development: Analysing data from eurozone

    Directory of Open Access Journals (Sweden)

    Ilias A. Makris

    2016-06-01

    Full Text Available Purpose – The purpose of this paper is to examine the effect of Innovative Activity on firm performance and growth. Active Research and Development is considered to be directly related with development, prosperity and growth, in micro and macro level and a key factor in hindering economic recession. Design/methodology/approach – We analyse economic data from listed firms of selected eurozone country-members in order to associate Research and Development with performance indicators in firm and country level. For that purpose, several firm data were collected from WorldScope data base and macroeconomic data from Worldbank database. The period examined is between 2002 and 2012, with a special focus on current financial crisis (after 2007. The empirical process includes, descriptive statistics and logistic regression analysis. Findings – Findings indicate the crucial effect of innovative process in economic performance and development in firm and country level. The latter highlights the urgent need for public support in order to spur innovative activity and high-tech exports, especially in countries that were heavily affected by recession. Research limitations/implications – Some research limitations are the large number of missing cases in WordScope database, as many firms after the beginning of current crisis exit stock market. Furthermore, the other part of the economy, the Small and Medium Enterprises does not exist in the analysis, as listed firms are mainly large and mature companies. Originality/value – The results tend to highlight the need for common policy measures in eurozone, in regard to such issues, instead of imposing horizontal budgetary constraints in specific countries (like Southern Europe, hindering the vicious recessionary circle.

  9. Development of the methodology and approaches to validate safety and accident management

    International Nuclear Information System (INIS)

    Asmolov, V.G.

    1997-01-01

    The article compares the development of the methodology and approaches to validate the nuclear power plant safety and accident management in Russia and advanced industrial countries. It demonstrates that the development of methods of safety validation is dialectically related to the accumulation of the knowledge base on processes and events during NPP normal operation, transients and emergencies, including severe accidents. The article describes the Russian severe accident research program (1987-1996), the implementation of which allowed Russia to reach the world level of the safety validation efforts, presents future high-priority study areas. Problems related to possible approaches to the methodological accident management development are discussed. (orig.)

  10. A study to develop the domestic functional requirements of the specific safety systems of CANDU

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Man Woong [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of); Lee, Jae Young; Park, Kun Chul [Handong Global Univ., Pohang (Korea, Republic of)] (and others)

    2003-03-15

    The present research has been made to develop and review critically the functional requirements of the specific safety systems of CANDU such as SDS-1, SDS2, ECCS, and containment. Based on R documents for this, a systematic study was made to develop the domestic regulation statements. Also, the conventional laws are carefully reviewed to see the compatibility to CANDU. Also, the safety assessment method for CANDU was studied by reviewing C documents and recommendation of IAEA. Through the present works, the vague policy in the CANDU safety regulation is cleaning up in a systematic form and a new frame to measure the objective risk of nuclear power plants was developed.

  11. Analysis and development of the automated emergency algorithm to control primary to secondary LOCA for SUNPP safety upgrading

    International Nuclear Information System (INIS)

    Kim, V.; Kuznetsov, V.; Balakan, G.; Gromov, G.; Krushynsky, A.; Sholomitsky, S.; Lola, I.

    2007-01-01

    The paper presents the results of the study conducted to support planned modernization of the South Ukraine nuclear power plant. The objective of the analysis has been to develop the automated emergency control algorithm for primary to secondary LOCA accident for SUNPP WWER-1000 safety upgrading. According to the analyses performed in the framework of safety assesment report, given accident is the most complex for control and has the largest contribution into the core damage frequency value. This is because of initial event diagnostics is difficult, emergency control is complicated for personnel, time available for decision making and actions performing is limited with coolant inventory for make-up, probability of steam dump valves on affected steam generator non-closing after opening is high, and as a consequence containment bypass, irretrievable loss of coolant and radioactive materials release into the environment are possible. Unit design modifications are directed on expansion of safety systems capabilities to overcome given accident and to facilitate the personnel actions on emergency control. Safety systems modification according to developed algorithm will allow to simplify accident control by personnel and enable to control the ECCS discharge limiting pressure below the affected steam generator steam dump valve opening pressure, and decrease the probability of the containment bypass sequences. The analysis of the primary-to-secondary LOCA thermal-hydraulics has been conducted with RELAP5/Mod 3.2, and involved development of the dedicated analytical model, calculations of various plant response accident scenarios, conducting of plant personnel intervention analyses using full-scale simulator, development and justification of the emergency control algorithm aimed on the minimization of negative consequences of the primary-to-secondary LOCA (Authors)

  12. Radiation Exposure Analyses Supporting the Development of Solar Particle Event Shielding Technologies

    Science.gov (United States)

    Walker, Steven A.; Clowdsley, Martha S.; Abston, H. Lee; Simon, Hatthew A.; Gallegos, Adam M.

    2013-01-01

    NASA has plans for long duration missions beyond low Earth orbit (LEO). Outside of LEO, large solar particle events (SPEs), which occur sporadically, can deliver a very large dose in a short amount of time. The relatively low proton energies make SPE shielding practical, and the possibility of the occurrence of a large event drives the need for SPE shielding for all deep space missions. The Advanced Exploration Systems (AES) RadWorks Storm Shelter Team was charged with developing minimal mass SPE storm shelter concepts for missions beyond LEO. The concepts developed included "wearable" shields, shelters that could be deployed at the onset of an event, and augmentations to the crew quarters. The radiation transport codes, human body models, and vehicle geometry tools contained in the On-Line Tool for the Assessment of Radiation In Space (OLTARIS) were used to evaluate the protection provided by each concept within a realistic space habitat and provide the concept designers with shield thickness requirements. Several different SPE models were utilized to examine the dependence of the shield requirements on the event spectrum. This paper describes the radiation analysis methods and the results of these analyses for several of the shielding concepts.

  13. Development of NUPREP PC Version and Input Structures for NUCIRC Single Channel Analyses

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Churl; Jun, Ji Su; Park, Joo Hwan

    2007-12-15

    The input file for a steady-state thermal-hydraulic code NUCIRC consists of common channel input data and specific channel input data in a case of single channel analysis. Even when all the data is ready for the 380 channels' single channel analyses, it takes long time and requires enormous effort to compose an input file by hand-editing. The automatic pre-processor for this tedious job is a NUPREP code. In this study, a NUPREP PC version has been developed from the source list in the program manual of NUCIRC-MOD2.000 that is imported in a form of an execution file. In this procedure, some errors found in PC executions and lost statements are fixed accordingly. It is confirmed that the developed NUPREP code produces input file correctly for the CANDU-6 single channel analysis. Additionally, the NUCIRC input structure and data format are summarized for a single channel analysis and the input CARDs required for the creep information of aged channels are listed.

  14. Development of NUPREP PC Version and Input Structures for NUCIRC Single Channel Analyses

    International Nuclear Information System (INIS)

    Yoon, Churl; Jun, Ji Su; Park, Joo Hwan

    2007-12-01

    The input file for a steady-state thermal-hydraulic code NUCIRC consists of common channel input data and specific channel input data in a case of single channel analysis. Even when all the data is ready for the 380 channels' single channel analyses, it takes long time and requires enormous effort to compose an input file by hand-editing. The automatic pre-processor for this tedious job is a NUPREP code. In this study, a NUPREP PC version has been developed from the source list in the program manual of NUCIRC-MOD2.000 that is imported in a form of an execution file. In this procedure, some errors found in PC executions and lost statements are fixed accordingly. It is confirmed that the developed NUPREP code produces input file correctly for the CANDU-6 single channel analysis. Additionally, the NUCIRC input structure and data format are summarized for a single channel analysis and the input CARDs required for the creep information of aged channels are listed

  15. Analyses of Injection-Coupled Combustion Instability from J-2X Gas Generator Development

    Science.gov (United States)

    Hulka, James R.; Kenny, R. Jeremy; Protz, Chris; Casiano, Matthew

    2011-01-01

    During development of the gas generator for the liquid oxygen/liquid hydrogen propellant J-2X rocket engine, combustion instabilities were observed near the frequency of the first longitudinal acoustic mode of the hot gas combustion chamber duct. These instabilities were similar to intermediate-frequency or buzz-type instabilities as described in historical programs, except for several aspects: 1) the frequencies were low, in the realm of chug; 2) at times the instability oscillation amplitudes were quite large, with peak-to-peak amplitudes exceeding 50% of the mean chamber pressure along with the appearance of harmonics; 3) the chamber excitation was related to but not exactly at the first longitudinal combustion chamber acoustic mode; and 4) the injector provided mass flow rate oscillations induced by capacitance and inertance effects in the injector rather than by organ pipe resonances of the coaxial oxidizer posts. This type of combustion instability is referred to as "injection coupling" because one critical driving source of the instability is mass flow rate oscillations from the injector. However, the type of injection coupling observed here is different than observed in previous instances of buzz instability with coaxial injectors, because of the lower frequencies and lack of influence from the oxidizer post organ pipe resonances. Test data and preliminary analyses of the initial combustion instabilities were presented in several papers at the 5th Liquid Propulsion Subcommittee meeting. Since that time, additional hot-fire tests with several new hardware configurations have been conducted, and additional analyses have been completed. The analytical models described in previous papers have been updated to include the influences of new geometrical configurations, including a different oxidizer injector manifold configuration and a branch pipe in the hot gas duct that supplies gaseous helium during the start transient to pre-spin the turbine. In addition, the

  16. Monitoring road safety development at regional level: A case study in the ASEAN region.

    Science.gov (United States)

    Chen, Faan; Wang, Jianjun; Wu, Jiaorong; Chen, Xiaohong; Zegras, P Christopher

    2017-09-01

    Persistent monitoring of progress, evaluating the results of interventions and recalibrating to achieve continuous improvement over time is widely recognized as being crucial towards the successful development of road safety. In the ASEAN (Association of Southeast Asian Nations) region there is a lack of well-resourced teams that contain multidisciplinary safety professionals, and specialists in individual countries, who are able to carry out this work effectively. In this context, not only must the monitoring framework be effective, it must also be easy to use and adapt. This paper provides a case study that can be easily reproduced; based on an updated and refined Road Safety Development Index (RSDI), by means of the RSR (Rank-sum ratio)-based model, for monitoring/reporting road safety development at regional level. The case study was focused on the road safety achievements in eleven Southeast Asian countries; identifying the areas of poor performance, potential problems and delays. These countries are finally grouped into several classes based on an overview of their progress and achievements regarding to road safety. The results allow the policymakers to better understand their own road safety progress toward their desired impact; more importantly, these results enable necessary interventions to be made in a quick and timely manner. Keeping action plans on schedule if things are not progressing as desired. This would avoid 'reinventing the wheel' and trial and error approaches to road safety, making the implementation of action plans more effective. Copyright © 2017 Elsevier Ltd. All rights reserved.

  17. Tofacitinib, an oral Janus kinase inhibitor, in patients from Mexico with rheumatoid arthritis: Pooled efficacy and safety analyses from Phase 3 and LTE studies.

    Science.gov (United States)

    Burgos-Vargas, Ruben; Cardiel, Mario; Xibillé, Daniel; Pacheco-Tena, César; Pascual-Ramos, Virginia; Abud-Mendoza, Carlos; Mahgoub, Ehab; Rahman, Mahboob; Fan, Haiyun; Rojo, Ricardo; García, Erika; Santana, Karina

    2017-05-25

    Tofacitinib is an oral Janus kinase inhibitor for the treatment of rheumatoid arthritis (RA). We characterized efficacy and safety of tofacitinib in Mexican patients from RA Phase 3 and long-term extension (LTE) studies. Data from Mexican patients with RA and an inadequate response to disease-modifying antirheumatic drugs (DMARDs) were taken from four Phase 3 studies (pooled across studies) and one open-label LTE study of tofacitinib. Patients received tofacitinib 5 or 10mg twice daily, adalimumab (one Phase 3 study) or placebo (four Phase 3 studies) as monotherapy or in combination with conventional synthetic DMARDs. Efficacy up to Month 12 (Phase 3) and Month 36 (LTE) was assessed by American College of Rheumatology 20/50/70 response rates, Disease Activity Score (erythrocyte sedimentation rate), and Health Assessment Questionnaire-Disability Index. Safety, including incidence rates (IRs; patients with events/100 patient-years) for adverse events (AEs) of special interest, was assessed throughout the studies. 119 and 212 Mexican patients were included in the Phase 3 and LTE analyses, respectively. Tofacitinib-treated patients in Phase 3 had numerically greater improvements in efficacy responses versus placebo at Month 3. Efficacy was sustained in Phase 3 and LTE studies. IRs for AEs of special interest were similar to those with tofacitinib in the global and Latin American RA populations. In Mexican patients from the tofacitinib global RA program, tofacitinib efficacy was demonstrated up to Month 12 in Phase 3 studies and Month 36 in the LTE study, with a safety profile consistent with tofacitinib global population. Copyright © 2017 Elsevier España, S.L.U. and Sociedad Española de Reumatología y Colegio Mexicano de Reumatología. All rights reserved.

  18. Development of Safety Grade PLC (POSAFE-Q) and Performance Test Results

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Chang Hwoi; Park, Won Man; Choi, Jong Gyun; Lee, Dong Young [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); No, Young Hun; Song, Seung Hwan [POSCON, Seoul (Korea, Republic of)

    2006-07-01

    The safety grade PLC (POSAFE-Q) is being developed in the Korea Nuclear Instrumentation and Control System (KNICS) R and D project. The PLC satisfies Safety Class 1E, Quality Class 1, and Seismic Category I. The software such as the RTOS and firmware are being developed according to the safety critical software life cycle. Especially, the formal method is applied to design the SRS (Software Requirement Spec.) and the SDS (Software Design Specification.) to be error-free. The POSAFE-Q has several modules such as processor module, input and output modules, communication modules, redundant processor module, redundant power modules, etc,. To verify the function and performance, several tests such as CT, IT and ST were performed. And also, the equipment qualification test for environment, EMI and EMC, and seismic ware performed. All tests are satisfied with the requirements and specification for safety grade PLC, and the criteria for safety system in nuclear power plants.

  19. Key Factors Affecting Construction Safety Performance in Developing Countries: Evidence from Cambodia

    Directory of Open Access Journals (Sweden)

    Serdar Durdyev

    2017-12-01

    Full Text Available Although proper safety management in construction is of utmost importance; anecdotal evidence suggests that safety is not adequately considered in many developing countries. This paper considers the key variables affecting construction safety performance in Cambodia. Using an empirical questionnaire survey targeting local construction professionals, respondents were invited to rate the level of importance of 30 variables identified from the seminal literature. The data set was subjected to factor analysis. Correlations between the variables show that five key factors underlie the challenges facing the local industry; management and organisation, resources, site management, cosmetic and workforce. It is found that the forefront construction professionals (top management and government authorities should take more responsibilities for further improvements in safety performance on project sites. Findings and recommendations of this study may be useful to construction professional who are seeking ways to improve safety records in developing countries.

  20. Developing an OMERACT Core Outcome Set for Assessing Safety Components in Rheumatology Trials

    DEFF Research Database (Denmark)

    Klokker, Louise; Tugwell, Peter; Furst, Daniel E

    2016-01-01

    in such COS. The Outcome Measures in Rheumatology (OMERACT) Filter 2.0 emphasizes the importance of measuring harms. The Safety Working Group was reestablished at the OMERACT 2016 with the objective to develop a COS for assessing safety components in trials across rheumatologic conditions. METHODS: The safety......OBJECTIVE: Failure to report harmful outcomes in clinical research can introduce bias favoring a potentially harmful intervention. While core outcome sets (COS) are available for benefits in randomized controlled trials in many rheumatic conditions, less attention has been paid to safety...... components of safety that should be measured in such trials by use of a patient-driven approach including qualitative data collection and statistical organization of data, and (3) developing a COS through consensus processes including everyone involved. RESULTS: Members of OMERACT including patients...

  1. Observational analyses of dramatic developments of a severe air pollution event in the Beijing area

    Science.gov (United States)

    Li, Ju; Sun, Jielun; Zhou, Mingyu; Cheng, Zhigang; Li, Qingchun; Cao, Xiaoyan; Zhang, Jingjiang

    2018-03-01

    A rapid development of a severe air pollution event in Beijing, China, at the end of November 2015 was investigated with unprecedented observations collected during the field campaign of the Study of Urban Rainfall and Fog/Haze (SURF-15). Different from previous statistical analyses of air pollution events and their correlations with meteorological environmental conditions in the area, the role of turbulent mixing in the pollutant transfer was investigated in detail. The analyses indicate that the major pollution source associated with high particulate matter of diameter 2.5 µm (PM2.5) was from south of Beijing. Before the day of the dramatic PM2.5 increase, the nighttime downslope flow from the mountains to the west and north of Beijing reduced the surface PM2.5 concentration northwest of Beijing. The nighttime surface stable boundary layer (SBL) not only kept the relatively less-polluted air near the surface, it also shielded the rough surface from the pollutant transfer by southwesterly winds above the SBL, leading to the fast transport of pollutants over the Beijing area at night. As the daytime convective turbulent mixing developed in the morning, turbulent mixing transported the elevated polluted air downward even though the weak surface wind was from northeast, leading to the dramatic increase of the surface PM2.5 concentration in the urban area. As a result of both turbulent mixing and advection processes with possible aerosol growth from secondary aerosol formation under the low-wind and high-humidity conditions, the PM2.5 concentration reached over 700 µg m-3 in the Beijing area by the end of the day. Contributions of the two transporting processes to the PM2.5 oscillations prior to this dramatic event were also analyzed. The study demonstrates the important role of large-eddy convective turbulent mixing in vertical transfer of pollutants and the role of the SBL in not only decoupling vertical transport of trace gases and aerosols but also in

  2. Road safety in developing countries: The role of research.

    NARCIS (Netherlands)

    Schreuder, D.A.

    1986-01-01

    Road accidents are a continuous burden for all countries, developing and developed alike. There are reasons, however, to pay special attention to developing countries as the situation often seems to be more unfavourable as in developed countries, and as the consequences of road accidents, can be

  3. Developing a safety report for an existing conversion facility

    International Nuclear Information System (INIS)

    Carisse, Hess

    2013-01-01

    A review of the process used to meet the regulatory requirements for a Safety Report at an existing conversion facility is described. This paper will cover the establishment of the regulatory criteria, selection of appropriate methodologies, identification of events and modeling of credible events. Once established there is on-going maintenance to deal with design changes and the need for periodic reviews will also be discussed. Challenges in dealing with the various phases, including incorporation of historical licensing documents, and lessons learned are presented. Of specific interest is the failure of the selected methodology to deal with infrastructure issues. One aspect of lessons learned that will be explored is the lack of an available mechanism for sharing information with similar fuel cycle facilities which is compounded by the fact that there are a small number of fuel cycle facilities compared to nuclear power plants. Possible approaches to dealing with this issue are also discussed. (authors)

  4. Status of nuclear safety developments and prospects for the future

    International Nuclear Information System (INIS)

    Birkhofer, A.; Laverie, M.

    1992-01-01

    During the eighties, the emphasis on safety questions progressively shifted from design and construction of plants towards the operation of existing plants. Lessons learned from incidents and accidents in national or foreign plants, experience gained from the operation and maintenance of plants, research and ageing of installations were some of the issues on which, at that time, the nuclear community focused its attention and concern. At the same time, numerous actions, dedicated to the strengthening of defense-in-depth which consists of piling up enough dispositions (or defense lines) to be able to deal with situations even beyond the initial design of plants, were implemented. Despite all these technical precautions to reduce the risk to the public and the environment, the acceptance level decreased in some countries, due to the impact of T.M.T. and Chernobyl, a certain misunderstanding between technicians and the public proceeding from a lack of openness and poor communication

  5. Research strategies for safety evaluation of nanomaterials, part VIII: International efforts to develop risk-based safety evaluations for nanomaterials.

    Science.gov (United States)

    Thomas, Karluss; Aguar, Pilar; Kawasaki, Hajime; Morris, Jeff; Nakanishi, Junko; Savage, Nora

    2006-07-01

    The use of nanotechnology in consumer and industrial applications will likely have a profound impact on a number of products from a variety of industrial sectors. Nanomaterials exhibit unique physical/chemical properties and impart enhancements to engineered materials, including better magnetic properties, improved electrical activity, and increased optical properties. The United States, Europe, and Japan have each initiated comprehensive programs to promote and expand the utility of nanotechnology for commercial applications. An important component of these programs is the development of reliable risk and safety evaluations for these materials to ensure their safety for human health and the environment. The scope of each of these programs includes efforts to assess the hazards posed by nanomaterials in realistic exposure conditions.

  6. Workplace activities to promote small attempts for safety. Toward development of safety culture in a nuclear power plant

    International Nuclear Information System (INIS)

    Fukui, Hirokazu; Sugiman, Toshio

    2007-01-01

    Activities that could possibly grow into learning activities for developing safety culture were explored by intensive fieldwork in a nuclear power plant depending on Engestroem's activity theory. As a first step to achieve this goal, workers' small attempts that might contribute to nurturing a safety culture were investigated. Eight kinds of activity were observed and interpreted as having the possibility to facilitate small recognition and small practice, i.e., activities including (1) workgroup as community, (2) other workgroups and other departments as community, (3) meeting drawing remarks as mediating artifacts, (4) study session and Off-the-Job-Training as mediating artifact, (5) award as mediating artifact, (6) extended leave as mediating artifact, (7) check sheet as mediating artifact, and (8) skill-transfer system as mediating artifact. (author)

  7. Developing safety culture in nuclear activities. Practical suggestions to assist progress

    International Nuclear Information System (INIS)

    1998-01-01

    The term 'safety culture' was introduced by the International Nuclear Safety Advisory Group (INSAG) in Summary Report on the Post-Accident Review Meeting on the Chernobyl Accident published by IAEA as safety Series No. 75-INSAG-1 in 1986, and expanded in Basic Safety principles for Nuclear Power Plants, Safety Series No. 75-INSAG-3 in 1988. This publication supplements INSAG-4 published in 1991 which includes the definition and concept of safety culture describing practices valuable in establishing and maintaining a sound safety culture in a number of countries. It is intended for those who design, construct, manufacture, operate, maintain or decommission nuclear facilities. It should be practically useful for all those involved in operating nuclear facilities. It will also provide a reference for groups such as regulators who have an interest in developing, improving and evaluating safety culture training and individuals engaged in nuclear activities, and for bodies such as ethics review committees who should take into account safety culture issues for certifying professional excellence in the medical field

  8. Development and evaluation of a pedestrian safety training program for elementary school bus riders

    Science.gov (United States)

    1994-12-01

    The objective of this study was to develop and evaluate a comprehensive pedestrian safety program for elementary (kindergarten through grade 6) school bus riders. Existing materials, crash data and state laws/regulations on school bus pedestrian safe...

  9. Development of Comprehensive Nuclear Safety Regulation Plan for 2007-2011

    International Nuclear Information System (INIS)

    Choi, Young Sung; Kim, Woong Sik; Park, Dong Keuk; Kim, Ho Ki

    2006-01-01

    The Article 8-2 of Atomic Energy Act requires the government to establish Atomic Energy Promotion Plan every five years. It sets out national nuclear energy policies in a systematic and consistent way. The plan presents the goals and basic directions of national nuclear energy policies on the basis of current status and prospects. Both areas of utilization and safety management of nuclear energy are included and various projects and schedules are delineated based on the national policy directions. The safety management area in this plan deals with the overall safety and regulation policy. Its detail projects and schedule should be developed in separate plans by responsible ministries under the mediation of the MOST. As a regulatory authority, MOST is responsible for safety management area and its technical support organization, KINS has developed Comprehensive Nuclear Safety Regulation Plan as an implementation plan of safety area. This paper presents the development process and specific projects contained in the Comprehensive Nuclear Safety Regulation Plan which is under development now

  10. [The origin and development of the application of ergonomics in patient safety].

    Science.gov (United States)

    Zhang, Longhao; Mao, Xuanyue; Jia, Pengli; Zhang, Mingming

    2015-07-01

    Human factor and ergonomics (HFE) is a discipline focusing on the studies of the relation between the safety and human behavior and system design. The application of HFE in healthcare and patient safety can be traced back to the 1950s. To date, the method and theory of HFE, including its principles and methodology, has played an important part in patient safety activities, and has gained more and more attention by the relevant health and hygienic institutions in developed countries, while its application in developing countries like China is still at the very initial stage with only a few research articles published, most of them theoretical rather than practical.

  11. Development of Necessary Technology for localizing of Nuclear Safety Grade I and C System

    International Nuclear Information System (INIS)

    Kwon, Kee Choon; Jang, Tong Il; Kim, Jung Tack

    2010-08-01

    Through KNICS and Nu-Tech 2012, a MMIS(Man-Machine Interface System) package for nuclear plants was localized and is expected to be applied to SUN 1,2 plants. This study is aimed to support the application of the MMIS package including the following technologies that meet strengthened regulation requirements and enhanced utility's performance requirements to nuclear plants. - Dedication to safety-grade computers and real time operating systems - The broadband communication network for safety information - Application of the automatic test for safety systems - Application of programmable logic controllers to nuclear plants - Development strategy for an integrated SW development tool for control device

  12. Development, expansion, and use of a stroke clinical trials resource for novel exploratory analyses.

    Science.gov (United States)

    Ali, Myzoon; Bath, Philip; Brady, Marian; Davis, Stephen; Diener, Hans-Christoph; Donnan, Geoffrey; Fisher, Marc; Hacke, Werner; Hanley, Daniel F; Luby, Marie; Tsivgoulis, G; Wahlgren, Nils; Warach, Steven; Lees, Kennedy R

    2012-02-01

    Analysis of reliable registry data can direct future research to influence clinical care. Data from the Virtual International Stroke Trials Archive have been used to test hypotheses and inform trial design. We sought to expand Virtual International Stroke Trials Archive into a broader stroke resource with new opportunities for research and international collaboration. Using procedures initially developed for an acute stroke trial archive, we invited trialists to lodge data on rehabilitation, secondary prevention, intracerebral haemorrhage, imaging, and observational stroke studies. We have extended Virtual International Stroke Trials Archive into six subsections: Virtual International Stroke Trials Archive-Acute (n = 28 190 patients' data), Virtual International Stroke Trials Archive-Rehab (n = 10 194), Virtual International Stroke Trials Archive-intracerebral haemorrhage (n = 1829), Virtual International Stroke Trials Archive-Prevention, Virtual International Stroke Trials Archive-Imaging (n = 1300), and Virtual International Stroke Trials Archive-Plus (n = 6573). Enrollment continues, with commitments for the contribution of six further trials to Virtual International Stroke Trials Archive-Prevention, 13 trials to Virtual International Stroke Trials Archive-Rehab, and one registry to Virtual International Stroke Trials Archive-Plus. Data on age, type of stroke, medical history, outcomes by modified Rankin scale and Barthel Index (BI), mortality, and adverse events are available for analyses. The Virtual International Stroke Trials Archive network encourages the development of young investigators and provides opportunities for international peer review and collaboration. Application of the original Virtual International Stroke Trials Archive concepts beyond acute stroke trials can extend the value of clinical research at low cost, without threatening commercial or intellectual property interests. This delivers valuable research output to inform

  13. Development of an algorithm for analysing the electronic measurement of medication adherence in routine HIV care.

    Science.gov (United States)

    Rotzinger, Aurélie; Cavassini, Matthias; Bugnon, Olivier; Schneider, Marie Paule

    2016-10-01

    Background Medication adherence is crucial for successful treatment. Various methods exist for measuring adherence, including electronic drug monitoring, pharmacy refills, pill count, and interviews. These methods are not equivalent, and no method can be considered as the gold standard. A combination of methods is therefore recommended. Objective To develop an algorithm for the management of routinely collected adherence data and to compare persistence and implementation curves using post-algorithm data (reconciled data) versus raw electronic drug monitoring data. Setting A community pharmacy located within a university medical outpatient clinic in Lausanne, Switzerland. Methods The algorithm was developed to take advantage of the strengths of each available adherence measurement method, with electronic drug monitoring as a cornerstone to capture the dynamics of patient behaviour, pill count as a complementary objective method to detect any discrepancy between the number of openings measured by electronic monitoring and the number of pills ingested per opening, and annotated interviews to interpret the discrepancy. The algorithm was tested using data from patients taking lopinavir/r and having participated in an adherence-enhancing programme for more than 3 months. Main outcome measure Adherence was calculated as the percentage of persistent patients (persistence) and the proportion of days with correct dosing over time (implementation) from inclusion to the end of the median follow-up period. Results A 10-step algorithm was established. Among 2041 analysed inter-visit periods, 496 (24 %) were classified as inaccurate, among which 372 (75 %) could be reconciled. The average implementation values were 85 % (raw data) and 91 % (reconciled data) (p electronic drug monitoring, pill count and patient interviews is possible within the setting of a medication adherence clinic. Electronic drug monitoring underestimates medication adherence, affecting subsequent

  14. 48{sup th} Annual meeting on nuclear technology (AMNT 2017). Key topic / Enhanced safety and operation excellence. Focus session: Uncertainty analyses in reactor core simulations

    Energy Technology Data Exchange (ETDEWEB)

    Zwermann, Winfried [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Garching (Germany). Forschungszentrum

    2017-12-15

    The supplementation of reactor simulations by uncertainty analyses is becoming increasingly important internationally due to the fact that the reliability of simulation calculations can be significantly increased by the quantification of uncertainties in comparison to the use of so-called conservative methods (BEPU- ''Best-Estimate plus Uncertainties''). While systematic uncertainty analyses for thermo-hydraulic calculations have been performed routinely for a long time, methods for taking into account uncertainties in nuclear data, which are the basis for neutron transport calculations, are under development. The Focus Session Uncertainty Analyses in Reactor Core Simulations was intended to provide an overview of international research and development with respect to supplementing reactor core simulations with uncertainty and sensitivity analyses, in research institutes as well as within the nuclear industry. The presented analyses not only focused on light water reactors, but also on advanced reactor systems. Particular emphasis was put on international benchmarks in the field. The session was chaired by Winfried Zwermann (Gesellschaft fuer Anlagen- und Reaktorsicherheit).

  15. Fire safety of ETICS with wood fibreboards for multi-storey buildings – first research and development results

    Directory of Open Access Journals (Sweden)

    Küppers Judith

    2016-01-01

    Full Text Available Wood fibreboards can serve as alternative sustainable insulation material for external thermal insulation component systems (ETICS. In Germany, the application of ETICS with wood fibreboards is restricted to low buildings. The restrictions are mainly caused by the smouldering tendency of the wood fibreboards. Thus, the development of an ETICS with wood fibreboards for multi-storey buildings complying with the requirements would provide a new scope of application for this sustainable insulation material. This paper presents first results of standardised and innovative investigation methods as basis for the development. These investigations and analyses concern the fire behaviour, especially smouldering processes, the plaster system, other constructive protection measures as well as risk and safety analyses.

  16. Challenges in developing competency-based training curriculum for food safety regulators in India

    Directory of Open Access Journals (Sweden)

    Anitha Thippaiah

    2014-01-01

    Full Text Available Context: The Food Safety and Standards Act have redefined the roles and responsibilities of food regulatory workforce and calls for highly skilled human resources as it involves complex management procedures. Aims: 1 Identify the competencies needed among the food regulatory workforce in India. 2 Develop a competency-based training curriculum for food safety regulators in the country. 3 Develop training materials for use to train the food regulatory workforce. Settings and Design: The Indian Institute of Public Health, Hyderabad, led the development of training curriculum on food safety with technical assistance from the Royal Society for Public Health, UK and the National Institute of Nutrition, India. The exercise was to facilitate the implementation of new Act by undertaking capacity building through a comprehensive training program. Materials and Methods: A competency-based training needs assessment was conducted before undertaking the development of the training materials. Results: The training program for Food Safety Officers was designed to comprise of five modules to include: Food science and technology, Food safety management systems, Food safety legislation, Enforcement of food safety regulations, and Administrative functions. Each module has a facilitator guide for the tutor and a handbook for the participant. Essentials of Food Hygiene-I (Basic level, II and III (Retail/ Catering/ Manufacturing were primarily designed for training of food handlers and are part of essential reading for food safety regulators. Conclusion: The Food Safety and Standards Act calls for highly skilled human resources as it involves complex management procedures. Despite having developed a comprehensive competency-based training curriculum by joint efforts by the local, national, and international agencies, implementation remains a challenge in resource-limited setting.

  17. Challenges in Developing Competency-based Training Curriculum for Food Safety Regulators in India

    Science.gov (United States)

    Thippaiah, Anitha; Allagh, Komal Preet; Murthy, G. V.

    2014-01-01

    Context: The Food Safety and Standards Act have redefined the roles and responsibilities of food regulatory workforce and calls for highly skilled human resources as it involves complex management procedures. Aims: 1) Identify the competencies needed among the food regulatory workforce in India. 2) Develop a competency-based training curriculum for food safety regulators in the country. 3) Develop training materials for use to train the food regulatory workforce. Settings and Design: The Indian Institute of Public Health, Hyderabad, led the development of training curriculum on food safety with technical assistance from the Royal Society for Public Health, UK and the National Institute of Nutrition, India. The exercise was to facilitate the implementation of new Act by undertaking capacity building through a comprehensive training program. Materials and Methods: A competency-based training needs assessment was conducted before undertaking the development of the training materials. Results: The training program for Food Safety Officers was designed to comprise of five modules to include: Food science and technology, Food safety management systems, Food safety legislation, Enforcement of food safety regulations, and Administrative functions. Each module has a facilitator guide for the tutor and a handbook for the participant. Essentials of Food Hygiene-I (Basic level), II and III (Retail/ Catering/ Manufacturing) were primarily designed for training of food handlers and are part of essential reading for food safety regulators. Conclusion: The Food Safety and Standards Act calls for highly skilled human resources as it involves complex management procedures. Despite having developed a comprehensive competency-based training curriculum by joint efforts by the local, national, and international agencies, implementation remains a challenge in resource-limited setting. PMID:25136155

  18. Research on the improvement of nuclear safety -The development of LOCA analysis codes for nuclear power plant-

    International Nuclear Information System (INIS)

    Jang, Won Pyo; Jung, Yung Jong; Kim, Kyung Doo; Jung, Jae Joon; Kim, Won Suk; Han, Doh Heui; Hah, Kooi Suk; Jung, Bub Dong; Lee, Yung Jin; Hwang, Tae Suk; Lee, Sang Yong; Park, Chan Uk; Choi, Han Rim; Lee, Sang Jong; Choi, Jong Hoh; Ban, Chang Hwan; Bae, Kyoo Hwan

    1995-07-01

    The present research aims at development of both a best estimate methodology on LOCA analysis and, as an application, performance analyses of safety systems. SBLOCA analyses have been continued to examine the capacity reduction effect of ECCS since the second project year. As a results, core uncovery, which is requirement of URD has not been occurred in 6'' cold leg break. Although core uncovery has been predicted when DVI line has been broken for DVI+4-Train HPIS, the calculated PCT has lied well within the criterion. The effect of safety injection position and SIT characteristics are also analyzed for LBLOCA. The results show that cold leg injection is the most effective way and the adaption of advanced SIT could lead to elimination of LPSI pump from the safety system. On the other hand, the quantified uncertainties obtained from THTF and FLECHT/SEASET which represents blowdown and reflood phenomena, respectively, have been confirmed using IET(LOFT test). The application uncertainty for Kori unit 3 has been analyzed. Finally, application of the best estimate methodology using the uncertainties concerned with the code, the bais, and the application, leads to overall uncertainty of about 200K for Kori unit 3. 244 figs, 22 tabs, 92 refs. (Author)

  19. Research on the improvement of nuclear safety -The development of LOCA analysis codes for nuclear power plant-

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Won Pyo; Jung, Yung Jong; Kim, Kyung Doo; Jung, Jae Joon; Kim, Won Suk; Han, Doh Heui; Hah, Kooi Suk; Jung, Bub Dong; Lee, Yung Jin; Hwang, Tae Suk; Lee, Sang Yong; Park, Chan Uk; Choi, Han Rim; Lee, Sang Jong; Choi, Jong Hoh; Ban, Chang Hwan; Bae, Kyoo Hwan [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    The present research aims at development of both a best estimate methodology on LOCA analysis and, as an application, performance analyses of safety systems. SBLOCA analyses have been continued to examine the capacity reduction effect of ECCS since the second project year. As a results, core uncovery, which is requirement of URD has not been occurred in 6`` cold leg break. Although core uncovery has been predicted when DVI line has been broken for DVI+4-Train HPIS, the calculated PCT has lied well within the criterion. The effect of safety injection position and SIT characteristics are also analyzed for LBLOCA. The results show that cold leg injection is the most effective way and the adaption of advanced SIT could lead to elimination of LPSI pump from the safety system. On the other hand, the quantified uncertainties obtained from THTF and FLECHT/SEASET which represents blowdown and reflood phenomena, respectively, have been confirmed using IET(LOFT test). The application uncertainty for Kori unit 3 has been analyzed. Finally, application of the best estimate methodology using the uncertainties concerned with the code, the bais, and the application, leads to overall uncertainty of about 200K for Kori unit 3. 244 figs, 22 tabs, 92 refs. (Author).

  20. Synergistic effects of the safety factor and shear flows on development of internal transport barriers in reversed shear plasmas

    International Nuclear Information System (INIS)

    Wang, A.K.; Dong, J.Q.; Qu, W.X.; Qiu, X.M.

    2002-01-01

    A new suppression mechanism of turbulent transport, characteristic of the synergism between safety factor and shear flows, is proposed to explain the internal transport barriers (ITBs) observed in neutral-beam-heated tokamak discharges with reversed magnetic shear. It is shown that the evolution of turbulent transport with the strength of the suppression mechanism reproduces the basic features of the formation and development of ITBs observed in experiments. In addition, the present analyses predict the possibility of global ion and electron heat transport barriers

  1. Approach to developing a ground-motion design basis for facilities important to safety at Yucca Mountain

    Energy Technology Data Exchange (ETDEWEB)

    King, J.L.

    1990-04-01

    The Department of Energy has proposed a methodology for developing a ground-motion design basis for prospective facilities at Yucca Mountain that are important to safety. The methodology utilizes a quasi-deterministic construct that is designed to provide a conservative, robust, and reproducible estimate of ground motion that has a one-in-ten chance of occurring during the preclosure period. This estimate is intended to define a ground-motion level for which the seismic design would ensure minimal disruption to operations; engineering analyses to ensure safe performance in the unlikely event that the design basis is exceeded are a part of the proposed methodology. 8 refs.

  2. NS [Nuclear Safety] update. Current safety and security activities and developments taking place in the Department of Nuclear Safety and Security. Issue no. 5, September 2007

    International Nuclear Information System (INIS)

    2007-09-01

    This newsletter reports on the denial of shipments of radioactive materials, the IAEA,s illicit trafficking database, the knowledge management conference and the IAEA Response Assistance network (RANET). Radioactive material needs to be transported for use in public health and industry. Transport of radioactive material is governed by national and international regulations which are based on the IAEA's Regulations for the Safe Transport of Radioactive Material. These Regulations, developed by experts around the world, ensure high standards of safety. However, even when complying with the Regulations, there continue to be instances where shipments have been denied or delayed. Denying or delaying a shipment of radioactive material for medical use can result in hardships to patients. In addition, radioactive materials for other activities such as sterilization or power generation are also affected. In 2006, to increase transparency, seek effective solutions and permit participation of interested parties, the IAEA Director General created a senior level International Steering Committee on Denials of Shipments of Radioactive Material, which includes representatives from IAEA Member States, international governmental and nongovernmental organizations and industry. The Committee's mandate is to coordinate international efforts at determining solution of issues related to the denial of shipments and facilitate the coordination of a comprehensive international work plan of activities. In the early 1990s a number of trafficking cases in the Czech Republic, Germany and the Russian Federation involved highly enriched uranium (HEU) and Plutonium, some in kilogram quantities. These cases raised serious concerns over the threat posed by nuclear materials which were apparently available for sale on the black market. In response, States asked the IAEA to establish a data base on illicit trafficking incidents. First established in the mid-1990s, this work was greatly accelerated and

  3. Technology Development for Integrated Safety Test of Spent Nuclear Fuel Transportation and Storage System

    International Nuclear Information System (INIS)

    Seo, Kiseog; Seo, J. S.; Lee, J. C.

    2012-05-01

    A dedicated review on the U. S. NRC Regulation 10 CFR Part 72 and regulatory guide NUREG/1536 has been performed. The safety requirements for spent nuclear fuel dry storage cask are analyzed and summarized in structural, thermal, shielding, criticality, materials, tests and maintenance aspects. Also a guideline for preparing the safety analysis report is provided. The heat flow analysis was performed by varying the dimensions of the heat flow test facility. From the heat flow analysis for the test facility, as the test facility became test facility. From the heat flow analysis for the test facility, as the test facility became bigger; the thermal effect became smaller. Therefore, the dimensions of the heat flow test facility was designed with 5m Χ 5m Χ 6m(H). Analyses of heat transfer characteristics and mechanism for spent PWR fuel assemblies, option study for production of the effective thermal conductivity and option study for effective thermal conductivity test have been performed to obtain the basic data for production of the effective thermal conductivity. It became clear that the diffusion coefficient of chloride ion of concrete remarkably increases along with the temperature rise, and that there is a linear relation between the logarithm values of the diffusion coefficients and the reciprocal of the temperature. It is understood to be able to express the temperature dependency of the diffusion coefficient roughly by an Arrhenius equation as the velocity coefficient is provided as the diffusion coefficient. The specifications and characteristics of storage facilities under operation including dual purpose casks were investigated. Components subject to material degradation were examined. Based on literature survey, investigating a drop analysis incorporating with material degradation, the basic data to develop an analysis methodology was obtained

  4. Development and implementation of a hospital-based patient safety program

    International Nuclear Information System (INIS)

    Frush, Karen S.; Alton, Michael; Frush, Donald P.

    2006-01-01

    Evidence from numerous studies indicates that large numbers of patients are harmed by medical errors while receiving health-care services in the United States today. The 1999 Institute of Medicine report on medical errors recommended that hospitals and health-care agencies ''establish safety programs to act as a catalyst for the development of a culture of safety'' [1]. In this article, we describe one approach to successful implementation of a hospital-based patient safety program. Although our experience at Duke University Health System will be used as an example, the needs, principles, and solutions can apply to a variety of other health-care practices. Key components include the development of safety teams, provision of tools that teams can use to support an environment of safety, and ongoing program modification to meet patient and staff needs and respond to changing priorities. By moving patient safety to the forefront of all that we do as health-care providers, we can continue to improve our delivery of health care to children and adults alike. This improvement is fostered when we enhance the culture of safety, develop a constant awareness of the possibility of human and system errors in the delivery of care, and establish additional safeguards to intercept medical errors in order to prevent harm to patients. (orig.)

  5. National program for the fostering and development of safety culture in the nuclear activities in Cuba

    International Nuclear Information System (INIS)

    Ferro Fernandez, R.; Guillen Campos, A.

    2002-01-01

    Since its appearance, as a result of the investigations of the accident in the Chernobyl Nuclear Power Plant, the term Safety Culture has been considered a key element to achieve a high level of safety in the nuclear installations, becoming a basic safety principle, internationally. The Cuban regulatory authority understood from very early the importance of the promotion and development of attitudes and characteristics in the organizations and personnel involved in the nuclear sector, reflecting a high Safety Culture, in order to propitiate a higher involvement of all employees in safety, contributing this way to the prevention of accidents in the nuclear facilities. Although the Cuban Nuclear Program was significantly reduced in the 90's, the regulatory authority has continued working in this direction and assimilated all the international experience for its application in its strategies for the development of a Safety Culture in the nuclear activities in the country. The present work summarizes the Cuban experience in the establishment of a National Program for the fostering and development of a Safety Culture. (author)

  6. Development of an auditable safety analysis in support of a radiological facility classification

    International Nuclear Information System (INIS)

    Kinney, M.D.; Young, B.

    1995-01-01

    In recent years, U.S. Department of Energy (DOE) facilities commonly have been classified as reactor, non-reactor nuclear, or nuclear facilities. Safety analysis documentation was prepared for these facilities, with few exceptions, using the requirements in either DOE Order 5481.1B, Safety Analysis and Review System; or DOE Order 5480.23, Nuclear Safety Analysis Reports. Traditionally, this has been accomplished by development of an extensive Safety Analysis Report (SAR), which identifies hazards, assesses risks of facility operation, describes and analyzes adequacy of measures taken to control hazards, and evaluates potential accidents and their associated risks. This process is complicated by analysis of secondary hazards and adequacy of backup (redundant) systems. The traditional SAR process is advantageous for DOE facilities with appreciable hazards or operational risks. SAR preparation for a low-risk facility or process can be cost-prohibitive and quite challenging because conventional safety analysis protocols may not readily be applied to a low-risk facility. The DOE Office of Environmental Restoration and Waste Management recognized this potential disadvantage and issued an EM limited technical standard, No. 5502-94, Hazard Baseline Documentation. This standard can be used for developing documentation for a facility classified as radiological, including preparation of an auditable (defensible) safety analysis. In support of the radiological facility classification process, the Uranium Mill Tailings Remedial Action (UMTRA) Project has developed an auditable safety analysis document based upon the postulation criteria and hazards analysis techniques defined in DOE Order 5480.23

  7. User-Centered Collaborative Design and Development of an Inpatient Safety Dashboard.

    Science.gov (United States)

    Mlaver, Eli; Schnipper, Jeffrey L; Boxer, Robert B; Breuer, Dominic J; Gershanik, Esteban F; Dykes, Patricia C; Massaro, Anthony F; Benneyan, James; Bates, David W; Lehmann, Lisa S

    2017-12-01

    Patient safety remains a key concern in hospital care. This article summarizes the iterative participatory development, features, functions, and preliminary evaluation of a patient safety dashboard for interdisciplinary rounding teams on inpatient medical services. This electronic health record (EHR)-embedded dashboard collects real-time data covering 13 safety domains through web services and applies logic to generate stratified alerts with an interactive check-box function. The technological infrastructure is adaptable to other EHR environments. Surveyed users perceived the tool as highly usable and useful. Integration of the dashboard into clinical care is intended to promote communication about patient safety and facilitate identification and management of safety concerns. Copyright © 2017 The Joint Commission. All rights reserved.

  8. A conceptual framework to development of construction safety culture in Indonesia

    Science.gov (United States)

    Armyn Machfudiyanto, Rossy; Latief, Yusuf

    2017-12-01

    Working accidents in the construction industry are among the highest in the world, affecting the three levels of both macro (National) mezzo (Enterprise) and micro (Projects) that need to be integrated in building a safety culture. The purpose of this research is to develop a conceptual framework in improving safety culture in the construction industry in Indonesia. The methodology was developed using literature study and deductive analysis which then performed expert validation to ensure the concept developed. The result of this research is that policy and institution as input to build safety culture which need to be followed up with increasing of company maturity which have implication to safety performance and construction project performance.

  9. Study on development of education model and its evaluation system for radiation safety

    CERN Document Server

    Seo, K W; Nam, Y M

    2002-01-01

    As one of the detailed action strategy of multi object preparedness for strengthening of radiation safety management by MOST, this project was performed, in order to promote the safety culture for user and radiation worker through effective education program. For the prevention of radiological accident and effective implementation of radiation safety education and training, this project has been carried out the development of education model and its evaluation system on radiation safety. In the development of new education model, education course was classified; new and old radiation worker, temporary worker, lecturer and manager. The education model includes the contents of expanding the education opportunity and workplace training. In the development of evaluation system, the recognition criteria for commission-education institute and inside-education institute which should establish by law were suggested for evaluation program. The recognition criteria contains classification, student, method, facilities, ...

  10. Developing international safety standards for the geological disposal of radioactive waste

    International Nuclear Information System (INIS)

    Metcalf, P.

    2001-01-01

    In the context of the International Atomic Energy Agency's (IAEA) programme to create a corpus of internationally accepted Radioactive Waste Safety Standards (RADWASS), focus is currently being placed on establishing standards for the 'geological disposal of radioactive waste'. This is a challenging task and to help the standards development process there is a need to stimulate discussion of some of the associated scientific and technical issues. A number of position papers developed in recent years by a subgroup of the Waste Safety Standards Committee (WASSC), the subgroup on Principles and Criteria for Radioactive Waste Disposal, address many of the relevant issues. These include a common safety based framework for radioactive waste disposal, appropriate time frames for safety assessment, different possible indicators of long-term safety, the safety implications of reversibility and retrievability, the assessment of possible human intrusion into the repository, the role and limitations of institutional control, establishing reference critical groups and biospheres for long-term assessment, and what is meant by 'compliance' with the standards. These papers will be discussed at a Specialists Meeting to be held at the IAEA in June 2001 as a means of establishing the extent to which they enjoy the general support of experts. In order to broaden that consensus, the conclusions reached at the Specialists Meeting on the issues listed above will be presented and discussed with participants at a number of international meetings. Later this year, a draft safety standard on the geological disposal of radioactive waste which takes account of the consensus positions reached through the various consultations will be submitted for the consideration of Waste Safety Standards Committee (WASSC), the officially approved body within the IAEA for the review and approval of waste safety standards. The Committee is made up of government appointed radioactive waste regulators

  11. Experience of Tecnatom in Developing a Strong Leadership for Safety and Performance

    International Nuclear Information System (INIS)

    González, F.; Villadóniga, J. I.

    2016-01-01

    This paper presents experience and insights of Tecnatom in the support of internal and external clients to develop a strong Leadership for Safety. Several cases are presented briefly: (a) The leadership and culture change activities for a utility, a radwaste company, and for Tecnatom itself. One important characteristic of the work performed is the detailed consideration of the underlying organizational culture that underpins the safety culture. Measurable improvements have been achieved and some of the key insights are shared in this paper. (b) The development and implementation of a leadership model with 17 competencies, including safety explicitly. One benefit of this model is that allows to perform a quantitative assessment of leadership effectiveness, something vital to be able to ensure that leadership development actions are truly supporting safety. The model uses an approach to development oriented to strengths and the use of companion competencies to further develop leadership. Moreover it aims to produce significant improvements on safety but also on performance, since both are not competing goals when the proper leadership model is selected. The training material prepared was shortlisted in the 2014 Nuclear Training Awards. (c) The design and implementation of a training development program on Safety Culture, and required competencies of Leadership, for Top Managers of the nuclear industry, as part of the project NUSHARE of the European Commission’s 7th research framework program. The program is sensible to the reduced time availability of Top Managers and uses a combination of learning approaches (webinars, micro-elearnings, web meetings) that provide higher flexibility for the learner, but complemented with other proven methods (group dialog, journaling, mentoring, etc.) to ensure that the program is effective. All these experiences reveal that to improve the organizational Safety Culture we need to enhance Leadership for Safety and Performance

  12. Development of CHF correlation “MG-NV” for low pressure and low velocity conditions applied to PWR safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Yumura, T.; Yodo, T.; Makino, Y.; Suemura, T. [Mitsubishi Heavy Industries, LTD., Kobe, Hyogo (Japan)

    2011-07-01

    The Critical Heat Flux (CHF) is one of the important parameters in the safety analysis of Pressurized Water Reactor (PWR). If the CHF is reached, an abrupt drop occurs in the heat transfer between the fuel rod cladding and the reactor coolant, which may induce a large temperature excursion of fuel cladding and a subsequent fuel failure. Therefore, accurate prediction of CHF is required in order to assure a sufficient safety margin in the PWR core. Mitsubishi Heavy Industries, ltd (MHI) is developing a new series of CHF correlations which covers various fuel designs and wide range of fluid conditions with sufficient reliability. In this paper, a new CHF correlation, MG-NV (Mitsubishi Generalized correlation for Non-Vane grid spacers) is presented. This correlation is one of the basic components of the new correlation series and was developed to cover low pressure and low velocity conditions where the rod bundle CHF data are limited. The CHF correlation was developed based on open CHF database and provides conservative but more reliable rod bundle CHF predictions compared with the conventional CHF correlations used in safety analyses at low pressure condition, such as Main Steam Line Break event. (author)

  13. NS [Nuclear Safety] update. Current safety and security activities and developments taking place in the Department of Nuclear Safety and Security. Issue no. 2, January 2007

    International Nuclear Information System (INIS)

    2006-08-01

    This newsletter reports on the training of cardiologists in radiation protection, IAEA's safety review services and the operational safety assessment review team (OSART), the international conference on management of spent fuel and the recent INSAG (International Nuclear Safety Group) publications. The IAEA has begun a major international initiative to train interventional cardiologists in radiation protection. Starting with the first course in May 2004, so far 6 regional and 3 national training courses have been conducted with the participation of over 400 health professionals putting the IAEA in a leading role in this area. A programme of two days' training has been developed, covering possible and observed radiation effects among patients and staff, international standards, dose management techniques, examples of good and bad practice and examples indicating prevention of possible injuries as a result of good practice in radiation protection. The training material is freely available on CD and will be placed on the Radiological Protection of Patients website at http://rpop.iaea.org/

  14. Development and validation of a new safety climate scale for petrochemical industries.

    Science.gov (United States)

    Jafari, Mohammad Javad; Eskandari, Davood; Valipour, Firouz; Mehrabi, Yadollah; Charkhand, Hossein; Mirghotbi, Mostafa

    2017-01-01

    While a considerable body of research has studied safety climate and its role as a leading indicator of organizational safety, much of this work has been conducted with Western manufacturing samples. The current study puts emphasis on the cross-validation of a safety climate model in the non-Western industrial context of Iranian petrochemical industries. The current study was performed in one petrochemical company in Iran. The scale was developed through conducting a literature review followed by a qualitative study with expert participation. After performing a screening process, the initial number of items on the scale was reduced to 68. Ten dimensions (including management commitment, workers' empowerment, communication, blame culture, safety training, job satisfaction, interpersonal relationship, supervision, continuous improvement, and reward system) together with 37 items were extracted from the exploratory factor analysis (EFA) to measure safety climate. Acceptable ranges of internal consistency statistics for the sub-scales were observed. Confirmatory factor analysis (CFA) confirmed the construct validity of the developed safety climate scale for the petrochemical industry workers. The results of reliability showed that the Cronbach's alpha coefficient for the designed scale was 0.94. The ICC was obtained 0.92. This study created a valid and reliable scale for measuring safety climate in petrochemical industries.

  15. The practice of pre-marketing safety assessment in drug development.

    Science.gov (United States)

    Chuang-Stein, Christy; Xia, H Amy

    2013-01-01

    The last 15 years have seen a substantial increase in efforts devoted to safety assessment by statisticians in the pharmaceutical industry. While some of these efforts were driven by regulations and public demand for safer products, much of the motivation came from the realization that there is a strong need for a systematic approach to safety planning, evaluation, and reporting at the program level throughout the drug development life cycle. An efficient process can help us identify safety signals early and afford us the opportunity to develop effective risk minimization plan early in the development cycle. This awareness has led many pharmaceutical sponsors to set up internal systems and structures to effectively conduct safety assessment at all levels (patient, study, and program). In addition to process, tools have emerged that are designed to enhance data review and pattern recognition. In this paper, we describe advancements in the practice of safety assessment during the premarketing phase of drug development. In particular, we share examples of safety assessment practice at our respective companies, some of which are based on recommendations from industry-initiated working groups on best practice in recent years.

  16. NS [Nuclear Safety] update. Current safety and security activities and developments taking place in the Department of Nuclear Safety and Security, Issue no. 11, June 2009

    International Nuclear Information System (INIS)

    2009-06-01

    The current issue presents information about the following topics: Nuclear Safety Review for the Year 2008; Feedback from IRS Topical Studies and Events Applied to Safety Standards; Education and Training Programmes at the IAEA Department of Nuclear Safety and Security; Peer Review of Operational Safety Performance (PROSPER)

  17. NS [Nuclear Safety] update. Current safety and security activities and developments taking place in the Department of Nuclear Safety and Security, Issue no. 10, March 2009

    International Nuclear Information System (INIS)

    2009-03-01

    The current issue contains information about the following meetings: Application of the Code of Conduct on the Safety of Research Reactors (the 'Code'). Environmental Modelling for Radiation Safety (EMRAS II); Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management (the Joint Convention). The document also gives an overview on International Nuclear Security Advisory Service (INSServ)

  18. Recent developments in the study of safety culture and its correlates: the role of leadership

    International Nuclear Information System (INIS)

    Peiro Francisco, J. M.; Garcia, J.; Martinez-Corcoles, M.

    2015-01-01

    Assessing safety culture is a key element in the strategies of reliability assurance in nuclear facilities. A diagnosis that provides a return on investment (ROI) requires a scientifically based conceptualization of the most relevant variables in safety culture and outcomes expected from it. Furthermore, it is important to develop assessment tools and designs which allow, not only checking the state of safety culture but evaluating the correlates in the different productive levels of analysis, taking into account the temporal dimensional that allows us to analyze the change and thereby identify the main determinants and mechanisms for improvement. (Author)

  19. IAEA safety guides in the light of recent developments in earthquake engineering

    International Nuclear Information System (INIS)

    Gurpinar, A.

    1988-11-01

    The IAEA safety guides 50-SG-S1 and 50-SG-S2 emphasize on the determination of the design basis earthquake ground motion and earthquake resistant design considerations for nuclear power plants, respectively. Since the elaboration of these safety guides years have elapsed and a review of some of these concepts is necessary, taking into account the information collected and the technical developments. In this article, topics within the scope of these safety guides are discussed. In particular, the results of some recent research which may have a bearing on the nuclear industry are highlighted. Conclusions and recommendations are presented. 6 fig., 19 refs. (F.M.)

  20. Developing a patient safety incident classification system for primary care. A literature review and Delphi-survey by the LINNEAUS collaboration on patient safety in primary care.

    Science.gov (United States)

    Klemp, Kerstin; Dovey, Susan; Valderas, Jose M; Rohe, Julia; Godycki-Cwirko, Maciek; Elliott, Phillip; Beyer, Martin; Gerlach, Ferdin M; Hoffmann, Barbara

    2015-09-01

    Despite awareness that comparative analysis of patient safety data from several data sources would promote risk reduction, there has been little effort to establish an incident classification system that is generally applicable to patient safety data in European primary care. To describe the development of a patient safety incident classification system for primary care. A systematic review was followed by an expert group discussion and a modified Delphi survey, to provide consensus statements. We developed a classification system providing a mechanism for classifying patient safety incidents across Europe, taking into account the varying organizational arrangements that exist for primary care. It takes into account organizational processes and outcomes related to patient safety incidents and can supplement existing classification systems. Classification systems are key tools in the analysis of patient safety incidents. A system that has relevance for primary care is now available.

  1. Opportunities to Develop Programs and Engage Amish Youth in Safety Education

    Directory of Open Access Journals (Sweden)

    S. Dee Jepsen

    2012-06-01

    Full Text Available Understanding and designing appropriate educational youth safety programs for the Amish requires an appreciation of their history, their distinctiveness in an American society built on economic, social and cultural change, and how the Amish themselves have changed over the years. The qualitative research study highlighted in this paper sought to determine culturally and age-appropriate curricula useful to community educators interested in youth safety programs for Amish and other conservative Anabaptist groups. Researchers identified rural safety topics of interest to Amish families to include lawn mowers, string trimmers, chemicals, water, livestock, confined spaces, tractors and skid loaders. Parents regularly involved children in daily farm chores, where they made assignments based on the child’s physical development, maturity, interest in the task, and birth-order. Findings suggest opportunities for cooperative extension professionals to develop and engage Amish children in safety education programs.

  2. Assessment on the Development of Occupational Health and Safety Management Based on OHSAS 18001

    International Nuclear Information System (INIS)

    Sigit Santoso

    2006-01-01

    This paper focused on the safety of a workplace, while the majority of the discussion is emphasized in the development of occupational health and safety management of the process system. The assessment on a development of occupational health and safety management based on the OHSAS 18001 has been done. The result indicates that OHSAS 18001 as an assessment specification for occupational health and safety management systems can be applied to any type of organization and industry, eventhough it does not give detailed specifications for design in a management system. The extent of the application depend on such factors as the OH&S policy of the organization, the nature of its activities and the risks and complexity of its operations. (author)

  3. What can nanosafety learn from drug development? The feasibility of “safety by design”

    DEFF Research Database (Denmark)

    Hjorth, Rune; van Hove, Lilian; Wickson, Fern

    2017-01-01

    Safety by design” (SbD) is an intuitively appealing concept that is on the rise within nanotoxicology and nanosafety research, as well as within nanotechnology research policy. It leans on principles established within drug discovery and development (DDD) and seeks to address safety early, as well...... as throughout product development. However, it remains unclear what the concept of SbD exactly entails for engineered nanomaterials (ENMs) or how it is envisioned to be implemented. Here, we review the concept as it is emerging in European research and compare its resemblance with the safety testing...... reduction is clearly desirable, the way SbD is currently communicated tends to treat safety as an inherent material property and that this is fundamentally problematic as it represents a recasting and reduction of societal issues into technical problems. SbD therefore faces a multitude of challenges, from...

  4. Employing Wireless Networks in Enhancing Occupational Safety at the Developing Seaport – Two Proposals

    Directory of Open Access Journals (Sweden)

    Bauk Sanja

    2017-03-01

    Full Text Available The paper proposes two Wireless Body Area Network Sensors (WBANSs scenarios at the logical and simulation levels for improving occupational safety and health conditions at the developing seaport environment. The Port of Bar (Montenegro is taken as an exemplar. The logical model is based on the actual position of the Port of Bar at the seaport market, its needs and capacities for the information systems innovation through technology transfer and diffusion. The simulation model analyses the channel between the body central unit (BCU of the worker’s on port wireless body sub-network and the port access point. The quality of the signal transmission at the physical layer has been tested through a source code generated in the Matlab. The code includes the BCU composite signal modulation, transmission, and demodulation, along with a noise and fading effects analysis. The results of the simulation experiments for the different transmission frequencies and distances between transmitter (worker’s BCU and receiver (port’s access point by using binary phase-shift keying (BPSK and quadratic phase-shift keying (QPSK modulation schemes are presented. Some directions for further investigations in this field are given, as well.

  5. Safety analysis code SCTRAN development for SCWR and its application to CGNPC SCWR

    International Nuclear Information System (INIS)

    Wu, Pan; Gou, Junli; Shan, Jianqiang; Jiang, Yang; Yang, Jue; Zhang, Bo

    2013-01-01

    Highlights: ► A new safety analysis code named SCTRAN is developed for SCWRs. ► Capability of SCTRAN is verified by comparing with code APROS and RELAP5-3D. ► A new passive safety system is proposed for CGNPC SCWR and analyzed with SCTRAN. ► CGNPC SCWR is able to cope with two critical accidents for SCWRs, LOFA and LOCA. - Abstract: Design analysis is one of the main difficulties during the research and design of SCWRs. Currently, the development of safety analysis code for SCWR is still in its infancy all around the world, and very few computer codes could carry out the trans-critical calculations where significant changes in water properties would take place. In this paper, a safety analysis code SCTRAN for SCWRs has been developed based on code RETRAN-02, the best estimate code used for safety analysis of light water reactors. The ability of SCTRAN code to simulate transients where both supercritical and subcritical regimes are encountered has been verified by comparing with APROS and RELAP5-3D codes. Furthermore, the LOFA and LOCA transients for the CGNPC SCWR design were analyzed with SCTRAN code. The characteristics and performance of the passive safety systems applied to CGNPC SCWR were evaluated. The results show that: (1) The SCTRAN computer code developed in this study is capable to perform design analysis for SCWRs; (2) During LOFA and LOCA accidents in a CGNPC SCWR, the passive safety systems would significantly mitigate the consequences of these transients and enhance the inherent safety

  6. Developing a patient safety incident classification system for primary care. A literature review and Delphi-survey by the LINNEAUS collaboration on patient safety in primary care

    OpenAIRE

    Klemp, Kerstin; Dovey, Susan; Valderas, Jose M; Rohe, Julia; Godycki-Cwirko, Maciek; Elliott, Phillip; Beyer, Martin; Gerlach, Ferdin M.; Hoffmann, Barbara

    2015-01-01

    ABSTRACT Background: Despite awareness that comparative analysis of patient safety data from several data sources would promote risk reduction, there has been little effort to establish an incident classification system that is generally applicable to patient safety data in European primary care. Objective: To describe the development of a patient safety incident classification system for primary care. Methods: A systematic review was followed by an expert group discussion and a modified Delp...

  7. Developing a patient-led electronic feedback system for quality and safety within Renal PatientView.

    Science.gov (United States)

    Giles, Sally J; Reynolds, Caroline; Heyhoe, Jane; Armitage, Gerry

    2017-03-01

    It is increasingly acknowledged that patients can provide direct feedback about the quality and safety of their care through patient reporting systems. The aim of this study was to explore the feasibility of patients, healthcare professionals and researchers working in partnership to develop a patient-led quality and safety feedback system within an existing electronic health record (EHR), known as Renal PatientView (RPV). Phase 1 (inception) involved focus groups (n = 9) and phase 2 (requirements) involved cognitive walkthroughs (n = 34) and 1:1 qualitative interviews (n = 34) with patients and healthcare professionals. A Joint Services Expert Panel (JSP) was convened to review the findings from phase 1 and agree the core principles and components of the system prototype. Phase 1 data were analysed using a thematic approach. Data from phase 1 were used to inform the design of the initial system prototype. Phase 2 data were analysed using the components of heuristic evaluation, resulting in a list of core principles and components for the final system prototype. Phase 1 identified four main barriers and facilitators to patients feeding back on quality and safety concerns. In phase 2, the JSP agreed that the system should be based on seven core principles and components. Stakeholders were able to work together to identify core principles and components for an electronic patient quality and safety feedback system in renal services. Tensions arose due to competing priorities, particularly around anonymity and feedback. Careful consideration should be given to the feasibility of integrating a novel element with differing priorities into an established system with existing functions and objectives. © 2016 European Dialysis and Transplant Nurses Association/European Renal Care Association.

  8. FLUOR HANFORD SAFETY MANAGEMENT PROGRAMS

    Energy Technology Data Exchange (ETDEWEB)

    GARVIN, L. J.; JENSEN, M. A.

    2004-04-13

    This document summarizes safety management programs used within the scope of the ''Project Hanford Management Contract''. The document has been developed to meet the format and content requirements of DOE-STD-3009-94, ''Preparation Guide for US. Department of Energy Nonreactor Nuclear Facility Documented Safety Analyses''. This document provides summary descriptions of Fluor Hanford safety management programs, which Fluor Hanford nuclear facilities may reference and incorporate into their safety basis when producing facility- or activity-specific documented safety analyses (DSA). Facility- or activity-specific DSAs will identify any variances to the safety management programs described in this document and any specific attributes of these safety management programs that are important for controlling potentially hazardous conditions. In addition, facility- or activity-specific DSAs may identify unique additions to the safety management programs that are needed to control potentially hazardous conditions.

  9. Rationale for the closure of the soil density unreviewed safety question and recommended structural analyses improvements for the Tank Waste Remidiation System underground storage facilities

    International Nuclear Information System (INIS)

    Morris, K.H.

    1998-01-01

    The purpose of this report is twofold. First, this report documents the technical evaluation supporting the Project Hanford Management Contract (PHMC) contractor recommendation to close the Unreviewed Safety Question (USQ) as originally evaluated in TF-94-0260, Soil Compaction Test Data Indicates Soil Density in Excess of Density Used in Tank Qualification Analysis for AP Tank Farm. Second, this report describes the status of existing structural analyses for the Tank Waste Remediation System (TWRS) waste storage structures and outlines the associated technical upgrades being considered by the contractor. This second feature of the report serves to communicate the distinction between the soil density issue which is the topic of the open USQ and other technical issues which are important to the contractor from a programmatic standpoint. Contractor actions to address the latter technical issues would support improvements in day-to-day operations (e.g., provide possible relaxations in soil load restrictions) but are not necessary to close the soil density USQ. Section 2.0 of this report documents the rationale for the PHMC contractor recommendation to the Department of Energy (DOE) to close the soil density USQ. Section 3.0 documents the recommended structural analyses improvements for the double-shell tanks (DSTs) which are the structures associated with the soil density USQ. Sections 4.0 and 5.0 provide, for completeness, the same information for single-shell tanks (SSTs), double-contained receiver tanks (DCRTs), catch tanks and inactive miscellaneous underground storage tanks (IMUSTs). Section 6.0 provides the conclusions of this report

  10. Developments in safety and operations culture in BNFL's thorp reprocessing plant, Sellafield, Cumbria

    International Nuclear Information System (INIS)

    Kett, P.J.

    2000-01-01

    One of the best descriptions of Culture is 'how we do things around here'. In a stable organisation it is extremely difficult to change any type of culture, whether it is an operations, customer service or safety culture. To change culture one of two elements are essential. There must be either a significant external pressure felt by all in the organisation or a change in senior management, with authority to set a new direction for the organisation. BNFL had a unique opportunity through the commissioning and operation of the Thorp Reprocessing Plant at Sellafield to shape a new Safety and Operations Culture. Both the key elements for change were present. Thorp was a high profile flagship plant that had attracted multinational investment. It incorporated new technology. The workforce had volunteered to operate the plant. A strong senior management team was specially selected. The plant was being commissioned in an environment where there was significant opposition by 'anti nuclear' groups. It was essential to both BNFL and the wider international nuclear community that Thorp was commissioned and operated safely. A strong operating culture was developed with safety as the corner stone. The culture comprises three key components. Rigorous plant safety case and risk assessments before work commences and modifications to the plant occur; A high level of involvement by all levels of the workforce in both operations and safety matters; Strong supportive leadership which does not allow safety standards to be compromised and encourages open debate on how to improve. During commissioning and early operation of Thorp the robustness of the Safety and Operations Culture was demonstrated. On several occasions, despite intense commercial pressure, operations were halted until the situation was resolved both technically and procedurally. This paper describes how the Safety and Operations Culture was developed. The key factors for success include recruitment, team selection

  11. The Efficacy and Development of Students' Problem-Solving Strategies During Compulsory Schooling: Logfile Analyses

    Science.gov (United States)

    Molnár, Gyöngyvér; Csapó, Benő

    2018-01-01

    The purpose of this study was to examine the role of exploration strategies students used in the first phase of problem solving. The sample for the study was drawn from 3rd- to 12th-grade students (aged 9–18) in Hungarian schools (n = 4,371). Problems designed in the MicroDYN approach with different levels of complexity were administered to the students via the eDia online platform. Logfile analyses were performed to ascertain the impact of strategy use on the efficacy of problem solving. Students' exploration behavior was coded and clustered through Latent Class Analyses. Several theoretically effective strategies were identified, including the vary-one-thing-at-a-time (VOTAT) strategy and its sub-strategies. The results of the analyses indicate that the use of a theoretically effective strategy, which extract all information required to solve the problem, did not always lead to high performance. Conscious VOTAT strategy users proved to be the best problem solvers followed by non-conscious VOTAT strategy users and non-VOTAT strategy users. In the primary school sub-sample, six qualitatively different strategy class profiles were distinguished. The results shed new light on and provide a new interpretation of previous analyses of the processes involved in complex problem solving. They also highlight the importance of explicit enhancement of problem-solving skills and problem-solving strategies as a tool for knowledge acquisition in new contexts during and beyond school lessons. PMID:29593606

  12. Seismic Safety Margins Research Program Phase I final report: fragilities development (Project VI)

    International Nuclear Information System (INIS)

    Bohn, M.P.; Cover, L.E.; Dong, R.G.; Vagliente, V.N.; Campbell, R.D.; Wesley, D.A.

    1982-12-01

    The Seismic Safety Margins Research Program is an NRC-funded program directed towards estimating the conservatism in the NRC Standard Review Plan seismic safety requirements with the ultimate goal of developing improved requirements. As part of this program, calculations of the seismic risk from a typical commercial nuclear reactor are being made. These calculations require a knowledge of the probability of failure (fragility) of safety-related components in the reactor system which actively participate in the hypothesized accident scenarios. This report describes the development of the required fragility relations and the data sources and data reduction techniques upon which they are based. Both building and component fragilities are covered. The building fragilities are for the Zion Unit 1 reactor which was the specific plant used for development of methodology in the program. Some of the component fragilities are site-specific also, but most have been developed to apply to generic categories of components

  13. Development methodology for the software life cycle process of the safety software

    International Nuclear Information System (INIS)

    Kim, D. H.; Lee, S. S.; Cha, K. H.; Lee, C. S.; Kwon, K. C.; Han, H. B.

    2002-01-01

    A methodology for developing software life cycle processes (SLCP) is proposed to develop the digital safety-critical Engineered Safety Features - Component Control System (ESF-CCS) successfully. A software life cycle model is selected as the hybrid model mixed with waterfall, prototyping, and spiral models and is composed of two stages , development stages of prototype of ESF-CCS and ESF-CCS. To produce the software life cycle (SLC) for the Development of the Digital Reactor Safety System, the Activities referenced in IEEE Std. 1074-1997 are mapped onto the hybrid model. The SLCP is established after the available OPAs (Organizational Process Asset) are applied to the SLC Activities, and the known constraints are reconciled. The established SLCP describes well the software life cycle activities with which the Regulatory Authority provides

  14. On the Use of Safety Certification Practices in Autonomous Field Robot Software Development

    DEFF Research Database (Denmark)

    Mogensen, Johann Thor Ingibergsson; Schultz, Ulrik Pagh; Kuhrmann, Marco

    2015-01-01

    reactions or performance in malfunctioning systems, and influence industry regarding software development and project management. However, academia seemingly did not reach the same degree of utilisation of standards. This paper presents the findings from a systematic mapping study in which we study...... the state-of-the-art in developing software for safety-critical software for autonomous field robots. The purpose of the study is to identify practices used for the development of autonomous field robots and how these practices relate to available safety standards. Our findings from reviewing 49 papers show...

  15. Nuclear Criticality Safety Organization guidance for the development of continuing technical training. Revision 1

    International Nuclear Information System (INIS)

    Carroll, K.J.; Taylor, R.G.; Worley, C.A.

    1997-01-01

    The Nuclear Criticality Safety Organization (NCSO) is committed to developing and maintaining a staff of highly qualified personnel to meet the current and anticipated needs in nuclear criticality safety at the Oak Ridge Y-12 Plant and throughout the DOE complex. Continuing technical training is training outside of the initial qualification program to address identified organization-wide needs. Typically, this training is used to improve organization performance in the conduct of business. This document provides guidelines for the development of the technical portions of the Continuing Training Program. It is not a step-by-step procedure, but a collection of considerations to be used during the development process

  16. Integrating teamwork, clinician occupational well-being and patient safety - development of a conceptual framework based on a systematic review.

    Science.gov (United States)

    Welp, Annalena; Manser, Tanja

    2016-07-19

    There is growing evidence that teamwork in hospitals is related to both patient outcomes and clinician occupational well-being. Furthermore, clinician well-being is associated with patient safety. Despite considerable research activity, few studies include all three concepts, and their interrelations have not yet been investigated systematically. To advance our understanding of these potentially complex interrelations we propose an integrative framework taking into account current evidence and research gaps identified in a systematic review. We conducted a literature search in six major databases (Medline, PsycArticles, PsycInfo, Psyndex, ScienceDirect, and Web of Knowledge). Inclusion criteria were: peer reviewed papers published between January 2000 and June 2015 investigating a statistical relationship between at least two of the three concepts; teamwork, patient safety, and clinician occupational well-being in hospital settings, including practicing nurses and physicians. We assessed methodological quality using a standardized rating system and qualitatively appraised and extracted relevant data, such as instruments, analyses and outcomes. The 98 studies included in this review were highly diverse regarding quality, methodology and outcomes. We found support for the existence of independent associations between teamwork, clinician occupational well-being and patient safety. However, we identified several conceptual and methodological limitations. The main barrier to advancing our understanding of the causal relationships between teamwork, clinician well-being and patient safety is the lack of an integrative, theory-based, and methodologically thorough approach investigating the three concepts simultaneously and longitudinally. Based on psychological theory and our findings, we developed an integrative framework that addresses these limitations and proposes mechanisms by which these concepts might be linked. Knowledge about the mechanisms underlying the

  17. Maintaining knowledge, training and infrastructure for research and development in nuclear safety. A note by the International Nuclear Safety Advisory Group

    International Nuclear Information System (INIS)

    International Nuclear Safety Advisory Group

    2001-01-01

    The purpose of this INSAG Note is to emphasize the importance of maintaining capabilities for nuclear research and education, especially with regard to safety aspects, so that nuclear safety may be maintained in IAEA Member States, and to alert Member States to the potential for significant harm if the infrastructure for research, development and education is not maintained

  18. Development and Evaluation of a Multi-Institutional Case Studies-Based Course in Food Safety

    Science.gov (United States)

    Pleitner, Aaron M.; Chapin, Travis K.; Hammons, Susan R.; Stelten, Anna Van; Nightingale, Kendra K.; Wiedmann, Martin; Johnston, Lynette M.; Oliver, Haley F.

    2015-01-01

    Developing novel, engaging courses in food safety is necessary to train professionals in this discipline. Courses that are interactive and case-based encourage development of critical thinking skills necessary for identifying and preventing foodborne disease outbreaks. The purpose of this study was to assess the efficacy of a case study…

  19. Principles of developing the knowledge portal on safety of nuclear facilities

    International Nuclear Information System (INIS)

    Klevtsov, A.; Orlov, V.Yu.; Trubchaninov, S.A.

    2010-01-01

    The general principles of developing the knowledge portal on safety of nuclear facilities are considered in the article. In future, these principles can be used for implementing the project on development of the knowledge portal for the State Nuclear Regulatory Committee of Ukraine.

  20. Occupational safety and health in nanotechnology and Organisation for Economic Cooperation and Development

    Science.gov (United States)

    Murashov, Vladimir; Engel, Stefan; Savolainen, Kai; Fullam, Brian; Lee, Michelle; Kearns, Peter

    2009-10-01

    The Organization for Economic Cooperation and Development (OECD), an intergovernmental organization, is playing a critical global role in ensuring that emerging technologies, such as nanotechnology, are developed responsibly. This article describes OECD activities around occupational safety and health of nanotechnology and provides state-of-the-science overview resulting from an OECD workshop on exposure assessment and mitigation for nanotechnology workplace.