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)

Analysing context-dependent deviations in interacting with safety-critical systems

International Nuclear Information System (INIS)

Paterno, Fabio; Santoro, Carmen

2006-01-01

Mobile technology is penetrating many areas of human life. This implies that the context of use can vary in many respects. We present a method that aims to support designers in managing the complex design space when considering applications with varying contexts and help them to identify solutions that support users in performing their activities while preserving usability and safety. The method is a novel combination of an analysis of both potential deviations in task performance and most suitable information representations based on distributed cognition. The originality of the contribution is in providing a conceptual tool for better understanding the impact of context of use on user interaction in safety-critical domains. In order to present our approach we provide an example in which the implications of introducing new support through mobile devices in a safety-critical system are identified and analysed in terms of potential hazards

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)

Regulatory support activities of JNES by thermal-hydraulic and safety analyses

International Nuclear Information System (INIS)

Kasahara, Fumio

2008-01-01

Current status and some related topics on regulatory support activities of Japan Nuclear Energy Safety Organization (JNES) by thermal-hydraulic and safety analyses are reported. The safety of nuclear facilities is secured primarily by plant owners and operators. However, the regulatory body NISA (Nuclear and Industrial Safety Agency) has conducted a strict regulation to confirm the adequacy of the site condition as well as the basic and detailed design. The JNES has been conducting independent analyses from applicants (audit analyses, etc.) by direction of NISA and supporting its review. In addition to the audit analysis, thermal-hydraulic and safety analyses are used in such areas as analytical evaluation for investigation of causes of accidents and troubles, level 2 PSA for risk informed regulation, etc. Recent activities of audit analyses are for the application of Tsuruga 3 and 4 (APWR), the spent fuel storage facility for the establishment, and the LMFBR Monju for the core change. For the trouble event evaluation, the criticality accident analysis of Sika1 was carried out and the evaluation of effectiveness of accident management (AM) measure for Tomari 3 (PWR) and Monju was performed. The analytical codes for these evaluations are continuously revised by reflecting the state-of-art technical information and validated using the information provided by the data from JAEA, OECD project, etc. (author)

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

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

Energy Technology Data Exchange (ETDEWEB)

2009-11-15

deterioration over the assessment period. The basic approach for prescribing such margins is to consider whether the design assessed in SR-Can Main report was sufficient to result in safety. In case this design would imply too strict requirements, and in cases the SR-Can design was judged inadequate or not sufficiently analysed in the SR-Can report, some additional analyses have been undertaken to provide a better basis for setting the design premises. The resulting design premises constitute design constraints, which, if all fulfilled, form a good basis for demonstrating repository safety, according to the analyses in SR-Can and subsequent analyses. Some of the design premises may be modified in future stages of SKB's programme, as a result of analyses based on more detailed site data and a more developed understanding of processes of importance for long-term safety. Furthermore, a different balance between design requirements may result in the same level of safety. This report presents one technically reasonable balance, whereas future development and evaluations may result in other balances being deemed as more optimal. It should also be noted that in developing the reference design, the production reports should give credible evidence that the final product after construction and quality control fulfils the specifications of the reference design. To cover uncertainties in production and quality control that may be difficult to quantify in detail at the present design stage, the developer of the reference design need usually consider a margin to the conditions that would verify the design premises, but whether there is a need for such margins lies outside the scope of the current document. The term 'withstand' is used in this document in descriptions of load cases on repository components. The statement that a component withstands a particular load means that it upholds its related safety function when exposed to the load in question. For example, if the

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

deterioration over the assessment period. The basic approach for prescribing such margins is to consider whether the design assessed in SR-Can Main report was sufficient to result in safety. In case this design would imply too strict requirements, and in cases the SR-Can design was judged inadequate or not sufficiently analysed in the SR-Can report, some additional analyses have been undertaken to provide a better basis for setting the design premises. The resulting design premises constitute design constraints, which, if all fulfilled, form a good basis for demonstrating repository safety, according to the analyses in SR-Can and subsequent analyses. Some of the design premises may be modified in future stages of SKB's programme, as a result of analyses based on more detailed site data and a more developed understanding of processes of importance for long-term safety. Furthermore, a different balance between design requirements may result in the same level of safety. This report presents one technically reasonable balance, whereas future development and evaluations may result in other balances being deemed as more optimal. It should also be noted that in developing the reference design, the production reports should give credible evidence that the final product after construction and quality control fulfils the specifications of the reference design. To cover uncertainties in production and quality control that may be difficult to quantify in detail at the present design stage, the developer of the reference design need usually consider a margin to the conditions that would verify the design premises, but whether there is a need for such margins lies outside the scope of the current document. The term 'withstand' is used in this document in descriptions of load cases on repository components. The statement that a component withstands a particular load means that it upholds its related safety function when exposed to the load in question. For example, if the canister is said to

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

Heat transfer calculations for the High Flux Isotope Reactor (HFIR). Technical specifications: bases for safety limits and limiting safety system settings

International Nuclear Information System (INIS)

Sims, T.M.; Swanks, J.H.

1977-09-01

Heat transfer analyses, in support of the preparation of the HFIR technical specifications, were made to establish the bases for the safety limits and limiting safety system settings applicable to the HFIR. The results of these analyses, along with the detailed bases, are presented

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.)

Building patient safety in intensive care nursing : Patient safety culture, team performance and simulation-based training

OpenAIRE

Ballangrud, Randi

2013-01-01

Aim: The overall aim of the thesis was to investigate patient safety culture, team performance and the use of simulation-based team training for building patient safety in intensive care nursing. Methods: Quantitative and qualitative methods were used. In Study I, 220 RNs from ten ICUs responded to a patient safety culture questionnaire analysed with statistics. Studies II-IV were based on an evaluation of a simulation-based team training programme. Studies II-III included 53 RNs from seven I...

Safety balance: Analysis of safety systems; Bilans de surete: analyse par les organismes de surete

Energy Technology Data Exchange (ETDEWEB)

Delage, M; Giroux, C

1990-12-01

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

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)

Safety and sensitivity analyses of a generic geologic disposal system for high-level radioactive waste

International Nuclear Information System (INIS)

Kimura, Hideo; Takahashi, Tomoyuki; Shima, Shigeki; Matsuzuru, Hideo

1994-11-01

This report describes safety and sensitivity analyses of a generic geologic disposal system for HLW, using a GSRW code and an automated sensitivity analysis methodology based on the Differential Algebra. An exposure scenario considered here is based on a normal evolution scenario which excludes events attributable to probabilistic alterations in the environment. The results of sensitivity analyses indicate that parameters related to a homogeneous rock surrounding a disposal facility have higher sensitivities to the output analyzed here than those of a fractured zone and engineered barriers. The sensitivity analysis methodology provides technical information which might be bases for the optimization of design of the disposal facility. Safety analyses were performed on the reference disposal system which involve HLW in amounts corresponding to 16,000 MTU of spent fuels. The individual dose equivalent due to the exposure pathway ingesting drinking water was calculated using both the conservative and realistic values of geochemical parameters. In both cases, the committed dose equivalent evaluated here is the order of 10 -7 Sv, and thus geologic disposal of HLW may be feasible if the disposal conditions assumed here remain unchanged throughout the periods assessed here. (author)

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

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

Nuclear power plants: Results of recent safety analyses

International Nuclear Information System (INIS)

Steinmetz, E.

1987-01-01

The contributions deal with the problems posed by low radiation doses, with the information currently available from analyses of the Chernobyl reactor accident, and with risk assessments in connection with nuclear power plant accidents. Other points of interest include latest results on fission product release from reactor core or reactor building, advanced atmospheric dispersion models for incident and accident analyses, reliability studies on safety systems, and assessment of fire hazard in nuclear installations. The various contributions are found as separate entries in the database. (DG) [de

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.)

Safety systems I/C equipment reliability analyses of the Kozloduy NPP units 3 and 4

Energy Technology Data Exchange (ETDEWEB)

Halev, G; Christov, N [Risk Engineering Ltd., Sofia (Bulgaria)

1996-12-31

The purpose of the analysis is to assess the safety systems I/C equipment reliability. The assessment includes: quantification of the safety systems unavailability due to component failures; definition of the minimal cut sets leading to the analysed safety systems failure; quantification of the I/C equipment importance measures of the dominant contribution components. The safety systems I/C equipment reliability has been analysed using PSAPACK (a code for probabilistic safety assessment). Fault trees for the following safety systems of the Kozloduy-3 and Kozloduy-4 reactors have been constructed: neutron flow control equipment, reactor protection system, main coolant pumps, pressurizer safety valves `Sempell`, steam dump systems, spray system, low pressure injection system, emergency feeding water system, essential service water system. THree separate reports have been issued containing the performed analyses and results. 1 ref.

Quality assurance requirements for the computer software and safety analyses

International Nuclear Information System (INIS)

Husarecek, J.

1992-01-01

The requirements are given as placed on the development, procurement, maintenance, and application of software for the creation or processing of data during the design, construction, operation, repair, maintenance and safety-related upgrading of nuclear power plants. The verification and validation processes are highlighted, and the requirements put on the software documentation are outlined. The general quality assurance principles applied to safety analyses are characterized. (J.B.). 1 ref

Safety And Transient Analyses For Full Core Conversion Of The Dalat Nuclear Research Reactor

International Nuclear Information System (INIS)

Luong Ba Vien; Le Vinh Vinh; Huynh Ton Nghiem; Nguyen Kien Cuong

2011-01-01

Preparing for full core conversion of Dalat Nuclear Research Reactor (DNRR), safety and transient analyses were carried out to confirm about ability to operate safely of proposed Low Enriched Uranium (LEU) working core. The initial LEU core consisting 92 LEU fuel assemblies and 12 Beryllium rods was analyzed under initiating events of uncontrolled withdrawal of a control rod, cooling pump failure, earthquake and fuel cladding fail. Working LEU core response were evaluated under these initial events based on RELAP/Mod3.2 computer code and other supported codes like ORIGEN, MCNP and MACCS2. Obtained results showed that safety of the reactor is maintained for all transients/accidents analyzed. (author)

Sensitivity and uncertainty analyses applied to criticality safety validation. Volume 2

International Nuclear Information System (INIS)

Broadhead, B.L.; Hopper, C.M.; Parks, C.V.

1999-01-01

This report presents the application of sensitivity and uncertainty (S/U) analysis methodologies developed in Volume 1 to the code/data validation tasks of a criticality safety computational study. Sensitivity and uncertainty analysis methods were first developed for application to fast reactor studies in the 1970s. This work has revitalized and updated the existing S/U computational capabilities such that they can be used as prototypic modules of the SCALE code system, which contains criticality analysis tools currently in use by criticality safety practitioners. After complete development, simplified tools are expected to be released for general use. The methods for application of S/U and generalized linear-least-square methodology (GLLSM) tools to the criticality safety validation procedures were described in Volume 1 of this report. Volume 2 of this report presents the application of these procedures to the validation of criticality safety analyses supporting uranium operations where enrichments are greater than 5 wt %. Specifically, the traditional k eff trending analyses are compared with newly developed k eff trending procedures, utilizing the D and c k coefficients described in Volume 1. These newly developed procedures are applied to a family of postulated systems involving U(11)O 2 fuel, with H/X values ranging from 0--1,000. These analyses produced a series of guidance and recommendations for the general usage of these various techniques. Recommendations for future work are also detailed

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

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

Energy Technology Data Exchange (ETDEWEB)

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.

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

Best Estimate plus Uncertainty (BEPU) Analyses in the IAEA Safety Standards

International Nuclear Information System (INIS)

Dusic, Milorad; )

2013-01-01

The Safety Standards Series establishes an essential basis for safety and represents the broadest international consensus. Safety Standards Series publications are categorized into: Safety Fundamental (Present the overall objectives, concepts and principles of protection and safety, they are the policy documents of the safety standards), Safety Requirements (Establish requirements that must be met to ensure the protection and safety of people and the environment, both now and in the future), and Safety Guides (Provide guidance, in the form of more detailed actions, conditions or procedures that can be used to comply with the Requirements). The incorporation of more detailed requirements, in accordance with national practice, may still be necessary. There should be only one set of international safety standards. Each safety standard will be reviewed by the relevant committee or by the commission every five years. Best Estimate plus Uncertainty (BEPU) Analyses are approached in the following IAEA Safety Standards: - Safety Requirements SSR 2/1 - Safety of NPPs, Design (Revision of NS-R-1); - General Safety Requirement GSR Part 4: Safety Assessment for Facilities and Activities; - Safety Guide SSG-2 Deterministic Safety Analysis for Nuclear Power Plants. NUSSC suggested that new safety guides should be accompanied by documents like TECDOCs or Safety Reports describing in detail their recommendations where appropriate. Special review is currently underway to identify needs for revision in the light of the Fukushima accident. Revision will concern, first, the Safety Requirements, and then, the Selected Safety Guides

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

Science.gov (United States)

Dozza, Marco; González, Nieves Pañeda

2013-11-01

New trends in research on traffic accidents include Naturalistic Driving Studies (NDS). NDS are based on large scale data collection of driver, vehicle, and environment information in real world. NDS data sets have proven to be extremely valuable for the analysis of safety critical events such as crashes and near crashes. However, finding safety critical events in NDS data is often difficult and time consuming. Safety critical events are currently identified using kinematic triggers, for instance searching for deceleration below a certain threshold signifying harsh braking. Due to the low sensitivity and specificity of this filtering procedure, manual review of video data is currently necessary to decide whether the events identified by the triggers are actually safety critical. Such reviewing procedure is based on subjective decisions, is expensive and time consuming, and often tedious for the analysts. Furthermore, since NDS data is exponentially growing over time, this reviewing procedure may not be viable anymore in the very near future. This study tested the hypothesis that automatic processing of driver video information could increase the correct classification of safety critical events from kinematic triggers in naturalistic driving data. Review of about 400 video sequences recorded from the events, collected by 100 Volvo cars in the euroFOT project, suggested that drivers' individual reaction may be the key to recognize safety critical events. In fact, whether an event is safety critical or not often depends on the individual driver. A few algorithms, able to automatically classify driver reaction from video data, have been compared. The results presented in this paper show that the state of the art subjective review procedures to identify safety critical events from NDS can benefit from automated objective video processing. In addition, this paper discusses the major challenges in making such video analysis viable for future NDS and new potential

Combining energy and power based safety metrics in controller design for domestic robots

NARCIS (Netherlands)

Tadele, T.S.; de Vries, Theodorus J.A.; Stramigioli, Stefano

This paper presents a general passivity based interaction controller design approach that utilizes a combined energy and power based safety norms to assert safety of domestic robots. Since these robots are expected to co-habit the same environment with a human user, analysing and ensuring their

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

German data for risk based fire safety assessment

International Nuclear Information System (INIS)

Roewekamp, M.; Berg, H.P.

1998-01-01

Different types of data are necessary to perform risk based fire safety assessments and, in particular, to quantify the fire event tree considering the plant specific conditions. Data on fire barriers, fire detection and extinguishing, including also data on secondary effects of a fire, have to be used for quantifying the potential hazard and damage states. The existing German database on fires in nuclear power plants (NPPs) is very small. Therefore, in general generic data, mainly from US databases, are used for risk based safety assessments. Due to several differences in the plant design and conditions generic data can only be used as conservative assumptions. World-wide existing generic data on personnel failures in case of fire fighting have only to be adapted to the plant specific conditions inside the NPP to be investigated. In contrary, unavailabilities of fire barrier elements may differ strongly depending on different standards, testing requirements, etc. In addition, the operational behaviour of active fire protection equipment may vary depending on type and manufacturer. The necessity for more detailed and for additional plant specific data was the main reason for generating updated German data on the operational behaviour of active fire protection equipment/features in NPPs to support risk based fire safety analyses being recommended to be carried out as an additional tool to deterministic fire hazard analyses in the frame of safety reviews. The results of these investigations revealed a broader and more realistic database for technical reliability of active fire protection means, but improvements as well as collection of further data are still necessary. (author)

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)

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)

Reliability analysis of software based safety functions

International Nuclear Information System (INIS)

Pulkkinen, U.

1993-05-01

The methods applicable in the reliability analysis of software based safety functions are described in the report. Although the safety functions also include other components, the main emphasis in the report is on the reliability analysis of software. The check list type qualitative reliability analysis methods, such as failure mode and effects analysis (FMEA), are described, as well as the software fault tree analysis. The safety analysis based on the Petri nets is discussed. The most essential concepts and models of quantitative software reliability analysis are described. The most common software metrics and their combined use with software reliability models are discussed. The application of software reliability models in PSA is evaluated; it is observed that the recent software reliability models do not produce the estimates needed in PSA directly. As a result from the study some recommendations and conclusions are drawn. The need of formal methods in the analysis and development of software based systems, the applicability of qualitative reliability engineering methods in connection to PSA and the need to make more precise the requirements for software based systems and their analyses in the regulatory guides should be mentioned. (orig.). (46 refs., 13 figs., 1 tab.)

Achieving reasonable conservatism in nuclear safety analyses

International Nuclear Information System (INIS)

Jamali, Kamiar

2015-01-01

In the absence of methods that explicitly account for uncertainties, seeking reasonable conservatism in nuclear safety analyses can quickly lead to extreme conservatism. The rate of divergence to extreme conservatism is often beyond the expert analysts’ intuitive feeling, but can be demonstrated mathematically. Too much conservatism in addressing the safety of nuclear facilities is not beneficial to society. Using certain properties of lognormal distributions for representation of input parameter uncertainties, example calculations for the risk and consequence of a fictitious facility accident scenario are presented. Results show that there are large differences between the calculated 95th percentiles and the extreme bounding values derived from using all input variables at their upper-bound estimates. Showing the relationship of the mean values to the key parameters of the output distributions, the paper concludes that the mean is the ideal candidate for representation of the value of an uncertain parameter. The mean value is proposed as the metric that is consistent with the concept of reasonable conservatism in nuclear safety analysis, because its value increases towards higher percentiles of the underlying positively skewed distribution with increasing levels of uncertainty. Insensitivity of the results to the actual underlying distributions is briefly demonstrated. - Highlights: • Multiple conservative assumptions can quickly diverge into extreme conservatism. • Mathematics and attractive properties provide basis for wide use of lognormal distribution. • Mean values are ideal candidates for representation of parameter uncertainties. • Mean values are proposed as reasonably conservative estimates of parameter uncertainties

Safety and deterministic failure analyses in high-beta D-D tokamak reactors

International Nuclear Information System (INIS)

Selcow, E.C.

1984-01-01

Safety and deterministic failure analyses were performed to compare major component failure characteristics for different high-beta D-D tokamak reactors. The primary focus was on evaluating damage to the reactor facility. The analyses also considered potential hazards to the general public and operational personnel. Parametric designs of high-beta D-D tokamak reactors were developed, using WILDCAT as the reference. The size, and toroidal field strength were reduced, and the fusion power increased in an independent manner. These changes were expected to improve the economics of D-D tokamaks. Issues examined using these designs were radiation induced failurs, radiation safety, first wall failure from plasma disruptions, and toroidal field magnet coil failure

Passive safety injection experiments and analyses (PAHKO)

International Nuclear Information System (INIS)

Tuunanen, J.

1998-01-01

PAHKO project involved experiments on the PACTEL facility and computer simulations of selected experiments. The experiments focused on the performance of Passive Safety Injection Systems (PSIS) of Advanced Light Water Reactors (ALWRs) in Small Break Loss-Of-Coolant Accident (SBLOCA) conditions. The PSIS consisted of a Core Make-up Tank (CMT) and two pipelines (Pressure Balancing Line, PBL, and Injection Line, IL). The examined PSIS worked efficiently in SBLOCAs although the flow through the PSIS stopped temporarily if the break was very small and the hot water filled the CMT. The experiments demonstrated the importance of the flow distributor in the CMT to limit rapid condensation. The project included validation of three thermal-hydraulic computer codes (APROS, CATHARE and RELAP5). The analyses showed the codes are capable to simulate the overall behaviour of the transients. The detailed analyses of the results showed some models in the codes still need improvements. Especially, further development of models for thermal stratification, condensation and natural circulation flow with small driving forces would be necessary for accurate simulation of the PSIS phenomena. (orig.)

Safety design analyses of Korea Advanced Liquid Metal Reactor

International Nuclear Information System (INIS)

Suk, S.D.; Park, C.K.

2000-01-01

The national long-term R and D program updated in 1997 requires Korea Atomic Energy Research Institute (KAERI) to complete by the year 2006 the basic design of Korea Advanced Liquid Metal Reactor (KALIMER), along with supporting R and D work, with the capability of resolving the issue of spent fuel storage as well as with significantly enhanced safety. KALIMER is a 150 MWe pool-type sodium cooled prototype reactor that uses metallic fuel. The conceptual design is currently under way to establish a self consistent design meeting a set of the major safety design requirements for accident prevention. Some of current emphasis include those for inherent and passive means of negative reactivity insertion and decay heat removal, high shutdown reliability, prevention of and protection from sodium chemical reaction, and high seismic margin, among others. All of these requirements affect the reactor design significantly and involve supporting R and D programs of substance. This paper summarizes some of the results of engineering and design analyses performed for the safety of KALIMER. (author)

Risk based limits for Operational Safety Requirements

International Nuclear Information System (INIS)

Cappucci, A.J. Jr.

1993-01-01

OSR limits are designed to protect the assumptions made in the facility safety analysis in order to preserve the safety envelope during facility operation. Normally, limits are set based on ''worst case conditions'' without regard to the likelihood (frequency) of a credible event occurring. In special cases where the accident analyses are based on ''time at risk'' arguments, it may be desirable to control the time at which the facility is at risk. A methodology has been developed to use OSR limits to control the source terms and the times these source terms would be available, thus controlling the acceptable risk to a nuclear process facility. The methodology defines a new term ''gram-days''. This term represents the area under a source term (inventory) vs time curve which represents the risk to the facility. Using the concept of gram-days (normalized to one year) allows the use of an accounting scheme to control the risk under the inventory vs time curve. The methodology results in at least three OSR limits: (1) control of the maximum inventory or source term, (2) control of the maximum gram-days for the period based on a source term weighted average, and (3) control of the maximum gram-days at the individual source term levels. Basing OSR limits on risk based safety analysis is feasible, and a basis for development of risk based limits is defensible. However, monitoring inventories and the frequencies required to maintain facility operation within the safety envelope may be complex and time consuming

Towards a Competency-based Vision for Construction Safety Education

Science.gov (United States)

Pedro, Akeem; Hai Chien, Pham; Park, Chan Sik

2018-04-01

Accidents still prevail in the construction industry, resulting in injuries and fatalities all over the world. Educational programs in construction should deliver safety knowledge and skills to students who will become responsible for ensuring safe construction work environments in the future. However, there is a gap between the competencies current pedagogical approaches target, and those required for safety in practice. This study contributes to addressing this issue in three steps. Firstly, a vision for competency-based construction safety education is conceived. Building upon this, a research scheme to achieve the vision is developed, and the first step of the scheme is initiated in this study. The critical competencies required for safety education are investigated through analyses of literature, and confirmed through surveys with construction and safety management professionals. Results from the study would be useful in establishing and orienting education programs towards current industry safety needs and requirements

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)

The impact of safety analyses on the design of the Hanford Waste Vitrification Plant

International Nuclear Information System (INIS)

Koppenaal, T.J.; Yee, A.K.; Reisdorf, J.B.; Hall, B.W.

1993-04-01

Accident analyses are being performed to evaluate and document the safety of the Hanford Waste Vitrification Plant (HWVP). The safety of the HWVP is assessed by evaluating worst-case accident scenarios and determining the dose to offsite and onsite receptors. Air dispersion modeling is done with the GENII computer code. Three accidents are summarized in this paper, and their effects on the safety and the design of the HWVP are demonstrated

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

International Nuclear Information System (INIS)

Wetzel, N.; Scharfe, A.

1998-01-01

The present report examines the possibilities and limits of a probabilistic safety analysis to evaluate passive safety systems and inherent safety characteristics. The inherent safety characteristics are based on physical principles, that together with the safety system lead to no damage. A probabilistic evaluation of the inherent safety characteristic is not made. An inventory of passive safety systems of accomplished nuclear power plant types in the Federal Republic of Germany was drawn up. The evaluation of the passive safety system in the analysis of the accomplished nuclear power plant types was examined. The analysis showed that the passive manner of working was always assumed to be successful. A probabilistic evaluation was not performed. The unavailability of the passive safety system was determined by the failure of active components which are necessary in order to activate the passive safety system. To evaluate the passive safety features in new concepts of nuclear power plants the AP600 from Westinghouse, the SBWR from General Electric and the SWR 600 from Siemens, were selected. Under these three reactor concepts, the SWR 600 is specially attractive because the safety features need no energy sources and instrumentation in this concept. First approaches for the assessment of the reliability of passively operating systems are summarized. Generally it can be established that the core melt frequency for the passive concepts AP600 and SBWR is advantageous in comparison to the probabilistic objectives from the European Pressurized Water Reactor (EPR). Under the passive concepts is the SWR 600 particularly interesting. In this concept the passive systems need no energy sources and instrumentation, and has active operational systems and active safety equipment. Siemens argues that with this concept the frequency of a core melt will be two orders of magnitude lower than for the conventional reactors. (orig.) [de

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

Risk-based safety indicators

International Nuclear Information System (INIS)

Szikszai, T.

1997-01-01

The presentation discusses the following issues: The objectives of the risk-based indicator programme. The characteristics of the risk-based indicators. The objectives of risk-based safety indicators - in monitoring safety; in PSA applications. What indicators? How to produce the risk based indicators? PSA requirements

Safety analyses for NHR-200

Energy Technology Data Exchange (ETDEWEB)

Jincai, Li; Zuying, Gao; Baocheng, Xu; Junxiao, He [Institute of Nuclear Energy and Technology, Tsingua Univ., Beijing (China)

1997-09-01

The NHR-200 is a commercial 200-MW District Heating Reactor developed in China. It is designed on the basis of design, construction and four-year operating experience of the 5MW Experimental Heating Reactor (NHR-5). It has special safety features which are briefly described in this paper. Accident classification and safety criteria are also explained. Some typical and serious accidents are studied theoretically, and their results are detailed in this paper. They demonstrate the excellent safety characteristics of HR-200. (author). 4 refs, 9 figs, 1 tab.

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

Internet of Things Based Combustible Ice Safety Monitoring System Framework

Science.gov (United States)

Sun, Enji

2017-05-01

As the development of human society, more energy is requires to meet the need of human daily lives. New energies play a significant role in solving the problems of serious environmental pollution and resources exhaustion in the present world. Combustible ice is essentially frozen natural gas, which can literally be lit on fire bringing a whole new meaning to fire and ice with less pollutant. This paper analysed the advantages and risks on the uses of combustible ice. By compare to other kinds of alternative energies, the advantages of the uses of combustible ice were concluded. The combustible ice basic physical characters and safety risks were analysed. The developments troubles and key utilizations of combustible ice were predicted in the end. A real-time safety monitoring system framework based on the internet of things (IOT) was built to be applied in the future mining, which provide a brand new way to monitoring the combustible ice mining safety.

Preliminary safety evaluation, based on initial site investigation data. Planning document

International Nuclear Information System (INIS)

Hedin, Allan

2002-12-01

This report is a planning document for the preliminary safety evaluations (PSE) to be carried out at the end of the initial stage of SKBs ongoing site investigations for a deep repository for spent nuclear fuel. The main purposes of the evaluations are to determine whether earlier judgements of the suitability of the candidate area for a deep repository with respect to long-term safety holds up in the light of borehole data and to provide feed-back to continued site investigations and site specific repository design. The preliminary safety evaluations will be carried out by a safety assessment group, based on a site model, being part of a site description, provided by a site modelling group and a repository layout within that model suggested by a repository engineering group. The site model contains the geometric features of the site as well as properties of the host rock. Several alternative interpretations of the site data will likely be suggested. Also the biosphere is included in the site model. A first task for the PSE will be to compare the rock properties described in the site model to previously established criteria for a suitable host rock. This report gives an example of such a comparison. In order to provide more detailed feedback, a number of thermal, hydrological, mechanical and chemical analyses of the site will also be included in the evaluation. The selection of analyses is derived from the set of geosphere and biosphere analyses preliminarily planned for the comprehensive safety assessment named SR-SITE, which will be based on a complete site investigation. The selection is dictated primarily by the expected feedback to continued site investigations and by the availability of data after the PSE. The repository engineering group will consider several safety related factors in suggesting a repository layout: Thermal calculations will be made to determine a minimum distance between canisters avoiding canister surface temperatures above 100 deg C

Use of the deterministic safety analyses in support to the NPP Krsko modification

International Nuclear Information System (INIS)

Feretic, D.; Cavlina, N.; Debrecin, N.; Grgic, D.; Bajs, T.; Spalj, S.

2004-01-01

The ultimate goal of the safety analysis is to verify that Nuclear Power Plant (NPP) meets safety and operational requirements. To this aim it is necessary to demonstrate that plant safety has not been deteriorated in the case of the modifications to the plant Systems, Structures and Components (SSC) or changes to the plant procedures. In addition, safety analyses are needed in the case of reassessment of an existing plant. The reasons for reassessment may be different, e.g. due to the changes in the methodology and assumptions used in the original design, if the original design basis or acceptance criteria may no longer be adequate, if the safety analysis tools used may have been superseded by more sophisticated methods or if the original design basis may no longer be met. The operation of the NPP Krsko has experienced numerous changes from the original design for the majority of the reasons that have been mentioned before. On the other side, the application of the large best-estimate thermalhydraulic codes has evolved to the wide spread support in the operation of the NPP: compliance with the regulatory goals, support to the PSA studies, analysis of the operational transients, plant modifications studies, equipment qualification, training of the operators, preparation of the operating procedures, etc. This trend has been followed at the Faculty of Electrical Engineering Zagreb (FER) and applied to the on-going needs due to the modifications and changes at NPP Krsko. In this paper, an overview of the deterministic safety analyses performed at FER in the support to the NPP Krsko modifications and changes is presented.(author)

Risk-based safety indicators

International Nuclear Information System (INIS)

Sedlak, J.

2001-12-01

The report is structured as follows: 1. Risk-based safety indicators: Typology of risk-based indicators (RBIs); Tools for defining RBIs; Requirements for the PSA model; Data sources for RBIs; Types of risks monitored; RBIs and operational safety indicators; Feedback from operating experience; PSO model modification for RBIs; RBI categorization; RBI assessment; RBI applications; Suitable RBI applications. 2. Proposal for risk-based indicators: Acquiring information from operational experience; Method of acquiring safety relevance coefficients for the systems from a PSA model; Indicator definitions; On-line indicators. 3. Annex: Application of RBIs worldwide. (P.A.)

Evaluation of geological documents available for provisional safety analyses of potential sites for nuclear waste repositories - Are additional geological investigations needed?

International Nuclear Information System (INIS)

2010-10-01

The procedure for selecting repository sites for all categories of radioactive waste in Switzerland is defined in the conceptual part of the Sectoral Plan for Deep Geological Repositories, which foresees a selection of sites in three stages. In Stage I, Nagra proposed geological siting regions based on criteria relating to safety and engineering feasibility. The Swiss Government (the Federal Council) is expected to decide on the siting proposals in 2011. The objective of Stage 2 is to prepare proposals for the location of the surface facilities within the planning perimeters defined by the Federal Council in its decision on Stage 1 and to identify potential sites. Nagra also has to carry out a provisional safety analysis for each site and a safety-based comparison of the sites. Based on this, and taking into account the results of the socio-economic-ecological impact studies, Nagra then has to propose at least two sites for each repository type to be carried through to Stage 3. The proposed sites will then be investigated in more detail in Stage 3 to ensure that the selection of the sites for the General Licence Applications is well founded. In order to realise the objectives of the upcoming Stage 2, the state of knowledge of the geological conditions at the sites has to be sufficient to perform the provisional safety analyses. Therefore, in preparation for Stage 2, the conceptual part of the Sectoral Plan requires Nagra to clarify the need for additional investigations aimed at providing input for the provisional safety analyses. The purpose of the present report is to document Nagra's technical-scientific assessment of this need. The focus is on evaluating the geological information based on processes and parameters that are relevant for safety and engineering feasibility. In evaluating the state of knowledge the key question is whether additional information could lead to a different decision regarding the selection of the sites to be carried through to Stage 3

Validation of risk-based performance indicators: Safety system function trends

International Nuclear Information System (INIS)

Boccio, J.L.; Vesely, W.E.; Azarm, M.A.; Carbonaro, J.F.; Usher, J.L.; Oden, N.

1989-10-01

This report describes and applies a process for validating a model for a risk-based performance indicator. The purpose of the risk-based indicator evaluated, Safety System Function Trend (SSFT), is to monitor the unavailability of selected safety systems. Interim validation of this indicator is based on three aspects: a theoretical basis, an empirical basis relying on statistical correlations, and case studies employing 25 plant years of historical data collected from five plants for a number of safety systems. Results using the SSFT model are encouraging. Application of the model through case studies dealing with the performance of important safety systems shows that statistically significant trends in, and levels of, system performance can be discerned which thereby can provide leading indications of degrading and/or improving performances. Methods for developing system performance tolerance bounds are discussed and applied to aid in the interpretation of the trends in this risk-based indicator. Some additional characteristics of the SSFT indicator, learned through the data-collection efforts and subsequent data analyses performed, are also discussed. The usefulness and practicality of other data sources for validation purposes are explored. Further validation of this indicator is noted. Also, additional research is underway in developing a more detailed estimator of system unavailability. 9 refs., 18 figs., 5 tabs

A fuzzy-logic-based approach to qualitative safety modelling for marine systems

International Nuclear Information System (INIS)

Sii, H.S.; Ruxton, Tom; Wang Jin

2001-01-01

Safety assessment based on conventional tools (e.g. probability risk assessment (PRA)) may not be well suited for dealing with systems having a high level of uncertainty, particularly in the feasibility and concept design stages of a maritime or offshore system. By contrast, a safety model using fuzzy logic approach employing fuzzy IF-THEN rules can model the qualitative aspects of human knowledge and reasoning processes without employing precise quantitative analyses. A fuzzy-logic-based approach may be more appropriately used to carry out risk analysis in the initial design stages. This provides a tool for working directly with the linguistic terms commonly used in carrying out safety assessment. This research focuses on the development and representation of linguistic variables to model risk levels subjectively. These variables are then quantified using fuzzy sets. In this paper, the development of a safety model using fuzzy logic approach for modelling various design variables for maritime and offshore safety based decision making in the concept design stage is presented. An example is used to illustrate the proposed approach

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.

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

Thermal and stress analyses of meltdown cups for LMFBR safety experiments using SLSF in-reactor loops

International Nuclear Information System (INIS)

Blomquist, C.A.; Pierce, R.D.; Pedersen, D.R.; Ariman, T.

1977-01-01

The test trains for the Sodium Loop Safety Facility (SLSF) in-reactor experiments, which simulate hypothetical LMFBR accidents, have a meltdown cup to protect the primary containment from the effects of molten materials. Thermal and stress analyses were performed on the cup which is designed to contain 3.6 kg of molten fuel and 2.4 kg of molten steel. Thermal analyses were performed with the Argonne-modified version fo the general heat transfer code THTB, based on the instantaneous addition of 3200 0 K molten fuel with a decay heat of 9 W/gm and 1920 0 K molten steel. These analyses have shown that the cup will adequately cool the molten materials. The stress analysis showed that the Inconel vessel would not fail from the pressure loading, it was also shown that brittle fracture of the tungsten liner from thermal gradients is unlikely. Therefore, the melt-down cup meets the structural design requirements. (Auth.)

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)

Safety analyses for an in-pile SCWR fuel qualification test loop

Energy Technology Data Exchange (ETDEWEB)

Schulenberg, T.; Raque, M. [Karlsruhe Inst. of Tech., Karlsruhe (Germany)

2014-07-01

As a nuclear facility cooled with supercritical water has never been built nor operated in the past, the planned SCWR fuel qualification test will give the first experience with supercritical water-cooled nuclear systems in general. With a fuel inventory of almost 1 kg of UO{sub 2} with almost 20% enrichment, the supercritical pressure test section inside a low pressure, pool type research reactor needs to be cooled properly even in case of a number of postulated design basis accidents. Depressurization systems and emergency cooling systems will need to be designed with similar reliability as for a prototype reactor to ensure the integrity of barriers retaining the radioactive material. The paper reports about the safety concept and summarizes the safety analyses which have been performed in this context. (author)

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.

Integrated vehicle-based safety systems light-vehicle field operational test key findings report.

Science.gov (United States)

2011-01-01

"This document presents key findings from the light-vehicle field operational test conducted as part of the Integrated Vehicle-Based Safety Systems program. These findings are the result of analyses performed by the University of Michigan Transportat...

The use of probabilistic safety assessment based maintenance indicators to increase the availability of safety related systems in nuclear power plants

International Nuclear Information System (INIS)

Kirchsteiger, C.

1991-04-01

This work describes the theoretical development of a Probabilistic Safety Assessment (PSA) based Performance Indicator (PI) model for a comprehensive Maintenance Efficiency Analysis (MEA) and its practical application to past operational history data of a certain Nuclear Power Plant. Plant specific equipment history and maintenance work order data have been collected and analysed using various advanced statistical procedures (nonparametric methods, multivariate analysis) in order to be able to estimate safety system related equipment and maintenance process trends. The main results of such a MEA case study are the trends in the (in)effectiveness of the performance of a selected safety system and its dominant maintenance related causes of its bad (good) equipment performance. Finally, the therefrom gained results are used to propose a new set of safety system based and maintenance related Performance Indicators, including suggestions for a corresponding plant specific maintenance data collection system. (author)

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

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

Directory of Open Access Journals (Sweden)

Shao Nian

2017-01-01

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

The long-term safety and performance analyses of the surface disposal facility for the Belgian category a waste at Dessel

Energy Technology Data Exchange (ETDEWEB)

Cool, Wim; Vermarien, Elise; Wacquier, William [ONDRAF/NIRAS Avenue des Arts 14, BE-1210 Bruxelles (Belgium); Perko, Janez [SCK-CEN Boeretang 200, BE-2400 Mol (Belgium)

2013-07-01

ONDRAF/NIRAS, the Belgian Agency for Radioactive Waste and Enriched Fissile Materials, and its partners have developed long-term safety and performance analyses in the framework of the license application for a surface disposal facility for low level radioactive waste (category A waste) at Dessel, Belgium. This paper focusses on the methodology of the safety assessments and on key results from the application of this methodology. An overview is given (1) of the performance analyses for the containment safety function of the disposal system and (2) of the radiological impact analyses confirming that radiological impacts are below applicable reference values and constraints and leading to radiological criteria for the waste and the facility. In this discussion, multiple indicators for performance and safety are used to illustrate the multi-faceted nature of long-term performance and safety of the surface disposal. This contributes to the multiple lines of reasoning for confidence building that a positive decision to proceed to the next stage of construction is justified. (authors)

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

Energy Technology Data Exchange (ETDEWEB)

Zuidema, Piet [Nagra, Wettingen (Switzerland)

2015-07-01

In Switzerland, the site selection process according to the ''Sectoral Plan for Deep Geological Repositories'' (BFE 2008) is underway since 2008. This process takes place in three stages. In stage 1 geological siting regions (six for the L/ILW repository and three for the HLW repository) have been identified, in stage 2 sites for the surface facilities have been identified for all siting regions in close co-operation with the sting regions and a narrowing down of the number of siting regions based on geological criteria will take place. In stage 3 the sites for a general license application are selected and the general license applications will be submitted which eventually will lead to the siting decision for both repository types. In the Swiss site selection process, safety has the highest priority. Many factors affect safety and thus a whole range of safety-related issues are considered in the identification and screening of siting possibilities. Besides dose calculations a range of quantitative and qualitative issues are considered. Dose calculations are performed in all three stages of the site selection process. In stage 1 generic safety calculations were made to develop criteria to be used for the identification of potential siting regions. In stage 2, dose calculations are made for comparing the different siting regions according to a procedure prescribed in detail by the regulator. Combined with qualitative evaluations this will lead to a narrowing down of the number of siting regions to at least two siting regions for each repository type. In stage 3 full safety cases will be prepared as part of the documentation for the general license applications. Besides the dose calculations, many other issues related to safety are analyzed in a quantitative and qualitative manner. These consider the 13 criteria defined in the Sectoral Plan and the corresponding indicators. The features analyzed cover the following broad themes: efficiency of

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

In Switzerland, the site selection process according to the ''Sectoral Plan for Deep Geological Repositories'' (BFE 2008) is underway since 2008. This process takes place in three stages. In stage 1 geological siting regions (six for the L/ILW repository and three for the HLW repository) have been identified, in stage 2 sites for the surface facilities have been identified for all siting regions in close co-operation with the sting regions and a narrowing down of the number of siting regions based on geological criteria will take place. In stage 3 the sites for a general license application are selected and the general license applications will be submitted which eventually will lead to the siting decision for both repository types. In the Swiss site selection process, safety has the highest priority. Many factors affect safety and thus a whole range of safety-related issues are considered in the identification and screening of siting possibilities. Besides dose calculations a range of quantitative and qualitative issues are considered. Dose calculations are performed in all three stages of the site selection process. In stage 1 generic safety calculations were made to develop criteria to be used for the identification of potential siting regions. In stage 2, dose calculations are made for comparing the different siting regions according to a procedure prescribed in detail by the regulator. Combined with qualitative evaluations this will lead to a narrowing down of the number of siting regions to at least two siting regions for each repository type. In stage 3 full safety cases will be prepared as part of the documentation for the general license applications. Besides the dose calculations, many other issues related to safety are analyzed in a quantitative and qualitative manner. These consider the 13 criteria defined in the Sectoral Plan and the corresponding indicators. The features analyzed cover the following broad themes: efficiency of

Application of geostatistical methods to long-term safety analyses for radioactive waste repositories

International Nuclear Information System (INIS)

Roehlig, K.J.

2001-01-01

Long-term safety analyses are an important part of the design and optimisation process as well as of the licensing procedure for final repositories for radioactive waste in deep geological formations. For selected scenarios describing possible evolutions of the repository system in the post-closure phase, quantitative consequence analyses are performed. Due to the complexity of the phenomena of concern and the large timeframes under consideration, several types of uncertainties have to be taken into account. The modelling work for the far-field (geosphere) surrounding or overlaying the repository is based on model calculations concerning the groundwater movement and the resulting migration of radionuclides which possibly will be released from the repository. In contrast to engineered systems, the geosphere shows a strong spatial variability of facies, materials and material properties. The paper presented here describes the first steps towards a quantitative approach for an uncertainty assessment taking into account this variability. Due to the availability of a large amount of data and information of several types, the Gorleben site (Germany) has been used for a case study in order to demonstrate the method. (orig.)

Fusion safety data base

International Nuclear Information System (INIS)

Laats, E.T.; Hardy, H.A.

1983-01-01

The purpose of this Fusion Safety Data Base Program is to provide a repository of data for the design and development of safe commercial fusion reactors. The program is sponsored by the United States Department of Energy (DOE), Office of Fusion Energy. The function of the program is to collect, examine, permanently store, and make available the safety data to the entire US magnetic-fusion energy community. The sources of data will include domestic and foreign fusion reactor safety-related research programs. Any participant in the DOE Program may use the Data Base Program from his terminal through user friendly dialog and can view the contents in the form of text, tables, graphs, or system diagrams

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

Ministry for Environment, Nature Conservation and Nuclear Safety (BMU). This expert group, led by the Federal Office for Radiation Protection (BfS), has the task to advise the BMU on all methodological issues for the implementation of probabilistic safety analyses and has elaborated two publications on methods and data for PSA with the aim to support a unified application of the PSA in Germany. With the publication 'Safety requirements for nuclear power plants', a modern version of a German nuclear safety regulations has been published. In this regulation the broad experience of the application of the periodic safety reviews have been incorporated as a key element of regulatory supervision. Further key findings from the European safety review of nuclear power plants were taken into account after the accident at Fukushima. The revision also paid special attention to the requirements and recommendations of WENRA and IAEA. In addition, the recommendations and guidelines of the Nuclear Safety Standards Commission (KTA) and the expert group on Probabilistic Safety Analysis (PSA FAK) have also been updated. The activities of the updates have been focused the natural external hazards 'earthquake' and 'flooding' in the German regulations: - Probabilistic assessment for retrofit measures in individual cases for all operating modes and the PSA level 1 and level 2 is possible. - Deterministic and probabilistic site hazard analysis for the events 'earthquake' and 'flood' are required. - For the event 'earthquake' according to IAEA plants receives a minimum design comparable to 0.1 g >concept. - Furthermore, a seismic instrumentation independent of the location of intensity is required for each installation. - The importance of quality assured plant walk downs to determine the specified plant conditions was explicitly emphasized and required measures to ensure. - Furthermore, the existing methods for their applicability verified the associated generic data base for PSA updated. - The

From extended integrity monitoring to the safety evaluation of satellite-based localisation system

International Nuclear Information System (INIS)

Legrand, Cyril; Beugin, Julie; Marais, Juliette; Conrard, Blaise; El-Koursi, El-Miloudi; Berbineau, Marion

2016-01-01

Global Navigation Satellite Systems (GNSS) such as GPS, already used in aeronautics for safety-related applications, can play a major role in railway safety by allowing a train to locate itself safely. However, in order to implement this positioning solution in any embedded system, its performances must be evaluated according to railway standards. The evaluation of GNSS performances is not based on the same attributes class than RAMS evaluation. Face to these diffculties, we propose to express the integrity attribute, performance of satellite-based localisation. This attribute comes from aeronautical standards and for a hybridised GNSS with inertial system. To achieve this objective, the integrity attribute must be extended to this kind of system and algorithms initially devoted to GNSS integrity monitoring only must be adapted. Thereafter, the formalisation of this integrity attribute permits us to analyse the safety quantitatively through the probabilities of integrity risk and wrong-side failure. In this paper, after an introductory discussion about the use of localisation systems in railway safety context together with integrity issues, a particular integrity monitoring is proposed and described. The detection events of this algorithm permit us to conclude about safety level of satellite-based localisation system.

Safety Psychology Applicating on Coal Mine Safety Management Based on Information System

Science.gov (United States)

Hou, Baoyue; Chen, Fei

In recent years, with the increase of intensity of coal mining, a great number of major accidents happen frequently, the reason mostly due to human factors, but human's unsafely behavior are affected by insecurity mental control. In order to reduce accidents, and to improve safety management, with the help of application security psychology, we analyse the cause of insecurity psychological factors from human perception, from personality development, from motivation incentive, from reward and punishment mechanism, and from security aspects of mental training , and put forward countermeasures to promote coal mine safety production,and to provide information for coal mining to improve the level of safety management.

Fault tree synthesis for software design analysis of PLC based safety-critical systems

International Nuclear Information System (INIS)

Koo, S. R.; Cho, C. H.; Seong, P. H.

2006-01-01

As a software verification and validation should be performed for the development of PLC based safety-critical systems, a software safety analysis is also considered in line with entire software life cycle. In this paper, we propose a technique of software safety analysis in the design phase. Among various software hazard analysis techniques, fault tree analysis is most widely used for the safety analysis of nuclear power plant systems. Fault tree analysis also has the most intuitive notation and makes both qualitative and quantitative analyses possible. To analyze the design phase more effectively, we propose a technique of fault tree synthesis, along with a universal fault tree template for the architecture modules of nuclear software. Consequently, we can analyze the safety of software on the basis of fault tree synthesis. (authors)

Integrated vehicle-based safety systems light-vehicle field operational test, methodology and results report.

Science.gov (United States)

2010-12-01

"This document presents the methodology and results from the light-vehicle field operational test conducted as part of the Integrated Vehicle-Based Safety Systems program. These findings are the result of analyses performed by the University of Michi...

Behavior based safety

International Nuclear Information System (INIS)

Sudhikumaran, T.V.; Mehta, S.C.; Goyal, D.K.

2009-01-01

Behaviour Based Safety (popularly known as BBS) is a new methodology for achieving injury free work place and total Safety Culture. BBS is successfully being implemented and is being practiced as a work methodology for achieving a loss and injury free work environment and work practice. Through BBS, it was brought out that the root causes of all Industrial accidents some how originate from the 'at risk' behaviour of some individual or group of individuals at some level. The policy of NPCIL is to excel in the field of Industrial and Fire Safety in comparison to international standards. This article indents to bring out the various parameters helping in installing BBS programme at any plant. (author)

The use of probabilistic safety assessment (PSA) based maintenance indicators to increase the availability of safety related systems in nuclear power plants

International Nuclear Information System (INIS)

Kirchsteiger, C.

1991-04-01

This work describes the theoretical development of a Probabilistic Safety Assessment (PSA) based Performance Indicator (PI) model for a comprehensive Maintenance Efficiency Analysis (MEA) and its practical application to past operational history data of a certain nuclear power plant. Plant specific equipment history and maintenance work on data have been collected and analysed using various advanced statistical procedures (nonparametric methods, multivariate analysis in order to be able to estimate safety system related equipment and maintenance process trends. The main results of such a MEA case study are the trends in the (in)effectiveness of the performance of a selected safety system and its dominant components as well as the detection of the dominant maintenance related causes of its bad (good) equipment performance. Finally, the therefrom gained results are used to propose a new set of safety system-based and maintenance-related performance indicators, including suggestions for a corresponding plant specific maintenance data collection system. (author)

Dry critical experiments and analyses performed in support of the Topaz-2 Safety Program

International Nuclear Information System (INIS)

Pelowitz, D.B.; Sapir, J.; Glushkov, E.S.; Ponomarev-Stepnoi, N.N.; Bubelev, V.G.; Kompanietz, G.B.; Krutov, A.M.; Polyakov, D.N.; Loynstev, V.A.

1994-01-01

In December 1991, the Strategic Defense Initiative Organization decided to investigate the possibility of launching a Russian Topaz-2 space nuclear power system. Functional safety requirements developed for the Topaz mission mandated that the reactor remain subcritical when flooded and immersed in water. Initial experiments and analyses performed in Russia and the United States indicated that the reactor could potentially become supercritical in several water- or sand-immersion scenarios. Consequently, a series of critical experiments was performed on the Narciss M-II facility at the Kurchatov Institute to measure the reactivity effects of water and sand immersion, to quantify the effectiveness of reactor modifications proposed to preclude criticality, and to benchmark the calculational methods and nuclear data used in the Topaz-2 safety analyses. In this paper we describe the Narciss M-II experimental configurations along with the associated calculational models and methods. We also present and compare the measured and calculated results for the dry experimental configurations

Dry critical experiments and analyses performed in support of the TOPAZ-2 safety program

International Nuclear Information System (INIS)

Pelowitz, D.B.; Sapir, J.; Glushkov, E.S.; Ponomarev-Stepnoi, N.N.; Bubelev, V.G.; Kompanietz, G.B.; Krutov, A.M.; Polyakov, D.N.; Lobynstev, V.A.

1995-01-01

In December 1991, the Strategic Defense Initiative Organization decided to investigate the possibility of launching a Russian Topaz-2 space nuclear power system. Functional safety requirements developed for the Topaz mission mandated that the reactor remain subcritical when flooded and immersed in water. Initial experiments and analyses performed in Russia and the United States indicated that the reactor could potentially become supercritical in several water- or sand-immersion scenarios. Consequently, a series of critical experiments was performed on the Narciss M-II facility at the Kurchatov Institute to measure the reactivity effects of water and sand immersion, to quantify the effectiveness of reactor modifications proposed to preclude criticality, and to benchmark the calculational methods and nuclear data used in the Topaz-2 safety analyses. In this paper we describe the Narciss M-II experimental configurations along with the associated calculational models and methods. We also present and compare the measured and calculated results for the dry experimental configurations. copyright 1995 American Institute of Physics

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

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

International Nuclear Information System (INIS)

2004-01-01

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

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

International Nuclear Information System (INIS)

2005-01-01

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

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

International Nuclear Information System (INIS)

2000-01-01

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

Licensing process for safety-critical software-based systems

Energy Technology Data Exchange (ETDEWEB)

Haapanen, P. [VTT Automation, Espoo (Finland); Korhonen, J. [VTT Electronics, Espoo (Finland); Pulkkinen, U. [VTT Automation, Espoo (Finland)

2000-12-01

System vendors nowadays propose software-based technology even for the most critical safety functions in nuclear power plants. Due to the nature of software faults and the way they cause system failures new methods are needed for the safety and reliability evaluation of these systems. In the research project 'Programmable automation systems in nuclear power plants (OHA)', financed together by the Radiation and Nuclear Safety Authority (STUK), the Ministry of Trade and Industry (KTM) and the Technical Research Centre of Finland (VTT), various safety assessment methods and tools for software based systems are developed and evaluated. As a part of the OHA-work a reference model for the licensing process for software-based safety automation systems is defined. The licensing process is defined as the set of interrelated activities whose purpose is to produce and assess evidence concerning the safety and reliability of the system/application to be licensed and to make the decision about the granting the construction and operation permissions based on this evidence. The parties of the licensing process are the authority, the licensee (the utility company), system vendors and their subcontractors and possible external independent assessors. The responsibility about the production of the evidence in first place lies at the licensee who in most cases rests heavily on the vendor expertise. The evaluation and gauging of the evidence is carried out by the authority (possibly using external experts), who also can acquire additional evidence by using their own (independent) methods and tools. Central issue in the licensing process is to combine the quality evidence about the system development process with the information acquired through tests, analyses and operational experience. The purpose of the licensing process described in this report is to act as a reference model both for the authority and the licensee when planning the licensing of individual applications

Licensing process for safety-critical software-based systems

International Nuclear Information System (INIS)

Haapanen, P.; Korhonen, J.; Pulkkinen, U.

2000-12-01

System vendors nowadays propose software-based technology even for the most critical safety functions in nuclear power plants. Due to the nature of software faults and the way they cause system failures new methods are needed for the safety and reliability evaluation of these systems. In the research project 'Programmable automation systems in nuclear power plants (OHA)', financed together by the Radiation and Nuclear Safety Authority (STUK), the Ministry of Trade and Industry (KTM) and the Technical Research Centre of Finland (VTT), various safety assessment methods and tools for software based systems are developed and evaluated. As a part of the OHA-work a reference model for the licensing process for software-based safety automation systems is defined. The licensing process is defined as the set of interrelated activities whose purpose is to produce and assess evidence concerning the safety and reliability of the system/application to be licensed and to make the decision about the granting the construction and operation permissions based on this evidence. The parties of the licensing process are the authority, the licensee (the utility company), system vendors and their subcontractors and possible external independent assessors. The responsibility about the production of the evidence in first place lies at the licensee who in most cases rests heavily on the vendor expertise. The evaluation and gauging of the evidence is carried out by the authority (possibly using external experts), who also can acquire additional evidence by using their own (independent) methods and tools. Central issue in the licensing process is to combine the quality evidence about the system development process with the information acquired through tests, analyses and operational experience. The purpose of the licensing process described in this report is to act as a reference model both for the authority and the licensee when planning the licensing of individual applications. Many of the

Reliability-based approaches for safety margin assessment in the French nuclear industry

International Nuclear Information System (INIS)

Ardillon, E.; Barthelet, B.; Meister, E.; Cambefort, P.; Hornet, P.; Le Delliou, P.

2003-01-01

The prevention of the fast fracture damage of the mechanical equipment important for the safety of nuclear islands of the French PWR relies on deterministic rules. These rules include flaw acceptance criteria involving safety factors applied to characteristic values (implicit margins) of the physical variables. The sets of safety factors that are currently under application in the industrial analyses with the agreement of the Safety Authority, are distributed across the two main physical parameters and have partly been based on a semi-probabilistic approach. After presenting the generic probabilistic pro-codification approach this paper shows its application to the evaluation of the performances of the existing regulatory flaw acceptance criteria. This application can be carried out in a realistic manner or in a more simplified one. These two approaches are applied to representative mechanical components. Their results are consistent. (author)

Reactivity initiated accident analyses for the safety assessment of upgraded JRR-3

International Nuclear Information System (INIS)

Harami, Taikan; Uemura, Mutsumi; Ohnishi, Nobuaki

1984-08-01

JRR-3, currently a heavy water moderated and cooled 10 MW reactor, is to be upgraded to a light water moderated and cooled, heavy water reflected 20 MW reactor. This report describes the analytical results of reactivity initiated accidents for the safety assessment of upgraded JRR-3. The following five cases have been selected for the assessment; (1) uncontrolled control rod withdrawal from zero power, (2) uncontrolled control rod withdrawal from full power, (3) removal of irradiation samples, (4) increase of primary coolant flow, (5) failure of heavy water tank. Parameter studies have been made for each of the above cases to cover possible uncertainties. All analyses have been made by a computer code EUREKA-2. The results show that the safety criteria for upgraded JRR-3 are all met and the adequacy of the design is confirmed. (author)

Methodology development for statistical evaluation of reactor safety analyses

International Nuclear Information System (INIS)

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

1976-07-01

In February 1975, Westinghouse Electric Corporation, under contract to Electric Power Research Institute, started a one-year program to develop methodology for statistical evaluation of nuclear-safety-related engineering analyses. The objectives of the program were to develop an understanding of the relative efficiencies of various computational methods which can be used to compute probability distributions of output variables due to input parameter uncertainties in analyses of design basis events for nuclear reactors and to develop methods for obtaining reasonably accurate estimates of these probability distributions at an economically feasible level. A series of tasks was set up to accomplish these objectives. Two of the tasks were to investigate the relative efficiencies and accuracies of various Monte Carlo and analytical techniques for obtaining such estimates for a simple thermal-hydraulic problem whose output variable of interest is given in a closed-form relationship of the input variables and to repeat the above study on a thermal-hydraulic problem in which the relationship between the predicted variable and the inputs is described by a short-running computer program. The purpose of the report presented is to document the results of the investigations completed under these tasks, giving the rationale for choices of techniques and problems, and to present interim conclusions

Intelligent monitoring-based safety system of massage robot

Institute of Scientific and Technical Information of China (English)

胡宁; 李长胜; 王利峰; 胡磊; 徐晓军; 邹雲鹏; 胡玥; 沈晨

2016-01-01

As an important attribute of robots, safety is involved in each link of the full life cycle of robots, including the design, manufacturing, operation and maintenance. The present study on robot safety is a systematic project. Traditionally, robot safety is defined as follows: robots should not collide with humans, or robots should not harm humans when they collide. Based on this definition of robot safety, researchers have proposed ex ante and ex post safety standards and safety strategies and used the risk index and risk level as the evaluation indexes for safety methods. A massage robot realizes its massage therapy function through applying a rhythmic force on the massage object. Therefore, the traditional definition of safety, safety strategies, and safety realization methods cannot satisfy the function and safety requirements of massage robots. Based on the descriptions of the environment of massage robots and the tasks of massage robots, the present study analyzes the safety requirements of massage robots; analyzes the potential safety dangers of massage robots using the fault tree tool; proposes an error monitoring-based intelligent safety system for massage robots through monitoring and evaluating potential safety danger states, as well as decision making based on potential safety danger states; and verifies the feasibility of the intelligent safety system through an experiment.

Review of Ontario Hydro Pickering 'A' and Bruce 'A' nuclear generating stations' accident analyses

International Nuclear Information System (INIS)

Serdula, K.J.

1988-01-01

Deterministic safety analysis for the Pickering 'A' and Bruce 'A' nuclear generating stations were reviewed. The methodology used in the evaluation and assessment was based on the concept of 'N' critical parameters defining an N-dimensional safety parameter space. The reviewed accident analyses were evaluated and assessed based on their demonstrated safety coverage for credible values and trajectories of the critical parameters within this N-dimensional safety parameter space. The reported assessment did not consider probability of occurrence of event. The reviewed analyses were extensive for potential occurrence of accidents under normal steady-state operating conditions. These analyses demonstrated an adequate assurance of safety for the analyzed conditions. However, even for these reactor conditions, items have been identified for consideration of review and/or further study, which would provide a greater assurance of safety in the event of an accident. Accident analyses based on a plant in a normal transient operating state or in an off-normal condition but within the allowable operating envelope are not as extensive. Improvements in demonstrations and/or justifications of safety upon potential occurrence of accidents would provide further assurance of adequacy of safety under these conditions. Some events under these conditions have not been analyzed because of their judged low probability; however, accident analyses in this area should be considered. Recommendations are presented relating to these items; it is also recommended that further study is needed of the Pickering 'A' special safety systems

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.

IMPROVING CONTROL ROOM DESIGN AND OPERATIONS BASED ON HUMAN FACTORS ANALYSES OR HOW MUCH HUMAN FACTORS UPGRADE IS ENOUGH ?

Energy Technology Data Exchange (ETDEWEB)

HIGGINS,J.C.; OHARA,J.M.; ALMEIDA,P.

2002-09-19

THE JOSE CABRERA NUCLEAR POWER PLANT IS A ONE LOOP WESTINGHOUSE PRESSURIZED WATER REACTOR. IN THE CONTROL ROOM, THE DISPLAYS AND CONTROLS USED BY OPERATORS FOR THE EMERGENCY OPERATING PROCEDURES ARE DISTRIBUTED ON FRONT AND BACK PANELS. THIS CONFIGURATION CONTRIBUTED TO RISK IN THE PROBABILISTIC SAFETY ASSESSMENT WHERE IMPORTANT OPERATOR ACTIONS ARE REQUIRED. THIS STUDY WAS UNDERTAKEN TO EVALUATE THE IMPACT OF THE DESIGN ON CREW PERFORMANCE AND PLANT SAFETY AND TO DEVELOP DESIGN IMPROVEMENTS.FIVE POTENTIAL EFFECTS WERE IDENTIFIED. THEN NUREG-0711 [1], PROGRAMMATIC, HUMAN FACTORS, ANALYSES WERE CONDUCTED TO SYSTEMATICALLY EVALUATE THE CR-LA YOUT TO DETERMINE IF THERE WAS EVIDENCE OF THE POTENTIAL EFFECTS. THESE ANALYSES INCLUDED OPERATING EXPERIENCE REVIEW, PSA REVIEW, TASK ANALYSES, AND WALKTHROUGH SIMULATIONS. BASED ON THE RESULTS OF THESE ANALYSES, A VARIETY OF CONTROL ROOM MODIFICATIONS WERE IDENTIFIED. FROM THE ALTERNATIVES, A SELECTION WAS MADE THAT PROVIDED A REASONABLEBALANCE BE TWEEN PERFORMANCE, RISK AND ECONOMICS, AND MODIFICATIONS WERE MADE TO THE PLANT.

Uncertainty and conservatism in safety evaluations based on a BEPU approach

International Nuclear Information System (INIS)

Yamaguchi, A.; Mizokami, S.; Kudo, Y.; Hotta, A.

2009-01-01

Atomic Energy Society of Japan has published 'Standard Method for Safety Evaluation using Best Estimate Code Based on Uncertainty and Scaling Analyses with Statistical Approach' to be applied to accidents and AOOs in the safety evaluation of LWRs. In this method, hereafter named as the AESJ-SSE (Statistical Safety Evaluation) method, identification and quantification of uncertainties will be performed and then a combination of the best estimate code and the evaluation of uncertainty propagation will be performed. Uncertainties are categorized into bias and variability. In general, bias is related to our state-of-knowledge on uncertainty objects (modeling, scaling, input data, etc.) while variability reflects stochastic features involved in these objects. Considering many kinds of uncertainties in thermal-hydraulics models and experimental databases show variabilities that will be strongly influenced by our state of knowledge, it seems reasonable that these variabilities are also related to state-of-knowledge. The design basis events (DBEs) that are employed for licensing analyses form a main part of the given or prior conservatism. The regulatory acceptance criterion is also regarded as the prior conservatism. In addition to these prior conservatisms, a certain amount of the posterior conservatism is added with maintaining intimate relationships with state-of-knowledge. In the AESJ-SSE method, this posterior conservatism can be incorporated into the safety evaluation in a combination of the following three ways, (1) broadening ranges of variability relevant to uncertainty objects, (2) employing more disadvantageous biases relevant to uncertainty objects and (3) adding an extra bias to the safety evaluation results. Knowing implemented quantitative bases of uncertainties and conservatism, the AESJ-SSE method provides a useful ground for rational decision-making. In order to seek for 'the best estimation' as well as reasonably setting the analytical margin, a degree

Job safety and awareness analysis of safety implementation among electrical workers in airport service company

Directory of Open Access Journals (Sweden)

Putra Perdana Suteja

2018-01-01

Full Text Available Electrical is a fundamental process in the company that has high risk and responsibility especially in public service company such as an airport. Hence, the company that operates activities in the airport has to identify and control the safety activities of workers. On the safety implementation, the lack of workers’ awareness is fundamental aspects to the safety failure. Therefore, this study aimed to analyse the safety awareness and identify risk in the electrical workplace. Safety awareness questionnaires are distributed to ten workers in order to analyse their awareness. Job safety analysis method used to identify the risk in the electrical workplace. The preliminary study stated that workers were not aware of personal protective equipment usage so that the awareness and behavioural need to be analysed. The result is the hazard was found such as electrical shock and noise for various intensity in the workplace. While electrical workers were aware of safety implementation but less of safety behaviour. Furthermore, the recommendation can be implemented are the implementation of behaviour-based safety (BBS, 5S implementation and accident report list.

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)

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

Health economics and outcomes methods in risk-based decision-making for blood safety.

Science.gov (United States)

Custer, Brian; Janssen, Mart P

2015-08-01

Analytical methods appropriate for health economic assessments of transfusion safety interventions have not previously been described in ways that facilitate their use. Within the context of risk-based decision-making (RBDM), health economics can be important for optimizing decisions among competing interventions. The objective of this review is to address key considerations and limitations of current methods as they apply to blood safety. Because a voluntary blood supply is an example of a public good, analyses should be conducted from the societal perspective when possible. Two primary study designs are recommended for most blood safety intervention assessments: budget impact analysis (BIA), which measures the cost to implement an intervention both to the blood operator but also in a broader context, and cost-utility analysis (CUA), which measures the ratio between costs and health gain achieved, in terms of reduced morbidity and mortality, by use of an intervention. These analyses often have important limitations because data that reflect specific aspects, for example, blood recipient population characteristics or complication rates, are not available. Sensitivity analyses play an important role. The impact of various uncertain factors can be studied conjointly in probabilistic sensitivity analyses. The use of BIA and CUA together provides a comprehensive assessment of the costs and benefits from implementing (or not) specific interventions. RBDM is multifaceted and impacts a broad spectrum of stakeholders. Gathering and analyzing health economic evidence as part of the RBDM process enhances the quality, completeness, and transparency of decision-making. © 2015 AABB.

Computer-based and web-based radiation safety training

Energy Technology Data Exchange (ETDEWEB)

Owen, C., LLNL

1998-03-01

The traditional approach to delivering radiation safety training has been to provide a stand-up lecture of the topic, with the possible aid of video, and to repeat the same material periodically. New approaches to meeting training requirements are needed to address the advent of flexible work hours and telecommuting, and to better accommodate individuals learning at their own pace. Computer- based and web-based radiation safety training can provide this alternative. Computer-based and web- based training is an interactive form of learning that the student controls, resulting in enhanced and focused learning at a time most often chosen by the student.

Safety analyses for sodium-cooled fast reactors with pelletized and sphere-pac oxide fuels within the FP-7 European project PELGRIMM - 15386

International Nuclear Information System (INIS)

Maschek, W.; Andriolo, L.; Matzerath-Boccaccini, C.; Delage, F.; Parisi, C.; Del Nevo, A.; Abbate, G.; Schmitt, D.

2015-01-01

The European FP-7 project PELGRIMM addresses the development of Minor-Actinide (MA) bearing oxide fuel for Sodium-cooled Fast Reactors. Optionally, both MA homogeneous recycling and heterogeneous recycling is investigated with pellet and sphere-pac fuel. A first safety assessment of sphere-pac fuelled cores should be given in the Work Package 4 of the project. This assessment is in continuity with the former FP-7 CP-ESFR project. Within the CP-ESFR project the CONF2 core design has been developed characterized by a core with a large upper sodium plenum to reduce the coolant void worth. This optimized core has been chosen for the safety analyses in PELGRIMM. The task within the PELGRIMM project is thus a safety assessment of the CONF2 core loaded either with pellets or with sphere-pac fuel. The investigations started with the design of the CONF2 core with sphere-pac fuel and the determination of core safety parameters and burn-up behavior. The neutronic analyses have been performed with the MCNPX code. Variants of the CONF2 core contain up to 4% Am in the fuel. The results revealed an extended void worth (core + upper plenum) for an Am free core of 1 up to 3 dollars for the 4% Am core. Thermal-hydraulic design analyses have been performed by RELAP5-3D. The accident simulations should be performed by different codes, some of which focus on the initiation phase of the accident, as SAS4A, BELLA and the MAT5DYN code, whereas the SIMMER-III code will also deal with the later accident phases and a potential whole core melting. The codes had to be adapted to the specifics of the sphere-pac fuel, in particular to the thermal conductivity and gap conditions. Analyses showed that the safety assessment has to take into account two main phases. Starting up the core, the green fuel shows a reduced fuel thermal conductivity. After restructuring within a couple of hours, the thermal conductivity recovers and the fuel temperature decreases. The main objective of the safety analyses

IT-CARES: an interactive tool for case-crossover analyses of electronic medical records for patient safety.

Science.gov (United States)

Caron, Alexandre; Chazard, Emmanuel; Muller, Joris; Perichon, Renaud; Ferret, Laurie; Koutkias, Vassilis; Beuscart, Régis; Beuscart, Jean-Baptiste; Ficheur, Grégoire

2017-03-01

The significant risk of adverse events following medical procedures supports a clinical epidemiological approach based on the analyses of collections of electronic medical records. Data analytical tools might help clinical epidemiologists develop more appropriate case-crossover designs for monitoring patient safety. To develop and assess the methodological quality of an interactive tool for use by clinical epidemiologists to systematically design case-crossover analyses of large electronic medical records databases. We developed IT-CARES, an analytical tool implementing case-crossover design, to explore the association between exposures and outcomes. The exposures and outcomes are defined by clinical epidemiologists via lists of codes entered via a user interface screen. We tested IT-CARES on data from the French national inpatient stay database, which documents diagnoses and medical procedures for 170 million inpatient stays between 2007 and 2013. We compared the results of our analysis with reference data from the literature on thromboembolic risk after delivery and bleeding risk after total hip replacement. IT-CARES provides a user interface with 3 columns: (i) the outcome criteria in the left-hand column, (ii) the exposure criteria in the right-hand column, and (iii) the estimated risk (odds ratios, presented in both graphical and tabular formats) in the middle column. The estimated odds ratios were consistent with the reference literature data. IT-CARES may enhance patient safety by facilitating clinical epidemiological studies of adverse events following medical procedures. The tool's usability must be evaluated and improved in further research. © The Author 2016. Published by Oxford University Press on behalf of the American Medical Informatics Association.

Safety management of software-based equipment

CERN Document Server

Boulanger, Jean-Louis

2013-01-01

A review of the principles of the safety of software-based equipment, this book begins by presenting the definition principles of safety objectives. It then moves on to show how it is possible to define a safety architecture (including redundancy, diversification, error-detection techniques) on the basis of safety objectives and how to identify objectives related to software programs. From software objectives, the authors present the different safety techniques (fault detection, redundancy and quality control). "Certifiable system" aspects are taken into account throughout the book. C

A Technique of Software Safety Analysis in the Design Phase for PLC Based Safety-Critical Systems

International Nuclear Information System (INIS)

Koo, Seo-Ryong; Kim, Chang-Hwoi

2017-01-01

The purpose of safety analysis, which is a method of identifying portions of a system that have the potential for unacceptable hazards, is firstly to encourage design changes that will reduce or eliminate hazards and, secondly, to conduct special analyses and tests that can provide increased confidence in especially vulnerable portions of the system. For the design and implementation phase of the PLC based systems, we proposed a technique for software design specification and analysis, and this technique enables us to generate software design specifications (SDSs) in nuclear fields. For the safety analysis in the design phase, we used architecture design blocks of NuFDS to represent the architecture of the software. On the basis of the architecture design specification, we can directly generate the fault tree and then use the fault tree for qualitative analysis. Therefore, we proposed a technique of fault tree synthesis, along with a universal fault tree template for the architecture modules of nuclear software. Through our proposed fault tree synthesis in this work, users can use the architecture specification of the NuFDS approach to intuitively compose fault trees that help analyze the safety design features of software.

Qualification of FPGA-Based Safety-Related PRM System

International Nuclear Information System (INIS)

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

2011-01-01

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

The implementation of a burnup credit based criticality safety assessment in the THORP head end plant

International Nuclear Information System (INIS)

Gulliford, J.; Edge, J.A.; Gracey, J.; Harris, N.

2003-01-01

A new criticality safety assessment based on Actinide-Only Burnup Credit has been developed to cover operations in BNFL's Thermal Oxide Reprocessing Plant (THORP). Reduction of the gadolinium concentration leads to significant reduction in active waste volumes. Detailed description of the methodology was presented at ICNC 1999 and the basic components of the approved safety case have remained unchanged from those proposed then. This paper presents a brief summary of the new methodology, and describes further analyses carried out to quantify additional safety margins. These additional margins are not credited in the derivation of the operating limits, but provide further evidence of the fault tolerance inherent in the new regime. As part of the arrangements to monitor the overall performance of the plant and instrumentation under the new regime, various analyses of plant data are made, including 'on-line' cross checks of measured versus expected fuel parameters (i.e. in addition to the checks on Residual Enrichment). Statistical analyses of data are made and compared with similar data from earlier batches. A summary of analyses made on some of the early fuel batches is presented here. A summary of the likely further development in the Burnup Credit methodology is given in this paper. (author)

Estimation of effective block conductivities based on discrete network analyses using data from the Aespoe site

International Nuclear Information System (INIS)

La Pointe, P.R.; Wallmann, P.; Follin, S.

1995-09-01

Numerical continuum codes may be used for assessing the role of regional groundwater flow in far-field safety analyses of a nuclear waste repository at depth. The focus of this project is to develop and evaluate one method based on Discrete Fracture Network (DFN) models to estimate block-scale permeability values for continuum codes. Data from the Aespoe HRL and surrounding area are used. 57 refs, 76 figs, 15 tabs

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

Jefferson Lab IEC 61508/61511 Safety PLC Based Safety System

International Nuclear Information System (INIS)

Mahoney, Kelly; Robertson, Henry

2009-01-01

This paper describes the design of the new 12 GeV Upgrade Personnel Safety System (PSS) at the Thomas Jefferson National Accelerator Facility (TJNAF). The new PSS design is based on the implementation of systems designed to meet international standards IEC61508 and IEC 61511 for programmable safety systems. In order to meet the IEC standards, TJNAF engineers evaluated several SIL 3 Safety PLCs before deciding on an optimal architecture. In addition to hardware considerations, software quality standards and practices must also be considered. Finally, we will discuss R and D that may lead to both high safety reliability and high machine availability that may be applicable to future accelerators such as the ILC.

Safety prediction for basic components of safety-critical software based on static testing

International Nuclear Information System (INIS)

Son, H.S.; Seong, P.H.

2000-01-01

The purpose of this work is to develop a safety prediction method, with which we can predict the risk of software components based on static testing results at the early development stage. The predictive model combines the major factor with the quality factor for the components, which are calculated based on the measures proposed in this work. The application to a safety-critical software system demonstrates the feasibility of the safety prediction method. (authors)

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

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.

Patient safety issues in office-based surgery and anaesthesia in Switzerland: a qualitative study.

Science.gov (United States)

McLennan, Stuart; Schwappach, David; Harder, Yves; Staender, Sven; Elger, Bernice

2017-08-01

To identify the spectrum of patient safety issues in office-based surgery and anaesthesia in Switzerland. Purposive sample of 23 experts in surgery and anaesthesia and quality and regulation in Switzerland. Data were collected via individual qualitative interviews using a researcher-developed semi-structured interview guide between March 2016 and September 2016. Interviews were transcribed and analysed using conventional content analysis. Issues were categorised under the headings "structure", "process", and "outcome". Experts identified two key overarching patient safety and regulatory issues in relation to office-based surgery and anaesthesia in Switzerland. First, experts repeatedly raised the current lack of data and transparency of the setting. It is unknown how many surgeons are operating in offices, how many and what types of operations are being done, and what the outcomes are. Secondly, experts also noted the limited oversight and regulation of the setting. While some standards exists, most experts felt that more minimal safety standards are needed regarding the requirements that must be met to do office-based surgery and what can and cannot be done in the office-based setting are needed, but they advocated a self-regulatory approach. There is a lack of empirical data regarding the quantity and quality office-based surgery and anaesthesia in Switzerland. Further research is needed to address these research gaps and inform health policy in relation to patient safety in office-based surgery and anaesthesia in Switzerland. Copyright © 2017. Published by Elsevier GmbH.

Active SMS-based influenza vaccine safety surveillance in Australian children.

Science.gov (United States)

Pillsbury, Alexis; Quinn, Helen; Cashman, Patrick; Leeb, Alan; Macartney, Kristine

2017-12-18

Australia's novel, active surveillance system, AusVaxSafety, monitors the post-market safety of vaccines in near real time. We analysed cumulative surveillance data for children aged 6 months to 4 years who received seasonal influenza vaccine in 2015 and/or 2016 to determine: adverse event following immunisation (AEFI) rates by vaccine brand, age and concomitant vaccine administration. Parent/carer reports of AEFI occurring within 3 days of their child receiving an influenza vaccine in sentinel immunisation clinics were solicited by Short Message Service (SMS) and/or email-based survey. Retrospective data from 2 years were combined to examine specific AEFI rates, particularly fever and medical attendance as a proxy for serious adverse events (SAE), with and without concomitant vaccine administration. As trivalent influenza vaccines (TIV) were funded in Australia's National Immunisation Program (NIP) in 2015 and quadrivalent (QIV) in 2016, respectively, we compared their safety profiles. 7402 children were included. Data were reported weekly through each vaccination season; no safety signals or excess of adverse events were detected. More children who received a concomitant vaccine had fever (7.5% versus 2.8%; p vaccine was associated with the highest increase in AEFI rates among children receiving a specified concomitant vaccine: 30.3% reported an AEFI compared with 7.3% who received an influenza vaccine alone (p safety profiles included low and expected AEFI rates (fever: 4.3% for TIV compared with 3.2% for QIV (p = .015); injection site reaction: 1.9% for TIV compared with 3.0% for QIV (p safety profile between brands. Active participant-reported data provided timely vaccine brand-specific safety information. Our surveillance system has particular utility in monitoring the safety of influenza vaccines, given that they may vary in composition annually. Copyright © 2017 Elsevier Ltd. All rights reserved.

Safety prediction for basic components of safety critical software based on static testing

International Nuclear Information System (INIS)

Son, H.S.; Seong, P.H.

2001-01-01

The purpose of this work is to develop a safety prediction method, with which we can predict the risk of software components based on static testing results at the early development stage. The predictive model combines the major factor with the quality factor for the components, both of which are calculated based on the measures proposed in this work. The application to a safety-critical software system demonstrates the feasibility of the safety prediction method. (authors)

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)

Safety based on organisational learning (SOL) - Conceptual approach and verification of a method for event analysis

International Nuclear Information System (INIS)

Miller, R.; Wilpert, B.; Fahlbruch, B.

1999-01-01

This paper discusses a method for analysing safety-relevant events in NPP which is known as 'SOL', safety based on organisational learning. After discussion of the specific organisational and psychological problems examined in the event analysis, the analytic process using the SOL approach is explained as well as the required general setting. The SOL approach has been tested both with scientific experiments and from the practical perspective, by operators of NPPs and experts from other branches of industry. (orig./CB) [de

Field Programmable Gate Array-based I and C Safety System

International Nuclear Information System (INIS)

Kim, Hyun Jeong; Kim, Koh Eun; Kim, Young Geul; Kwon, Jong Soo

2014-01-01

Programmable Logic Controller (PLC)-based I and C safety system used in the operating nuclear power plants has the disadvantages of the Common Cause Failure (CCF), high maintenance costs and quick obsolescence, and then it is necessary to develop the other platform to replace the PLC. The Field Programmable Gate Array (FPGA)-based Instrument and Control (I and C) safety system is safer and more economical than Programmable Logic Controller (PLC)-based I and C safety system. Therefore, in the future, FPGA-based I and C safety system will be able to replace the PLC-based I and C safety system in the operating and the new nuclear power plants to get benefited from its safety and economic advantage. FPGA-based I and C safety system shall be implemented and verified by applying the related requirements to perform the safety function

Field Programmable Gate Array-based I and C Safety System

Energy Technology Data Exchange (ETDEWEB)

Kim, Hyun Jeong; Kim, Koh Eun; Kim, Young Geul; Kwon, Jong Soo [KEPCO, Daejeon (Korea, Republic of)

2014-08-15

Programmable Logic Controller (PLC)-based I and C safety system used in the operating nuclear power plants has the disadvantages of the Common Cause Failure (CCF), high maintenance costs and quick obsolescence, and then it is necessary to develop the other platform to replace the PLC. The Field Programmable Gate Array (FPGA)-based Instrument and Control (I and C) safety system is safer and more economical than Programmable Logic Controller (PLC)-based I and C safety system. Therefore, in the future, FPGA-based I and C safety system will be able to replace the PLC-based I and C safety system in the operating and the new nuclear power plants to get benefited from its safety and economic advantage. FPGA-based I and C safety system shall be implemented and verified by applying the related requirements to perform the safety function.

Precursor analyses - The use of deterministic and PSA based methods in the event investigation process at nuclear power plants

International Nuclear Information System (INIS)

2004-09-01

The efficient feedback of operating experience (OE) is a valuable source of information for improving the safety and reliability of nuclear power plants (NPPs). It is therefore essential to collect information on abnormal events from both internal and external sources. Internal operating experience is analysed to obtain a complete understanding of an event and of its safety implications. Corrective or improvement measures may then be developed, prioritized and implemented in the plant if considered appropriate. Information from external events may also be analysed in order to learn lessons from others' experience and prevent similar occurrences at our own plant. The traditional ways of investigating operational events have been predominantly qualitative. In recent years, a PSA-based method called probabilistic precursor event analysis has been developed, used and applied on a significant scale in many places for a number of plants. The method enables a quantitative estimation of the safety significance of operational events to be incorporated. The purpose of this report is to outline a synergistic process that makes more effective use of operating experience event information by combining the insights and knowledge gained from both approaches, traditional deterministic event investigation and PSA-based event analysis. The PSA-based view on operational events and PSA-based event analysis can support the process of operational event analysis at the following stages of the operational event investigation: (1) Initial screening stage. (It introduces an element of quantitative analysis into the selection process. Quantitative analysis of the safety significance of nuclear plant events can be a very useful measure when it comes to selecting internal and external operating experience information for its relevance.) (2) In-depth analysis. (PSA based event evaluation provides a quantitative measure for judging the significance of operational events, contributors to

Developing design premises for a KBS-3V repository based on results from the safety assessment - 16027

International Nuclear Information System (INIS)

Andersson, Johan; Hedin, Allan

2009-01-01

As a part of the planned license application for a final repository for spent nuclear fuel the Swedish Nuclear Fuel and Waste Management Co. (SKB), has developed design premises from a long term safety aspect of a KBS-3V repository for spent nuclear fuel. The purpose is to provide requirements from a long term safety aspect, to form the basis for the development of the reference design of the repository and to justify that design. Design premises typically concern specification on what mechanical loads the barriers must withstand, restrictions on the composition of barrier materials or acceptance criteria for the various underground excavations. These design constraints, if all fulfilled by the actual design, should form a good basis for demonstrating repository safety. The justification for these design premises is derived from SKB's most recent safety assessment SR-Can complemented by a few additional analyses. Some of the design premises may be modified in future stages of SKB's program, as a result of analyses based on more detailed site data and a more developed understanding of processes of importance for long-term safety. (authors)

Implementation of an indicator-based safety management system for the EnKK NPP's

International Nuclear Information System (INIS)

Bassing, G.; Nasellu, M.; Ritter, J.

2004-01-01

This presentation at the Eurosafe Berlin 2004 will give an overview of the activities on safety management of EnBW Kernkraft GmbH taken up with a notifiable event in August 2001 at KKP 2 nuclear site. After this event EnBW announced the development and introduction of an indicator-based safety management system (SMS) at all sites of its nuclear power plants. A SMS team with members from all NPP sites was built which had to analyse all processes based on the DIN EN ISO 9000 philosophy and to control them by indicators. The regulatory authorities and their experts would accompany this process in a suitable fashion and monitor and review it after its introduction. This presentation shows the process during the development of the system, the status of its introduction and the general involvement of the regulator. (orig.)

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_i_n_o_r_g, 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

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

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

Sorption data bases for argillaceous rocks and bentonite for the provisional safety analyses for SGT-E2

International Nuclear Information System (INIS)

Baeyens, B.; Thoenen, T.; Bradbury, M. H.; Marques Fernandes, M.

2014-11-01

In Stage 1 of the Sectoral Plan for Deep Geological Repositories, four rock types have been identified as being suitable host rocks for a radioactive waste repository, namely, Opalinus Clay for a high-level (HLW) and a low- and intermediate-level (L/ILW) repository, and 'Brauner Dogger', Effingen Member and Helvetic Marls for a L/ILW repository. Sorption data bases (SDBs) for all of these host rocks are required for the provisional safety analyses, including all of the bounding porewater and mineralogical composition combinations. In addition, SDBs are needed for the rock formations lying below Opalinus Clay (lower confining units) and for the bentonite backfill in the HLW repository. In some previous work Bradbury et al. (2010) have described a methodology for developing sorption data bases for argillaceous rocks and compacted bentonite. The main factors influencing the sorption in such systems are the phyllosilicate mineral content, particular the 2:1 clay mineral content (illite/smectite/illite-smectite mixed layers) and the water chemistry which determines the radionuclide species in the aqueous phase. The source sorption data were taken predominantly from measurements on illite (or montmorillonite in the case of bentonite) and converted to the defined conditions in each system considered using a series of so called conversion factors to take into account differences in mineralogy, in pH and in radionuclide speciation. Finally, a Lab → Field conversion factor was applied to adapt sorption data measured in dispersed systems (batch experiments) to intact rock under in-situ conditions. This methodology to develop sorption data bases has been applied to the selected host rocks, lower confining units and compacted bentonite taking into account the mineralogical and porewater composition ranges defined. Confidence in the validity and correctness of this methodology has been built up through additional studies: (i) sorption values obtained in the manner

Sorption data bases for argillaceous rocks and bentonite for the provisional safety analyses for SGT-E2

Energy Technology Data Exchange (ETDEWEB)

Baeyens, B.; Thoenen, T.; Bradbury, M. H.; Marques Fernandes, M.

2014-11-15

In Stage 1 of the Sectoral Plan for Deep Geological Repositories, four rock types have been identified as being suitable host rocks for a radioactive waste repository, namely, Opalinus Clay for a high-level (HLW) and a low- and intermediate-level (L/ILW) repository, and 'Brauner Dogger', Effingen Member and Helvetic Marls for a L/ILW repository. Sorption data bases (SDBs) for all of these host rocks are required for the provisional safety analyses, including all of the bounding porewater and mineralogical composition combinations. In addition, SDBs are needed for the rock formations lying below Opalinus Clay (lower confining units) and for the bentonite backfill in the HLW repository. In some previous work Bradbury et al. (2010) have described a methodology for developing sorption data bases for argillaceous rocks and compacted bentonite. The main factors influencing the sorption in such systems are the phyllosilicate mineral content, particular the 2:1 clay mineral content (illite/smectite/illite-smectite mixed layers) and the water chemistry which determines the radionuclide species in the aqueous phase. The source sorption data were taken predominantly from measurements on illite (or montmorillonite in the case of bentonite) and converted to the defined conditions in each system considered using a series of so called conversion factors to take into account differences in mineralogy, in pH and in radionuclide speciation. Finally, a Lab → Field conversion factor was applied to adapt sorption data measured in dispersed systems (batch experiments) to intact rock under in-situ conditions. This methodology to develop sorption data bases has been applied to the selected host rocks, lower confining units and compacted bentonite taking into account the mineralogical and porewater composition ranges defined. Confidence in the validity and correctness of this methodology has been built up through additional studies: (i) sorption values obtained in the manner

Kowledge-based dynamic network safety calculations. Wissensbasierte dynamische Netzsicherheitsberechnungen

Energy Technology Data Exchange (ETDEWEB)

Kulicke, B [Inst. fuer Hochspannungstechnik und Starkstromanlagen, Berlin (Germany); Schlegel, S [Inst. fuer Hochspannungstechnik und Starkstromanlagen, Berlin (Germany)

1993-06-28

An important part of network operation management is the estimation and maintenance of the security of supply. So far the control personnel has only been supported by static network analyses and safety calculations. The authors describe an expert system, which is coupled to a real time simulation program on a transputer basis, for dynamic network safety calculations. They also introduce the system concept and the most important functions of the expert system. (orig.)

Experiences in assessing safety culture

International Nuclear Information System (INIS)

Spitalnik, J.

2002-01-01

Based on several Safety Culture self-assessment applications in nuclear organisations, the paper stresses relevant aspects to be considered when programming an assessment of this type. Reasons for assessing Safety Culture, basic principles to take into account, necessary resources, the importance of proper statistical analyses, the feed-back of results, and the setting up of action plans to enhance Safety Culture are discussed. (author)

Safety case plan 2008

International Nuclear Information System (INIS)

2008-07-01

Following the guidelines set forth by the Ministry of Trade and Industry (now Ministry of Employment and Economy) Posiva is preparing to submit the construction license application for a spent fuel repository by the end of the year 2012. The long-term safety section supporting the license application is based on a safety case, which, according to the internationally adopted definition, is a compilation of the evidence, analyses and arguments that quantify and substantiate the safety and the level of expert confidence in the safety of the planned repository. In 2005, Posiva presented a plan to prepare such a safety case. The present report provides a revised plan of the safety case contents mentioned above. The update of the safety case plan takes into account the recommendations made by the Radiation and Nuclear Safety Authority (STUK) about improving the focus and further developing the plan. Accordingly, particular attention is given to the quality management of the safety case work, the management of uncertainties and the scenario methodology. The quality management is based on the ISO 9001:2000 standard process thinking enhanced with special features arising from STUK's YVL Guides. The safety case production process is divided into four main sub-processes. The conceptualisation and methodology sub-process defines the framework for the assessment. The critical data handling and modelling sub-process links Posiva's main technical and scientific activities to the production of the safety case. The assessment sub-process analyses the consequences of the evolution of the disposal system in various scenarios, classified either as part of the expected evolution or as disruptive scenarios. The compliance and confidence sub-process is responsible for final evaluation of compliance of the assessment results with the regulatory criteria and the overall confidence in the safety case. As in the previous safety case plan, the safety case will be based on several reports, but

Deep Borehole Disposal Safety Analysis.

Energy Technology Data Exchange (ETDEWEB)

Freeze, Geoffrey A. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Stein, Emily [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Price, Laura L. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); MacKinnon, Robert J. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Tillman, Jack Bruce [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States)

2016-10-01

This report presents a preliminary safety analysis for the deep borehole disposal (DBD) concept, using a safety case framework. A safety case is an integrated collection of qualitative and quantitative arguments, evidence, and analyses that substantiate the safety, and the level of confidence in the safety, of a geologic repository. This safety case framework for DBD follows the outline of the elements of a safety case, and identifies the types of information that will be required to satisfy these elements. At this very preliminary phase of development, the DBD safety case focuses on the generic feasibility of the DBD concept. It is based on potential system designs, waste forms, engineering, and geologic conditions; however, no specific site or regulatory framework exists. It will progress to a site-specific safety case as the DBD concept advances into a site-specific phase, progressing through consent-based site selection and site investigation and characterization.

Reviewing PSA-based analyses to modify technical specifications at nuclear power plants

International Nuclear Information System (INIS)

Samanta, P.K.; Martinez-Guridi, G.; Vesely, W.E.

1995-12-01

Changes to Technical Specifications (TSs) at nuclear power plants (NPPs) require review and approval by the United States Nuclear Regulatory Commission (USNRC). Currently, many requests for changes to TSs use analyses that are based on a plant's probabilistic safety assessment (PSA). This report presents an approach to reviewing such PSA-based submittals for changes to TSs. We discuss the basic objectives of reviewing a PSA-based submittal to modify NPP TSs; the methodology of reviewing a TS submittal, and the differing roles of a PSA review, a PSA Computer Code review, and a review of a TS submittal. To illustrate this approach, we discuss our review of changes to allowed outage time (AOT) and surveillance test interval (STI) in the TS for the South Texas Project Nuclear Generating Station. Based on this experience gained, a check-list of items is given for future reviewers; it can be used to verify that the submittal contains sufficient information, and also that the review has addressed the relevant issues. Finally, recommended steps in the review process and the expected findings of each step are discussed

Studying the Safety Impact of Autonomous Vehicles Using Simulation-Based Surrogate Safety Measures

OpenAIRE

Morando, Mark Mario; Tian, Qingyun; Truong, Long T.; Vu, Hai L.

2018-01-01

Autonomous vehicle (AV) technology has advanced rapidly in recent years with some automated features already available in vehicles on the market. AVs are expected to reduce traffic crashes as the majority of crashes are related to driver errors, fatigue, alcohol, or drugs. However, very little research has been conducted to estimate the safety impact of AVs. This paper aims to investigate the safety impacts of AVs using a simulation-based surrogate safety measure approach. To this end, safety...

Ship Power System Analysis Based on Safety Aspects

Directory of Open Access Journals (Sweden)

Urbaha Margarita

2017-08-01

Full Text Available This article analyses the reasons for the reduction of insulating resistance, processes influencing them and isolation diagnostic methods. It provides a short description of electrical safety situation on ships with isolated neutral electrical power systems. It also covers the methods of protecting personnel from electric shock or preventing ignition or arching damage at the fault location with the help of fault current compensation. Principal fault current compensation circuit diagrams are analysed by using the minimum value and time of transient fault current as criteria.

Advanced handbook for accident analyses of German nuclear power plants; Weiterentwicklung eines Handbuches fuer Stoerfallanalysen deutscher Kernkraftwerke

Energy Technology Data Exchange (ETDEWEB)

Kerner, Alexander; Broecker, Annette; Hartung, Juergen; Mayer, Gerhard; Pallas Moner, Guim

2014-09-15

The advanced handbook of safety analyses (HSA) comprises a comprehensive electronic collection of knowledge for the compilation and conduction of safety analyses in the area of reactor, plant and containment behaviour as well as results of existing safety analyses (performed by GRS in the past) with characteristic specifications and further background information. In addition, know-how from the analysis software development and validation process is presented and relevant rules and regulations with regard to safety demonstration are provided. The HSA comprehensively covers the topic thermo-hydraulic safety analyses (except natural hazards, man-made hazards and malicious acts) for German pressurized and boiling water reactors for power and non-power operational states. In principle, the structure of the HSA-content represents the analytical approach utilized by safety analyses and applying the knowledge from safety analyses to technical support services. On the basis of a multilevel preparation of information to the topics ''compilation of safety analyses'', ''compilation of data bases'', ''assessment of safety analyses'', ''performed safety analyses'', ''rules and regulation'' and ''ATHLET-validation'' the HSA addresses users with different background, allowing them to enter the HSA at different levels. Moreover, the HSA serves as a reference book, which is designed future-oriented, freely configurable related to the content, completely integrated into the GRS internal portal and prepared to be used by a growing user group.

Improving construction site safety through leader-based verbal safety communication.

Science.gov (United States)

Kines, Pete; Andersen, Lars P S; Spangenberg, Soren; Mikkelsen, Kim L; Dyreborg, Johnny; Zohar, Dov

2010-10-01

The construction industry is one of the most injury-prone industries, in which production is usually prioritized over safety in daily on-site communication. Workers have an informal and oral culture of risk, in which safety is rarely openly expressed. This paper tests the effect of increasing leader-based on-site verbal safety communication on the level of safety and safety climate at construction sites. A pre-post intervention-control design with five construction work gangs is carried out. Foremen in two intervention groups are coached and given bi-weekly feedback about their daily verbal safety communications with their workers. Foremen-worker verbal safety exchanges (experience sampling method, n=1,693 interviews), construction site safety level (correct vs. incorrect, n=22,077 single observations), and safety climate (seven dimensions, n=105 questionnaires) are measured over a period of up to 42 weeks. Baseline measurements in the two intervention and three control groups reveal that foremen speak with their workers several times a day. Workers perceive safety as part of their verbal communication with their foremen in only 6-16% of exchanges, and the levels of safety at the sites range from 70-87% (correct observations). Measurements from baseline to follow-up in the two intervention groups reveal that safety communication between foremen and workers increases significantly in one of the groups (factor 7.1 increase), and a significant yet smaller increase is found when the two intervention groups are combined (factor 4.6). Significant increases in the level of safety are seen in both intervention groups (7% and 12% increases, respectively), particularly in regards to 'access ways' and 'railings and coverings' (39% and 84% increases, respectively). Increases in safety climate are seen in only one of the intervention groups with respect to their 'attention to safety.' No significant trend changes are seen in the three control groups on any of the three measures

The ConCom Safety Management Scale: developing and testing a measurement instrument for control-based and commitment-based safety management approaches in hospitals.

Science.gov (United States)

Alingh, Carien W; Strating, Mathilde M H; van Wijngaarden, Jeroen D H; Paauwe, Jaap; Huijsman, Robbert

2018-03-06

Nursing management is considered important for patient safety. Prior research has predominantly focused on charismatic leadership styles, although it is questionable whether these best characterise the role of nurse managers. Managerial control is also relevant. Therefore, we aimed to develop and test a measurement instrument for control-based and commitment-based safety management of nurse managers in clinical hospital departments. A cross-sectional survey design was used to test the newly developed questionnaire in a sample of 2378 nurses working in clinical departments. The nurses were asked about their perceptions of the leadership behaviour and management practices of their direct supervisors. Psychometric properties were evaluated using confirmatory factor analysis and reliability estimates. The final 33-item questionnaire showed acceptable goodness-of-fit indices and internal consistency (Cronbach's α of the subscales range: 0.59-0.90). The factor structure revealed three subdimensions for control-based safety management: (1) stressing the importance of safety rules and regulations; (2) monitoring compliance; and (3) providing employees with feedback. Commitment-based management consisted of four subdimensions: (1) showing role modelling behaviour; (2) creating safety awareness; (3) showing safety commitment; and (4) encouraging participation. Construct validity of the scale was supported by high factor loadings and provided preliminary evidence that control-based and commitment-based safety management are two distinct yet related constructs. The findings were reconfirmed in a cross-validation procedure. The results provide initial support for the construct validity and reliability of our ConCom Safety Management Scale. Both management approaches were found to be relevant for managing patient safety in clinical hospital departments. The scale can be used to deepen our understanding of the influence of patient safety management on healthcare professionals

Fusion reactor passive safety and ignitor risk-based regulation

International Nuclear Information System (INIS)

Zucchetti, M.

1995-01-01

Passive design features are more reliable than operator action of successful operation of active safety systems. Passive safety has usually been adopted for fission. The achievement of an inventory-based passive safety is difficult if the fusion reactor uses neutronic reactions. Ignitor is a high-magnetic field tokamak designed to study the physics of ignited plasmas. The safety goal for Ignitor is classification as a mobility-based passively safe machine

Software system safety

Science.gov (United States)

Uber, James G.

1988-01-01

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

A web-based tool for the Comprehensive Unit-based Safety Program (CUSP).

Science.gov (United States)

Pronovost, Peter J; King, Jay; Holzmueller, Christine G; Sawyer, Melinda; Bivens, Shauna; Michael, Michelle; Haig, Kathy; Paine, Lori; Moore, Dana; Miller, Marlene

2006-03-01

An organization's ability to change is driven by its culture, which in turn has a significant impact on safety. The six-step Comprehensive Unit-Based Safety Program (CUSP) is intended to improve local culture and safety. A Web-based project management tool for CUSP was developed and then pilot tested at two hospitals. HOW ECUSP WORKS: Once a patient safety concern is identified (step 3), a unit-level interdisciplinary safety committee determines issue criticality and starts up the projects (step 4), which are managed using project management tools within eCUSP (step 5). On a project's completion, the results are disseminated through a shared story (step 6). OSF St. Joseph's Medical Center-The Medical Birthing Center (Bloomington, Illinois), identified 11 safety issues, implemented 11 projects, and created 9 shared stories--including one for its Armband Project. The Johns Hopkins Hospital (Baltimore) Medical Progressive Care (MPC4) Unit identified 5 safety issues and implemented 4 ongoing projects, including the intravenous (IV) Tubing Compliance Project. The eCUSP tool's success depends on an organizational commitment to creating a culture of safety.

Construction safety monitoring based on the project's characteristic with fuzzy logic approach

Science.gov (United States)

Winanda, Lila Ayu Ratna; Adi, Trijoko Wahyu; Anwar, Nadjadji; Wahyuni, Febriana Santi

2017-11-01

Construction workers accident is the highest number compared with other industries and falls are the main cause of fatal and serious injuries in high rise projects. Generally, construction workers accidents are caused by unsafe act and unsafe condition that can occur separately or together, thus a safety monitoring system based on influencing factors is needed to achieve zero accident in construction industry. The dynamic characteristic in construction causes high mobility for workers while doing the task, so it requires a continuously monitoring system to detect unsafe condition and to protect workers from potential hazards. In accordance with the unique nature of project, fuzzy logic approach is one of the appropriate methods for workers safety monitoring on site. In this study, the focus of discussion is based on the characteristic of construction projects in analyzing "potential hazard" and the "protection planning" to be used in accident prevention. The data have been collected from literature review, expert opinion and institution of safety and health. This data used to determine hazard identification. Then, an application model is created using Delphi programming. The process in fuzzy is divided into fuzzification, inference and defuzzification, according to the data collection. Then, the input and final output data are given back to the expert for assessment as a validation of application model. The result of the study showed that the potential hazard of construction workers accident could be analysed based on characteristic of project and protection system on site and fuzzy logic approach can be used for construction workers accident analysis. Based on case study and the feedback assessment from expert, it showed that the application model can be used as one of the safety monitoring tools.

Applying Sensor-Based Technology to Improve Construction Safety Management.

Science.gov (United States)

Zhang, Mingyuan; Cao, Tianzhuo; Zhao, Xuefeng

2017-08-11

Construction sites are dynamic and complicated systems. The movement and interaction of people, goods and energy make construction safety management extremely difficult. Due to the ever-increasing amount of information, traditional construction safety management has operated under difficult circumstances. As an effective way to collect, identify and process information, sensor-based technology is deemed to provide new generation of methods for advancing construction safety management. It makes the real-time construction safety management with high efficiency and accuracy a reality and provides a solid foundation for facilitating its modernization, and informatization. Nowadays, various sensor-based technologies have been adopted for construction safety management, including locating sensor-based technology, vision-based sensing and wireless sensor networks. This paper provides a systematic and comprehensive review of previous studies in this field to acknowledge useful findings, identify the research gaps and point out future research directions.

IMPACTS OF GROUP-BASED SIGNAL CONTROL POLICY ON DRIVER BEHAVIOR AND INTERSECTION SAFETY

Directory of Open Access Journals (Sweden)

Keshuang TANG

2008-01-01

Full Text Available Unlike the typical stage-based policy commonly applied in Japan, the group-based control (often called movement-based in the traffic control industry in Japan refers to such a control pattern that the controller is capable of separately allocating time to each signal group instead of stage based on traffic demand. In order to investigate its applicability at signalized intersections in Japan, an intersection located in Yokkaichi City of Mie Prefecture was selected as an experimental application site by the Japan Universal Traffic Management Society (UTMS. Based on the data collected at the intersection before and after implementing the group-based control policy respectively, this study evaluated the impacts of such a policy on driver behavior and intersection safety. To specify those impacts, a few models utilizing cycle-based data were first developed to interpret the occurrence probability and rate of red-light-running (RLR. Furthermore, analyses were performed on the yellow-entry time (Ye of the last cleared vehicle and post encroachment time (PET during the phase switching. Conclusions supported that the group-based control policy, along with certain other factors, directly or indirectly influenced the RLR behavior of through and right-turn traffics. Meanwhile, it has potential safety benefits as well, indicated by the declined Ye and increased PET values.

Model-based safety architecture framework for complex systems

NARCIS (Netherlands)

Schuitemaker, Katja; Rajabali Nejad, Mohammadreza; Braakhuis, J.G.; Podofillini, Luca; Sudret, Bruno; Stojadinovic, Bozidar; Zio, Enrico; Kröger, Wolfgang

2015-01-01

The shift to transparency and rising need of the general public for safety, together with the increasing complexity and interdisciplinarity of modern safety-critical Systems of Systems (SoS) have resulted in a Model-Based Safety Architecture Framework (MBSAF) for capturing and sharing architectural

Implementation of Recommendations from the One System Comparative Evaluation of the Hanford Tank Farms and Waste Treatment Plant Safety Bases

International Nuclear Information System (INIS)

Garrett, Richard L.; Niemi, Belinda J.; Paik, Ingle K.; Buczek, Jeffrey A.; Lietzow, J.; McCoy, F.; Beranek, F.; Gupta, M.

2013-01-01

A Comparative Evaluation was conducted for One System Integrated Project Team to compare the safety bases for the Hanford Waste Treatment and Immobilization Plant Project (WTP) and Tank Operations Contract (TOC) (i.e., Tank Farms) by an Expert Review Team. The evaluation had an overarching purpose to facilitate effective integration between WTP and TOC safety bases. It was to provide One System management with an objective evaluation of identified differences in safety basis process requirements, guidance, direction, procedures, and products (including safety controls, key safety basis inputs and assumptions, and consequence calculation methodologies) between WTP and TOC. The evaluation identified 25 recommendations (Opportunities for Integration). The resolution of these recommendations resulted in 16 implementation plans. The completion of these implementation plans will help ensure consistent safety bases for WTP and TOC along with consistent safety basis processes. procedures, and analyses. and should increase the likelihood of a successful startup of the WTP. This early integration will result in long-term cost savings and significant operational improvements. In addition, the implementation plans lead to the development of eight new safety analysis methodologies that can be used at other U.S. Department of Energy (US DOE) complex sites where URS Corporation is involved

Safety analysis procedures for PHWR

International Nuclear Information System (INIS)

Min, Byung Joo; Kim, Hyoung Tae; Yoo, Kun Joong

2004-03-01

The methodology of safety analyses for CANDU reactors in Canada, a vendor country, uses a combination of best-estimate physical models and conservative input parameters so as to minimize the uncertainty of the plant behavior predictions. As using the conservative input parameters, the results of the safety analyses are assured the regulatory requirements such as the public dose, the integrity of fuel and fuel channel, the integrity of containment and reactor structures, etc. However, there is not the comprehensive and systematic procedures for safety analyses for CANDU reactors in Korea. In this regard, the development of the safety analyses procedures for CANDU reactors is being conducted not only to establish the safety analyses system, but also to enhance the quality assurance of the safety assessment. In the first phase of this study, the general procedures of the deterministic safety analyses are developed. The general safety procedures are covered the specification of the initial event, selection of the methodology and accident sequences, computer codes, safety analysis procedures, verification of errors and uncertainties, etc. Finally, These general procedures of the safety analyses are applied to the Large Break Loss Of Coolant Accident (LBLOCA) in Final Safety Analysis Report (FSAR) for Wolsong units 2, 3, 4

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

International Nuclear Information System (INIS)

Koh, Jung Soo

1997-02-01

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

Nature-Based Strategies for Improving Urban Health and Safety.

Science.gov (United States)

Kondo, Michelle C; South, Eugenia C; Branas, Charles C

2015-10-01

Place-based programs are being noticed as key opportunities to prevent disease and promote public health and safety for populations at-large. As one key type of place-based intervention, nature-based and green space strategies can play an especially large role in improving health and safety for dwellers in urban environments such as US legacy cities that lack nature and greenery. In this paper, we describe the current understanding of place-based influences on public health and safety. We focus on nonchemical environmental factors, many of which are related to urban abandonment and blight. We then review findings from studies of nature-based interventions regarding impacts on health, perceptions of safety, and crime. Based on our findings, we suggest that further research in this area will require (1) refined measures of green space, nature, and health and safety for cities, (2) interdisciplinary science and cross-sector policy collaboration, (3) observational studies as well as randomized controlled experiments and natural experiments using appropriate spatial counterfactuals and mixed methods, and (4) return-on-investment calculations of potential economic, social, and health costs and benefits of urban greening initiatives.

Mining Behavior Based Safety Data to Predict Safety Performance

Energy Technology Data Exchange (ETDEWEB)

Jeffrey C. Joe

2010-06-01

The Idaho National Laboratory (INL) operates a behavior based safety program called Safety Observations Achieve Results (SOAR). This peer-to-peer observation program encourages employees to perform in-field observations of each other's work practices and habits (i.e., behaviors). The underlying premise of conducting these observations is that more serious accidents are prevented from occurring because lower level “at risk” behaviors are identified and corrected before they can propagate into culturally accepted “unsafe” behaviors that result in injuries or fatalities. Although the approach increases employee involvement in safety, the premise of the program has not been subject to sufficient empirical evaluation. The INL now has a significant amount of SOAR data on these lower level “at risk” behaviors. This paper describes the use of data mining techniques to analyze these data to determine whether they can predict if and when a more serious accident will occur.

[Does simulator-based team training improve patient safety?].

Science.gov (United States)

Trentzsch, H; Urban, B; Sandmeyer, B; Hammer, T; Strohm, P C; Lazarovici, M

2013-10-01

Patient safety became paramount in medicine as well as in emergency medicine after it was recognized that preventable, adverse events significantly contributed to morbidity and mortality during hospital stay. The underlying errors cannot usually be explained by medical technical inadequacies only but are more due to difficulties in the transition of theoretical knowledge into tasks under the conditions of clinical reality. Crew Resource Management and Human Factors which determine safety and efficiency of humans in complex situations are suitable to control such sources of error. Simulation significantly improved safety in high reliability organizations, such as the aerospace industry.Thus, simulator-based team training has also been proposed for medical areas. As such training is consuming in cost, time and human resources, the question of the cost-benefit ratio obviously arises. This review outlines the effects of simulator-based team training on patient safety. Such course formats are not only capable of creating awareness and improvements in safety culture but also improve technical team performance and emphasize team performance as a clinical competence. A few studies even indicated improvement of patient-centered outcome, such as a reduced rate of adverse events but further studies are required in this respect. In summary, simulator-based team training should be accepted as a suitable strategy to improve patient safety.

ATHENA/INTRA analyses for ITER, NSSR-2

International Nuclear Information System (INIS)

Shen, Kecheng; Eriksson, John; Sjoeberg, A.

1999-02-01

The present report is a summary report including thermal-hydraulic analyses made at Studsvik Eco and Safety AB for the ITER NSSR-2 safety documentation. The objective of the analyses was to reveal the safety characteristics of various heat transfer systems at specified operating conditions and to indicate the conditions for which there were obvious risks of jeopardising the structural integrity of the coolant systems. In the latter case also some analyses were made to indicate conceivable mitigating measures for maintaining the integrity.The analyses were primarily concerned with the First Wall and Divertor heat transfer systems. Several enveloping transients were analysed with associated specific flow and heat load boundary conditions. The analyses were performed with the ATHENA and INTRA codes

ATHENA/INTRA analyses for ITER, NSSR-2

Energy Technology Data Exchange (ETDEWEB)

Shen, Kecheng; Eriksson, John; Sjoeberg, A

1999-02-01

The present report is a summary report including thermal-hydraulic analyses made at Studsvik Eco and Safety AB for the ITER NSSR-2 safety documentation. The objective of the analyses was to reveal the safety characteristics of various heat transfer systems at specified operating conditions and to indicate the conditions for which there were obvious risks of jeopardising the structural integrity of the coolant systems. In the latter case also some analyses were made to indicate conceivable mitigating measures for maintaining the integrity.The analyses were primarily concerned with the First Wall and Divertor heat transfer systems. Several enveloping transients were analysed with associated specific flow and heat load boundary conditions. The analyses were performed with the ATHENA and INTRA codes 8 refs, 14 figs, 15 tabs

Development of safety analysis technology for LMR

International Nuclear Information System (INIS)

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

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

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.

Selected problems and results of the transient event and reliability analyses for the German safety study

International Nuclear Information System (INIS)

Hoertner, H.

1977-01-01

For the investigation of the risk of nuclear power plants loss-of-coolant accidents and transients have to be analyzed. The different functions of the engineered safety features installed to cope with transients are explained. The event tree analysis is carried out for the important transient 'loss of normal onsite power'. Preliminary results of the reliability analyses performed for quantitative evaluation of this event tree are shown. (orig.) [de

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

Development of an FPGA-based controller for safety critical application

International Nuclear Information System (INIS)

Xing, A.; De Grosbois, J.; Sklyar, V.; Archer, P.; Awwal, A.

2011-01-01

In implementing safety functions, Field Programmable Gate Arrays (FPGA) technology offers a distinct combination of benefits and advantages over microprocessor-based systems. FPGAs can be designed such that the final product is purely hardware, without any overhead runtime software, bringing the design closer to a conventional hardware-based solution. On the other hand, FPGAs can implement more complex safety logic that would generally require microprocessor-based safety systems. There are now qualified FPGA-based platforms available on the market with a credible use history in safety applications in nuclear power plants. Atomic Energy of Canada (AECL), in collaboration with RPC Radiy, has initiated a development program to define a vigorous FPGA engineering process suitable for implementing safety critical functions at the application development level. This paper provides an update on the FPGA development program along with the proposed design model using function block diagrams for the development of safety controllers in CANDU applications. (author)

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.)

Safety strategy and safety analysis of nuclear power plants

International Nuclear Information System (INIS)

Franzen, L.F.

1976-01-01

The safety strategy for nuclear power plants is characterized by the fact that the high level of safety was attained not as a result of experience, but on the basis of preventive accident analyses and the finding derived from such analyses. Although, in these accident analyses, the deterministic approach is predominant, it is supplemented by reliability analyses. The accidents analyzed in nuclear licensing procedures cover a wide spectrum from minor incidents to the design basis accidents which determine the design of the safety devices. The initial and boundary conditions, which are essentail for accident analyses, and the determination of the loads occurring in various states during regular operation and in accidents flow into the design of the individual systems and components. The inevitable residual risk and its origins are discussed. (orig.) [de

Development of reliability database for safety-related I and C component based on operating experience of KSNP

International Nuclear Information System (INIS)

Jang, S. C.; Han, S. H.; Min, K. R.

2001-01-01

Reliability database for safety-related I and C components has been developed, based on domestic operating experience of total 8.63 years from four units-Yonggwang Units 3 and 4, and Ulchin Units 3 and 4. This plant-specific data of safety-related I and C components has compared with operating experience for CE-supplied plants in U.S.A. As a results, we found that on the whole the domestic reliability data was similar to CE-supplied plants in USA, through lots of failures occurred early in the commercial operation were included in our analyses without percolation

TVSA-T fuel assembly for 'Temelin' NPP. Main results of design and safety analyses. Trends of development

International Nuclear Information System (INIS)

Samojlov, O.B.; Kajdalov, V.B.; Falkov, A.A.; Bolnov, V.A.; Morozkin, O.N.; Molchanov, V.L.; Ugryumov, A.V.

2010-01-01

TVSA is a fuel assembly with rigid skeleton formed by 6 angle pieces and SG is successfully operated at 17 VVER-1000 power units of Kalinin NPP, as well as at Ukrainian and Bulgarian NPPs. Based on a contract for fuel supply to the Temelin NPP, the TVSA-T fuel assembly was developed, building on proven solutions confirmed by operation of TVSA modifications during 4-6 years and by the results of post-irradiation examination. The TVSA-T design includes combined spacer grids (SG+MG) and by fuel column elongation by 150 mm. A set of analyses and experiments was performed to validate the design, including thermal hydraulic tests, validation of critical heat flux correlation for TVSA-T, integrated mechanical, vibration and lifetime tests. A licence to use the fuel has been granted by the Czech State Office for Nuclear Safety. The TVSA-T core is currently in operation at the Temelin-1 reactor unit. The presentation is concluded as follows: TVSA-T fuel assembly for Temelin has been validated. The TVSA-T design is based on approved technical decisions and meets the current requirements for lifetime, operational maneuverability and safety. The results of post-irradiation examination of TVSA-T operated at the Kalinin-1 unit for 4 years confirm the assembly operability, skeleton stiffness, geometric stability and normal fuel rod cladding condition. The properties of the TVSA fuel with MG allow the core power to be increased up to 3300 MW to match the envisaged future VVER (MIR-1200) design, providing allowable fuel rod power FΔh =1.63 (to implement effective fuel cycles). (P.A.)

A web-based endpoint adjudication system for interim analyses in clinical trials.

Science.gov (United States)

Nolen, Tracy L; Dimmick, Bill F; Ostrosky-Zeichner, Luis; Kendrick, Amy S; Sable, Carole; Ngai, Angela; Wallace, Dennis

2009-02-01

A data monitoring committee (DMC) is often employed to assess trial progress and review safety data and efficacy endpoints throughout a trail. Interim analyses performed for the DMC should use data that are as complete and verified as possible. Such analyses are complicated when data verification involves subjective study endpoints or requires clinical expertise to determine each subject's status with respect to the study endpoint. Therefore, procedures are needed to obtain adjudicated data for interim analyses in an efficient manner. In the past, methods for handling such data included using locally reported results as surrogate endpoints, adjusting analysis methods for unadjudicated data, or simply performing the adjudication as rapidly as possible. These methods all have inadequacies that make their sole usage suboptimal. For a study of prophylaxis for invasive candidiasis, adjudication of both study eligibility criteria and clinical endpoints prior to two interim analyses was required. Because the study was expected to enroll at a moderate rate and the sponsor required adjudicated endpoints to be used for interim analyses, an efficient process for adjudication was required. We created a web-based endpoint adjudication system (WebEAS) that allows for expedited review by the endpoint adjudication committee (EAC). This system automatically identifies when a subject's data are complete, creates a subject profile from the study data, and assigns EAC reviewers. The reviewers use the WebEAS to review the subject profile and submit their completed review form. The WebEAS then compares the reviews, assigns an additional review as a tiebreaker if needed, and stores the adjudicated data. The study for which this system was originally built was administratively closed after 10 months with only 38 subjects enrolled. The adjudication process was finalized and the WebEAS system activated prior to study closure. Some website accessibility issues presented initially. However

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)

Labor unions and safety climate: perceived union safety values and retail employee safety outcomes.

Science.gov (United States)

Sinclair, Robert R; Martin, James E; Sears, Lindsay E

2010-09-01

Although trade unions have long been recognized as a critical advocate for employee safety and health, safety climate research has not paid much attention to the role unions play in workplace safety. We proposed a multiple constituency model of workplace safety which focused on three central safety stakeholders: top management, ones' immediate supervisor, and the labor union. Safety climate research focuses on management and supervisors as key stakeholders, but has not considered whether employee perceptions about the priority their union places on safety contributes contribute to safety outcomes. We addressed this gap in the literature by investigating unionized retail employee (N=535) perceptions about the extent to which their top management, immediate supervisors, and union valued safety. Confirmatory factor analyses demonstrated that perceived union safety values could be distinguished from measures of safety training, workplace hazards, top management safety values, and supervisor values. Structural equation analyses indicated that union safety values influenced safety outcomes through its association with higher safety motivation, showing a similar effect as that of supervisor safety values. These findings highlight the need for further attention to union-focused measures related to workplace safety as well as further study of retail employees in general. We discuss the practical implications of our findings and identify several directions for future safety research. 2009 Elsevier Ltd. All rights reserved.

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)

A Nuclear Safety System based on Industrial Computer

International Nuclear Information System (INIS)

Kim, Ji Hyeon; Oh, Do Young; Lee, Nam Hoon; Kim, Chang Ho; Kim, Jae Hack

2011-01-01

The Plant Protection System(PPS), a nuclear safety Instrumentation and Control (I and C) system for Nuclear Power Plants(NPPs), generates reactor trip on abnormal reactor condition. The Core Protection Calculator System (CPCS) is a safety system that generates and transmits the channel trip signal to the PPS on an abnormal condition. Currently, these systems are designed on the Programmable Logic Controller(PLC) based system and it is necessary to consider a new system platform to adapt simpler system configuration and improved software development process. The CPCS was the first implementation using a micro computer in a nuclear power plant safety protection system in 1980 which have been deployed in Ulchin units 3,4,5,6 and Younggwang units 3,4,5,6. The CPCS software was developed in the Concurrent Micro5 minicomputer using assembly language and embedded into the Concurrent 3205 computer. Following the micro computer based CPCS, PLC based Common-Q platform has been used for the ShinKori/ShinWolsong units 1,2 PPS and CPCS, and the POSAFE-Q PLC platform is used for the ShinUlchin units 1,2 PPS and CPCS. In developing the next generation safety system platform, several factors (e.g., hardware/software reliability, flexibility, licensibility and industrial support) can be considered. This paper suggests an Industrial Computer(IC) based protection system that can be developed with improved flexibility without losing system reliability. The IC based system has the advantage of a simple system configuration with optimized processor boards because of improved processor performance and unlimited interoperability between the target system and development system that use commercial CASE tools. This paper presents the background to selecting the IC based system with a case study design of the CPCS. Eventually, this kind of platform can be used for nuclear power plant safety systems like the PPS, CPCS, Qualified Indication and Alarm . Pami(QIAS-P), and Engineering Safety

A Nuclear Safety System based on Industrial Computer

Energy Technology Data Exchange (ETDEWEB)

Kim, Ji Hyeon; Oh, Do Young; Lee, Nam Hoon; Kim, Chang Ho; Kim, Jae Hack [Korea Electric Power Corporation Engineering and Construction, Daejeon (Korea, Republic of)

2011-05-15

The Plant Protection System(PPS), a nuclear safety Instrumentation and Control (I and C) system for Nuclear Power Plants(NPPs), generates reactor trip on abnormal reactor condition. The Core Protection Calculator System (CPCS) is a safety system that generates and transmits the channel trip signal to the PPS on an abnormal condition. Currently, these systems are designed on the Programmable Logic Controller(PLC) based system and it is necessary to consider a new system platform to adapt simpler system configuration and improved software development process. The CPCS was the first implementation using a micro computer in a nuclear power plant safety protection system in 1980 which have been deployed in Ulchin units 3,4,5,6 and Younggwang units 3,4,5,6. The CPCS software was developed in the Concurrent Micro5 minicomputer using assembly language and embedded into the Concurrent 3205 computer. Following the micro computer based CPCS, PLC based Common-Q platform has been used for the ShinKori/ShinWolsong units 1,2 PPS and CPCS, and the POSAFE-Q PLC platform is used for the ShinUlchin units 1,2 PPS and CPCS. In developing the next generation safety system platform, several factors (e.g., hardware/software reliability, flexibility, licensibility and industrial support) can be considered. This paper suggests an Industrial Computer(IC) based protection system that can be developed with improved flexibility without losing system reliability. The IC based system has the advantage of a simple system configuration with optimized processor boards because of improved processor performance and unlimited interoperability between the target system and development system that use commercial CASE tools. This paper presents the background to selecting the IC based system with a case study design of the CPCS. Eventually, this kind of platform can be used for nuclear power plant safety systems like the PPS, CPCS, Qualified Indication and Alarm . Pami(QIAS-P), and Engineering Safety

A new approach to determine the environmental qualification requirements for the safety related equipment

International Nuclear Information System (INIS)

Hasnaoui, C.; Parent, G.

2000-01-01

The objective of the environmental qualification of safety related equipment is to ensure that the plant defense-in-depth is not compromised by common mode failures following design basis accidents with a harsh environment. A new approach based on safety functions has been developed to determine what safety-related equipment is required to function during and after a design basis accident, as well as their environmental qualification requirements. The main feature of this approach is to use auxiliary safety functions established from safety requirements as credited in the safety analyses. This approach is undertaken in three steps: identification of the auxiliary safety functions of each main safety function; determination of the main equipment groups required for each auxiliary safety function; and review of the safety analyses for design basis accidents in order to determine the credited auxiliary safety functions and their mission times for each accident scenario. Some of the benefits of the proposed approach for the determination of the safety environmental qualification requirements are: a systematic approach for the review of safety analyses based on a safety function check list, and the insurance, with the availability of the safety functions, that Gentilly-2 defense-in-depth would not be compromised by design basis accidents with a harsh environment. (author)

Human factors in safety assessment. Safety culture assessment

International Nuclear Information System (INIS)

Zhang Li; Deng Zhiliang; Wang Yiqun; Huang Weigang

1996-01-01

This paper analyses the present conditions and problems in enterprises safety assessment, and introduces the characteristics and effects of safety culture. The authors think that safety culture must be used as a 'soul' to form the pattern of modern safety management. Furthermore, they propose that the human safety and synthetic safety management assessment in a system should be changed into safety culture assessment. Finally, the assessment indicators are discussed

Nature-based strategies for improving urban health and safety

Science.gov (United States)

Michelle C. Kondo; Eugenia C. South; Charles C. Branas

2015-01-01

Place-based programs are being noticed as key opportunities to prevent disease and promote public health and safety for populations at-large. As one key type of place-based intervention, nature-based and green space strategies can play an especially large role in improving health and safety for dwellers in urban environments such as US legacy cities that lack nature...

THE FLUORBOARD A STATISTICALLY BASED DASHBOARD METHOD FOR IMPROVING SAFETY

International Nuclear Information System (INIS)

PREVETTE, S.S.

2005-01-01

The FluorBoard is a statistically based dashboard method for improving safety. Fluor Hanford has achieved significant safety improvements--including more than a 80% reduction in OSHA cases per 200,000 hours, during its work at the US Department of Energy's Hanford Site in Washington state. The massive project on the former nuclear materials production site is considered one of the largest environmental cleanup projects in the world. Fluor Hanford's safety improvements were achieved by a committed partnering of workers, managers, and statistical methodology. Safety achievements at the site have been due to a systematic approach to safety. This includes excellent cooperation between the field workers, the safety professionals, and management through OSHA Voluntary Protection Program principles. Fluor corporate values are centered around safety, and safety excellence is important for every manager in every project. In addition, Fluor Hanford has utilized a rigorous approach to using its safety statistics, based upon Dr. Shewhart's control charts, and Dr. Deming's management and quality methods

Overview of Risk Mitigation for Safety-Critical Computer-Based Systems

Science.gov (United States)

Torres-Pomales, Wilfredo

2015-01-01

This report presents a high-level overview of a general strategy to mitigate the risks from threats to safety-critical computer-based systems. In this context, a safety threat is a process or phenomenon that can cause operational safety hazards in the form of computational system failures. This report is intended to provide insight into the safety-risk mitigation problem and the characteristics of potential solutions. The limitations of the general risk mitigation strategy are discussed and some options to overcome these limitations are provided. This work is part of an ongoing effort to enable well-founded assurance of safety-related properties of complex safety-critical computer-based aircraft systems by developing an effective capability to model and reason about the safety implications of system requirements and design.

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)

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

Directory of Open Access Journals (Sweden)

Ai Lin Evelyn Teo

2016-12-01

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

MATHEMATICAL APPARATUS FOR KNOWLEDGE BASE PROJECT MANAGEMENT OF OCCUPATIONAL SAFETY

Directory of Open Access Journals (Sweden)

Валентина Николаевна ПУРИЧ

2015-05-01

Full Text Available The occupational safety project (OSP management is aimed onto a rational choice implementation. With respect to the subjectivity of management goals the project selection is considered as a minimum formalization level information process, The proposed project selection model relies upon the enterprise’s occupational and industrial safety assessment using fuzzy logic and linguistic variables based on occupational safety knowledge base.

KHNP Safety Culture Framework based on Global Standard, and Lessons learned from Safety Culture Evaluation

International Nuclear Information System (INIS)

Kim, Younggab; Hur, Nam Young; Jeong, Hyeon Jong

2015-01-01

In order to eliminate the vague fears of the people about the nuclear power and operate continuously NPPs, a strong safety culture of NPPs should be demonstrated. Strong safety culture awareness of workers can overcome social distrust about NPPs. KHNP has been a variety efforts to improve and establish safety culture of NPPs. Safety culture framework applying global standards was set up and safety culture assessment has been carried out periodically to enhance safety culture of workers. In addition, KHNP developed various safety culture contents and they are being used in NPPs by workers. As a result of these efforts, safety culture awareness of workers is changed positively and the safety environment of NPPs is expected to be improved. KHNP makes an effort to solve areas for improvement derived from safety culture assessment. However, there are some areas to take a long time in completing the work. Therefore, these actions are necessary to be carried out consistently and continuously. KHNP also developed recently safety culture enhancement system based on web. All information related to safety culture in KHNP will be shared through this web system and this system will be used to safety culture assessment. In addition to, KHNP plans to develop safety culture indicators for monitoring the symptoms of safety culture weakening

KHNP Safety Culture Framework based on Global Standard, and Lessons learned from Safety Culture Evaluation

Energy Technology Data Exchange (ETDEWEB)

Kim, Younggab; Hur, Nam Young; Jeong, Hyeon Jong [KHNP Central Research Institute, Daejeon (Korea, Republic of)

2015-05-15

In order to eliminate the vague fears of the people about the nuclear power and operate continuously NPPs, a strong safety culture of NPPs should be demonstrated. Strong safety culture awareness of workers can overcome social distrust about NPPs. KHNP has been a variety efforts to improve and establish safety culture of NPPs. Safety culture framework applying global standards was set up and safety culture assessment has been carried out periodically to enhance safety culture of workers. In addition, KHNP developed various safety culture contents and they are being used in NPPs by workers. As a result of these efforts, safety culture awareness of workers is changed positively and the safety environment of NPPs is expected to be improved. KHNP makes an effort to solve areas for improvement derived from safety culture assessment. However, there are some areas to take a long time in completing the work. Therefore, these actions are necessary to be carried out consistently and continuously. KHNP also developed recently safety culture enhancement system based on web. All information related to safety culture in KHNP will be shared through this web system and this system will be used to safety culture assessment. In addition to, KHNP plans to develop safety culture indicators for monitoring the symptoms of safety culture weakening.

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

International Nuclear Information System (INIS)

Koh, Jung Soo; Seong, Poong Hyun

1997-01-01

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

Review of accident analyses of RB experimental reactor

International Nuclear Information System (INIS)

Pesic, M.

2003-01-01

The RB reactor is a uranium fuel heavy water moderated critical assembly that has been put and kept in operation by the VINCA Institute of Nuclear Sciences, Belgrade, Serbia and Montenegro, since April 1958. The first complete Safety Analysis Report of the RB reactor was prepared in 1961/62; yet, the first accident analysis had been made in late 1958 with the aim to examine a power transition and the total equivalent doses received by the staff during the reactivity accident that occurred on October 15, 1958. Since 1960, the RB reactor has been modified a few times. Beside the initial natural uranium metal fuel rods, new types of fuel (TVR-S types of Russian origin) consisting of 2% enriched uranium metal and 80% enriched U0 2 , dispersed in aluminum matrix, have been available since 1962 and 1976, respectively. Modifications of the control and safety systems of the reactor were made occasionally. Special reactor cores were designed and constructed using all three types of fuel elements, as well as the coupled fast-thermal ones. The Nuclear Safety Committee of the VINCA Institute, an independent regulator)' body, approved for usage all these modifications of the RB reactor on the basis of the Preliminary Safety' Analysis Reports, which, beside proposed technical modifications and new regulation rules, included safety analyses of various possible accidents. A special attention was given (and a new safety methodology was proposed) to thorough analyses of the design-based accidents related to the coupled fast-thermal cores that included central zones of the reactor filled by the fuel elements without any moderator. In this paper, an overview of some accidents, methodologies and computation tools used for the accident analyses of the RB reactor is given. (author)

Review of accident analyses of RB experimental reactor

Directory of Open Access Journals (Sweden)

Pešić Milan P.

2003-01-01

Full Text Available The RB reactor is a uranium fuel heavy water moderated critical assembly that has been put and kept in operation by the VTNCA Institute of Nuclear Sciences, Belgrade, Serbia and Montenegro, since April 1958. The first complete Safety Analysis Report of the RB reactor was prepared in 1961/62 yet, the first accident analysis had been made in late 1958 with the aim to examine a power transition and the total equivalent doses received by the staff during the reactivity accident that occurred on October 15, 1958. Since 1960, the RB reactor has been modified a few times. Beside the initial natural uranium metal fuel rods, new types of fuel (TVR-S types of Russian origin consisting of 2% enriched uranium metal and 80% enriched UO2 dispersed in aluminum matrix, have been available since 1962 and 1976 respectively. Modifications of the control and safety systems of the reactor were made occasionally. Special reactor cores were designed and constructed using all three types of fuel elements as well as the coupled fast-thermal ones. The Nuclear Safety Committee of the VINĆA Institute, an independent regulatory body, approved for usage all these modifications of the RB reactor on the basis of the Preliminary Safety Analysis Reports, which, beside proposed technical modifications and new regulation rules, included safety analyses of various possible accidents. A special attention was given (and a new safety methodology was proposed to thorough analyses of the design-based accidents related to the coupled fast-thermal cores that included central zones of the reactor filled by the fuel elements without any moderator. In this paper, an overview of some accidents, methodologies and computation tools used for the accident analyses of the RB reactor is given.

The Relationship Among School Safety, School Liking, and Students' Self-Esteem: Based on a Multilevel Mediation Model.

Science.gov (United States)

Zhang, Xinghui; Xuan, Xin; Chen, Fumei; Zhang, Cai; Luo, Yuhan; Wang, Yun

2016-03-01

Perceptions of school safety have an important effect on students' development. Based on the model of "context-process-outcomes," we examined school safety as a context variable to explore how school safety at the school level affected students' self-esteem. We used hierarchical linear modeling to examine the link between school safety at the school level and students' self-esteem, including school liking as a mediator. The data were from the National Children's Study of China (NCSC), in which 6618 fourth- to fifth-grade students in 79 schools were recruited from 100 counties in 31 provinces in China. Multilevel mediation analyses showed that the positive relationship between school safety at the school level and self-esteem was partially mediated by school liking, controlling for demographics at both student and school levels. Furthermore, a sex difference existed in the multilevel mediation model. For boys, school liking fully mediated the relationship between school safety at the school level and self-esteem. However, school liking partially mediated the relationship between school safety at the school level and self-esteem among girls. School safety should receive increasing attention from policymakers because of its impact on students' self-esteem. © 2016, American School Health Association.

General design safety principles for nuclear power plants

International Nuclear Information System (INIS)

1986-01-01

This Safety Guide provides the safety principles and the approach that have been used to implement the Code in the Safety Guides. These safety principles and the approach are tied closely to the safety analyses needed to assist the design process, and are used to verify the adequacy of nuclear power plant designs. This Guide also provides a framework for the use of other design Safety Guides. However, although it explains the principles on which the other Safety Guides are based, the requirements for specific applications of these principles are mostly found in the other Guides

Risk-based rules for crane safety systems

Energy Technology Data Exchange (ETDEWEB)

Ruud, Stian [Section for Control Systems, DNV Maritime, 1322 Hovik (Norway)], E-mail: Stian.Ruud@dnv.com; Mikkelsen, Age [Section for Lifting Appliances, DNV Maritime, 1322 Hovik (Norway)], E-mail: Age.Mikkelsen@dnv.com

2008-09-15

The International Maritime Organisation (IMO) has recommended a method called formal safety assessment (FSA) for future development of rules and regulations. The FSA method has been applied in a pilot research project for development of risk-based rules and functional requirements for systems and components for offshore crane systems. This paper reports some developments in the project. A method for estimating target reliability for the risk-control options (safety functions) by means of the cost/benefit decision criterion has been developed in the project and is presented in this paper. Finally, a structure for risk-based rules is proposed and presented.

Risk-based rules for crane safety systems

International Nuclear Information System (INIS)

Ruud, Stian; Mikkelsen, Age

2008-01-01

The International Maritime Organisation (IMO) has recommended a method called formal safety assessment (FSA) for future development of rules and regulations. The FSA method has been applied in a pilot research project for development of risk-based rules and functional requirements for systems and components for offshore crane systems. This paper reports some developments in the project. A method for estimating target reliability for the risk-control options (safety functions) by means of the cost/benefit decision criterion has been developed in the project and is presented in this paper. Finally, a structure for risk-based rules is proposed and presented

A Critical Review of Practice of Equating the Reactivity of Spent Fuel to Fresh Fuel in Burnup Credit Criticality Safety Analyses for PWR Spent Fuel Pool Storage

International Nuclear Information System (INIS)

Wagner, J.C.; Parks, C.V.

2000-01-01

This research examines the practice of equating the reactivity of spent fuel to that of fresh fuel for the purpose of performing burnup credit criticality safety analyses for PWR spent fuel pool (SFP) storage conditions. The investigation consists of comparing k inf estimates based on reactivity equivalent fresh fuel enrichment (REFFE) to k inf estimates using the actual spent fuel isotopics. Analyses of selected storage configurations common in PWR SFPs show that this practice yields nonconservative results (on the order of a few tenths of a percent) in configurations in which the spent fuel is adjacent to higher-reactivity assemblies (e.g., fresh or lower-burned assemblies) and yields conservative results in configurations in which spent fuel is adjacent to lower-reactivity assemblies (e.g., higher-burned fuel or empty cells). When the REFFE is determined based on unborated water moderation, analyses for storage conditions with soluble boron present reveal significant nonconservative results associated with the use of the REFFE. This observation is considered to be important, especially considering the recent allowance of credit for soluble boron up to 5% in reactivity. Finally, it is shown that the practice of equating the reactivity of spent fuel to fresh fuel is acceptable, provided the conditions for which the REFFE was determined remain unchanged. Determination of the REFFE for a reference configuration and subsequent use of the REFFE for different configurations violates the basis used for the determination of the REFFE and, thus, may lead to inaccurate, and possibly, nonconservative estimates of reactivity. A significant concentration (approx. 2000 ppm) of soluble boron is typically (but not necessarily required to be) present in PWR SFPs, of which only a portion (le 500 ppm) may be credited in safety analyses. Thus, a large subcritical margin currently exists that more than accounts for errors or uncertainties associated with the use of the REFFE

Detection and analyse of hazardous roads in rural areas

DEFF Research Database (Denmark)

Sørensen, Michael

2003-01-01

For the last period of 5-10 years the notion "Grey roads" (hazardous roads) has appeared in Danish traffic safety work and improvement of these roads has become a very important part of the traffic safety work in many countries. The problem is, that the notion never has been clearly defined......, and therefore there are no unambiguos methods to point out and analyse "Grey roads". In this article based on a ph.D.-project a method to detecting "Grey roads" is introduced....

Philosophy and safety requirements for land-based nuclear installations

International Nuclear Information System (INIS)

Kellermann, Otto

1978-01-01

The main ideas of safety philosophy for land-based nuclear installations are presented together with their background of protection goals. Today's requirements for design and quality assurance are deductively shown. Finally a proposition is made for a new balancing of safety philosophy according to the high safety level that nuclear installations have reached

Systems reliability analyses and risk analyses for the licencing procedure under atomic law

International Nuclear Information System (INIS)

Berning, A.; Spindler, H.

1983-01-01

For the licencing procedure under atomic law in accordance with Article 7 AtG, the nuclear power plant as a whole needs to be assessed, plus the reliability of systems and plant components that are essential to safety are to be determined with probabilistic methods. This requirement is the consequence of safety criteria for nuclear power plants issued by the Home Department (BMI). Systems reliability studies and risk analyses used in licencing procedures under atomic law are identified. The stress is on licencing decisions, mainly for PWR-type reactors. Reactor Safety Commission (RSK) guidelines, examples of reasoning in legal proceedings and arguments put forth by objectors are also dealt with. Correlations are shown between reliability analyses made by experts and licencing decisions by means of examples. (orig./HP) [de

[Safety monitoring of cell-based medicinal products (CBMPs)].

Science.gov (United States)

Funk, Markus B; Frech, Marion; Spranger, Robert; Keller-Stanislawski, Brigitte

2015-11-01

Cell-based medicinal products (CBMPs), a category of advanced-therapy medicinal products (ATMPs), are authorised for the European market by the European Commission by means of the centralized marketing authorisation. By conforming to the German Medicinal Products Act (Sec. 4b AMG), national authorisation can be granted by the Paul-Ehrlich-Institut in Germany exclusively for ATMPs not based on a routine manufacturing procedure. In both procedures, quality, efficacy, and safety are evaluated and the risk-benefit balance is assessed. For the centralised procedure, mainly controlled clinical trial data must be submitted, whereas the requirements for national procedures could be modified corresponding to the stage of development of the ATMP. After marketing authorization, the marketing authorization/license holder is obligated to report all serious adverse reactions to the competent authority and to provide periodic safety update reports. If necessary, post-authorization safety studies could be imposed. On the basis of these regulatory measures, the safety of advanced therapies can be monitored and improved.

Collaborative, cross-national studies on health and safety in seafaring for evidence-based Maritime policy and regulations.

Science.gov (United States)

Jensen, Olaf C

2009-01-01

Until recently, maritime health and safety policies and regulations were sparsely based on health and safety research, and only a small number of countries contributed to new research. To strengthen maritime health and safety research activities by presenting a study example and discussing the possibilities and needs for more national and cross-national research. In a cross-national epidemiological study example, the seafarers from eleven countries completed small, anonymous questionnaires concerning the working conditions on their latest tours at sea while waiting for their health examinations. Significant disparities were pointed out among the nationalities, e.g., the length of the tours at sea, the proportional distribution of officers and non-officers, the mean age structure, the injury incidence rates, and the differences of occupational safety standards. The analysis of all data together increased the statistical strength of the multivariate analyses and allowed for valid comparisons among the nationalities. The questionnaire data was used successfully in the collaborative study example, but other data sources and methods are useful for health and safety research in seafaring as well. More national and cross-national research on maritime health and safety is warranted.

Status of science and technology with respect of preparation and evaluation of accident analyses and the use of analysis simulators

International Nuclear Information System (INIS)

Pointner, Winfried; Cuesta Morales, Alejandra; Draeger, Peer; Hartung, Juergen; Jakubowski, Zygmunt; Meyer, Gerhard; Palazzo, Simone; Moner, Guim Pallas; Perin, Yann; Pasichnyk, Ihor

2014-07-01

The scope of the work was to elaborate the prerequisites for short term accident analyses including recommendations for the application of new methodologies and computational procedures and technical aspects of safety evaluation. The following work packages were performed: Knowledge base for best estimate accident analyses; analytical studies on the PWR plant behavior in case of multiple safety system failures; extension and maintenance of the data base for plant specific analysis simulators.

Meta-analysis of surgical safety checklist effects on teamwork, communication, morbidity, mortality, and safety.

Science.gov (United States)

Lyons, Vanessa E; Popejoy, Lori L

2014-02-01

The purpose of this study is to examine the effectiveness of surgical safety checklists on teamwork, communication, morbidity, mortality, and compliance with safety measures through meta-analysis. Four meta-analyses were conducted on 19 studies that met the inclusion criteria. The effect size of checklists on teamwork and communication was 1.180 (p = .003), on morbidity and mortality was 0.123 (p = .003) and 0.088 (p = .001), respectively, and on compliance with safety measures was 0.268 (p teamwork and communication, reduce morbidity and mortality, and improve compliance with safety measures. This meta-analysis is limited in its generalizability based on the limited number of studies and the inclusion of only published research. Future research is needed to examine possible moderating variables for the effects of surgical safety checklists.

Preliminary safety evaluation for CSR1000 with passive safety system

International Nuclear Information System (INIS)

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

2014-01-01

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

Development of safety analysis technology for integral reactor

Energy Technology Data Exchange (ETDEWEB)

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)

Effectiveness of web-based tailored advice on parents' child safety behaviors: randomized controlled trial.

Science.gov (United States)

van Beelen, Mirjam Elisabeth Johanna; Beirens, Tinneke Monique Jozef; den Hertog, Paul; van Beeck, Eduard Ferdinand; Raat, Hein

2014-01-24

Injuries at home are a major cause of death, disability, and loss of quality of life among young children. Despite current safety education, required safety behavior of parents is often lacking. To prevent various childhood disorders, the application of Web-based tools has increased the effectiveness of health promotion efforts. Therefore, an intervention with Web-based, tailored, safety advice combined with personal counseling (E-Health4Uth home safety) was developed and applied. To evaluate the effect of E-Health4Uth home safety on parents' safety behaviors with regard to the prevention of falls, poisoning, drowning, and burns. A randomized controlled trial was conducted (2009-2011) among parents visiting well-baby clinics in the Netherlands. Parents were randomly assigned to the intervention group (E-Health4Uth home safety intervention) or to the control condition consisting of usual care. Parents in the intervention condition completed a Web-based safety behavior assessment questionnaire; the resulting tailored safety advice was discussed with their child health care professional at a well-baby visit (age approximately 11 months). Parents in the control condition received counseling using generic safety information leaflets at this well-baby visit. Parents' child safety behaviors were derived from self-report questionnaires at baseline (age 7 months) and at follow-up (age 17 months). Each specific safety behavior was classified as safe/unsafe and a total risk score was calculated. Logistic and linear regression analyses were used to reveal differences in safety behavior between the intervention and the control condition at follow-up. A total of 1292 parents (response rate 44.79%) were analyzed. At follow-up, parents in the intervention condition (n=643) showed significantly less unsafe behavior compared to parents in the control condition (n=649): top of staircase (23.91% vs. 32.19%; OR 0.65, 95% CI 0.50-0.85); bottom of staircase (63.53% vs. 71.94%; OR 0

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)

Reliability estimation of safety-critical software-based systems using Bayesian networks

International Nuclear Information System (INIS)

Helminen, A.

2001-06-01

Due to the nature of software faults and the way they cause system failures new methods are needed for the safety and reliability evaluation of software-based safety-critical automation systems in nuclear power plants. In the research project 'Programmable automation system safety integrity assessment (PASSI)', belonging to the Finnish Nuclear Safety Research Programme (FINNUS, 1999-2002), various safety assessment methods and tools for software based systems are developed and evaluated. The project is financed together by the Radiation and Nuclear Safety Authority (STUK), the Ministry of Trade and Industry (KTM) and the Technical Research Centre of Finland (VTT). In this report the applicability of Bayesian networks to the reliability estimation of software-based systems is studied. The applicability is evaluated by building Bayesian network models for the systems of interest and performing simulations for these models. In the simulations hypothetical evidence is used for defining the parameter relations and for determining the ability to compensate disparate evidence in the models. Based on the experiences from modelling and simulations we are able to conclude that Bayesian networks provide a good method for the reliability estimation of software-based systems. (orig.)

An Introduction of Behavior-Based Safety Program in Nuclear Power Plants

International Nuclear Information System (INIS)

Lee, Yong Hee; Lim, Hyeon Kyo

2011-01-01

There are many methods and approaches for a human error assessment that is valuable for investigating the causes of undesirable events and counter-plans to prevent their recurrence in the nuclear power plants (NPPs). There is behavior-based safety refers to the process of using a proactive approach to safety and health management. It either focuses on risk of behaviors that can lead to an injury, or on safe behaviors that can contribute to injury prevention. Early applications of behavior based safety included the construction and manufacturing industries, but today behavior based safety is applied to a wide variety of industries and service lines. This behavior based safety program can offer a set of significant human error countermeasures to be considered for human error in NPPs as well as other fields of industry. The current methods for the human error prevention in NPPs are several techniques such as Self-Check, Peer Check, Concurrent Verification, 3-way Communication, etc. However, it is not enough to grasp the whole human error problems in operations because the things are needed in fields are a behavior technique not a simple knowledge. Therefore, we applied a behavior based safety program on the current methods

Pathway-based analyses.

Science.gov (United States)

Kent, Jack W

2016-02-03

New technologies for acquisition of genomic data, while offering unprecedented opportunities for genetic discovery, also impose severe burdens of interpretation and penalties for multiple testing. The Pathway-based Analyses Group of the Genetic Analysis Workshop 19 (GAW19) sought reduction of multiple-testing burden through various approaches to aggregation of highdimensional data in pathways informed by prior biological knowledge. Experimental methods testedincluded the use of "synthetic pathways" (random sets of genes) to estimate power and false-positive error rate of methods applied to simulated data; data reduction via independent components analysis, single-nucleotide polymorphism (SNP)-SNP interaction, and use of gene sets to estimate genetic similarity; and general assessment of the efficacy of prior biological knowledge to reduce the dimensionality of complex genomic data. The work of this group explored several promising approaches to managing high-dimensional data, with the caveat that these methods are necessarily constrained by the quality of external bioinformatic annotation.

Handbook of methods for risk-based analyses of technical specifications

International Nuclear Information System (INIS)

Samanta, P.K.; Kim, I.S.; Mankamo, T.; Vesely, W.E.

1994-12-01

Technical Specifications (TS) requirements for nuclear power plants define the Limiting Conditions for Operation (LCOs) and Surveillance Requirements (SRs) to assure safety during operation. In general, these requirements are based on deterministic analysis and engineering judgments. Experiences with plant operation indicate that some elements of the requirements are unnecessarily restrictive, while a few may not be conducive to safety. The US Nuclear Regulatory Commission (USNRC) Office of Research has sponsored research to develop systematic risk-based methods to improve various aspects of TS requirements. This handbook summarizes these risk-based methods. The scope of the handbook includes reliability and risk-based methods for evaluating allowed outage times (AOTs), scheduled or preventive maintenance, action statements requiring shutdown where shutdown risk may be substantial, surveillance test intervals (STIs), and management of plant configurations resulting from outages of systems, or components. For each topic, the handbook summarizes analytic methods with data needs, outlines the insights to be gained, lists additional references, and gives examples of evaluations

Handbook of methods for risk-based analyses of technical specifications

Energy Technology Data Exchange (ETDEWEB)

Samanta, P.K.; Kim, I.S. [Brookhaven National Lab., Upton, NY (United States); Mankamo, T. [Avaplan Oy, Espoo (Finland); Vesely, W.E. [Science Applications International Corp., Dublin, OH (United States)

1994-12-01

Technical Specifications (TS) requirements for nuclear power plants define the Limiting Conditions for Operation (LCOs) and Surveillance Requirements (SRs) to assure safety during operation. In general, these requirements are based on deterministic analysis and engineering judgments. Experiences with plant operation indicate that some elements of the requirements are unnecessarily restrictive, while a few may not be conducive to safety. The US Nuclear Regulatory Commission (USNRC) Office of Research has sponsored research to develop systematic risk-based methods to improve various aspects of TS requirements. This handbook summarizes these risk-based methods. The scope of the handbook includes reliability and risk-based methods for evaluating allowed outage times (AOTs), scheduled or preventive maintenance, action statements requiring shutdown where shutdown risk may be substantial, surveillance test intervals (STIs), and management of plant configurations resulting from outages of systems, or components. For each topic, the handbook summarizes analytic methods with data needs, outlines the insights to be gained, lists additional references, and gives examples of evaluations.

Diversity requirements for safety critical software-based automation systems

International Nuclear Information System (INIS)

Korhonen, J.; Pulkkinen, U.; Haapanen, P.

1998-03-01

System vendors nowadays propose software-based systems even for the most critical safety functions in nuclear power plants. Due to the nature and mechanisms of influence of software faults new methods are needed for the safety and reliability evaluation of these systems. In the research project 'Programmable automation systems in nuclear power plants (OHA)' various safety assessment methods and tools for software based systems are developed and evaluated. This report first discusses the (common cause) failure mechanisms in software-based systems, then defines fault-tolerant system architectures to avoid common cause failures, then studies the various alternatives to apply diversity and their influence on system reliability. Finally, a method for the assessment of diversity is described. Other recently published reports in OHA-report series handles the statistical reliability assessment of software based (STUK-YTO-TR 119), usage models in reliability assessment of software-based systems (STUK-YTO-TR 128) and handling of programmable automation in plant PSA-studies (STUK-YTO-TR 129)

Development of SAGE, A computer code for safety assessment analyses for Korean Low-Level Radioactive Waste Disposal

International Nuclear Information System (INIS)

Zhou, W.; Kozak, Matthew W.; Park, Joowan; Kim, Changlak; Kang, Chulhyung

2002-01-01

This paper describes a computer code, called SAGE (Safety Assessment Groundwater Evaluation) to be used for evaluation of the concept for low-level waste disposal in the Republic of Korea (ROK). The conceptual model in the code is focused on releases from a gradually degrading engineered barrier system to an underlying unsaturated zone, thence to a saturated groundwater zone. Doses can be calculated for several biosphere systems including drinking contaminated groundwater, and subsequent contamination of foods, rivers, lakes, or the ocean by that groundwater. The flexibility of the code will permit both generic analyses in support of design and site development activities, and straightforward modification to permit site-specific and design-specific safety assessments of a real facility as progress is made toward implementation of a disposal site. In addition, the code has been written to easily interface with more detailed codes for specific parts of the safety assessment. In this way, the code's capabilities can be significantly expanded as needed. The code has the capability to treat input parameters either deterministic ally or probabilistic ally. Parameter input is achieved through a user-friendly Graphical User Interface.

Cyber Security Threats to Safety-Critical, Space-Based Infrastructures

Science.gov (United States)

Johnson, C. W.; Atencia Yepez, A.

2012-01-01

Space-based systems play an important role within national critical infrastructures. They are being integrated into advanced air-traffic management applications, rail signalling systems, energy distribution software etc. Unfortunately, the end users of communications, location sensing and timing applications often fail to understand that these infrastructures are vulnerable to a wide range of security threats. The following pages focus on concerns associated with potential cyber-attacks. These are important because future attacks may invalidate many of the safety assumptions that support the provision of critical space-based services. These safety assumptions are based on standard forms of hazard analysis that ignore cyber-security considerations This is a significant limitation when, for instance, security attacks can simultaneously exploit multiple vulnerabilities in a manner that would never occur without a deliberate enemy seeking to damage space based systems and ground infrastructures. We address this concern through the development of a combined safety and security risk assessment methodology. The aim is to identify attack scenarios that justify the allocation of additional design resources so that safety barriers can be strengthened to increase our resilience against security threats.

Efficacy and Safety Extrapolation Analyses for Atomoxetine in Young Children with Attention-Deficit/Hyperactivity Disorder.

Science.gov (United States)

Upadhyaya, Himanshu; Kratochvil, Christopher; Ghuman, Jaswinder; Camporeale, Angelo; Lipsius, Sarah; D'Souza, Deborah; Tanaka, Yoko

2015-12-01

This extrapolation analysis qualitatively compared the efficacy and safety profile of atomoxetine from Lilly clinical trial data in 6-7-year-old patients with attention-deficit/hyperactivity disorder (ADHD) with that of published literature in 4-5-year-old patients with ADHD (two open-label [4-5-year-old patients] and one placebo-controlled study [5-year-old patients]). The main efficacy analyses included placebo-controlled Lilly data and the placebo-controlled external study (5-year-old patients) data. The primary efficacy variables used in these studies were the ADHD Rating Scale-IV Parent Version, Investigator Administered (ADHD-RS-IV-Parent:Inv) total score, or the Swanson, Nolan and Pelham (SNAP-IV) scale score. Safety analyses included treatment-emergent adverse events (TEAEs) and vital signs. Descriptive statistics (means, percentages) are presented. Acute atomoxetine treatment improved core ADHD symptoms in both 6-7-year-old patients (n=565) and 5-year-old patients (n=37) (treatment effect: -10.16 and -7.42). In an analysis of placebo-controlled groups, the mean duration of exposure to atomoxetine was ∼ 7 weeks for 6-7-year-old patients and 9 weeks for 5-year-old patients. Decreased appetite was the most common TEAE in atomoxetine-treated patients. The TEAEs observed at a higher rate in 5-year-old versus 6-7-year-old patients were irritability (36.8% vs. 3.6%) and other mood-related events (6.9% each vs. atomoxetine may improve ADHD symptoms, but possibly to a lesser extent than in older children, with some adverse events occurring at a higher rate in 5-year-old patients.

Allowed outage time for test and maintenance - Optimization of safety

International Nuclear Information System (INIS)

Cepin, M.; Mavko, B.

1997-01-01

The main objective of the project is the development and application of methodologies for improvement and optimization of test and maintenance activities for safety related equipment in NPPs on basis of their enhanced safety. The probabilistic safety assessment serves as a base, which does not mean the replacement of the deterministic analyses but the consideration of probabilistic safety assessment results as complement to deterministic results. 15 refs, 2 figs

Studying the Safety Impact of Autonomous Vehicles Using Simulation-Based Surrogate Safety Measures

Directory of Open Access Journals (Sweden)

Mark Mario Morando

2018-01-01

Full Text Available Autonomous vehicle (AV technology has advanced rapidly in recent years with some automated features already available in vehicles on the market. AVs are expected to reduce traffic crashes as the majority of crashes are related to driver errors, fatigue, alcohol, or drugs. However, very little research has been conducted to estimate the safety impact of AVs. This paper aims to investigate the safety impacts of AVs using a simulation-based surrogate safety measure approach. To this end, safety impacts are explored through the number of conflicts extracted from the VISSIM traffic microsimulator using the Surrogate Safety Assessment Model (SSAM. Behaviours of human-driven vehicles (HVs and AVs (level 4 automation are modelled within the VISSIM’s car-following model. The safety investigation is conducted for two case studies, that is, a signalised intersection and a roundabout, under various AV penetration rates. Results suggest that AVs improve safety significantly with high penetration rates, even when they travel with shorter headways to improve road capacity and reduce delay. For the signalised intersection, AVs reduce the number of conflicts by 20% to 65% with the AV penetration rates of between 50% and 100% (statistically significant at p<0.05. For the roundabout, the number of conflicts is reduced by 29% to 64% with the 100% AV penetration rate (statistically significant at p<0.05.

Development of safety evaluation guidelines for base-isolated buildings in Japan

International Nuclear Information System (INIS)

Aoyama, Hiroyuki

1989-01-01

This paper describes the safety evaluation guidelines and the review process for non-nuclear base-isolated buildings proposed for construction in Japan. The paper discusses the guidelines application for two types of soil: hard soil and intermediate soil (soft soil was excluded.); safety evaluation items included in the level C design review; and safety margin of base isolation. Lessons learned through these design review efforts have potential applicability to design of seismic base isolation for nuclear power plants

Waste Isolation Pilot Plant Safety Analysis Report

International Nuclear Information System (INIS)

1995-11-01

The following provides a summary of the specific issues addressed in this FY-95 Annual Update as they relate to the CH TRU safety bases: Executive Summary; Site Characteristics; Principal Design and Safety Criteria; Facility Design and Operation; Hazards and Accident Analysis; Derivation of Technical Safety Requirements; Radiological and Hazardous Material Protection; Institutional Programs; Quality Assurance; and Decontamination and Decommissioning. The System Design Descriptions'' (SDDS) for the WIPP were reviewed and incorporated into Chapter 3, Principal Design and Safety Criteria and Chapter 4, Facility Design and Operation. This provides the most currently available final engineering design information on waste emplacement operations throughout the disposal phase up to the point of permanent closure. Also, the criteria which define the TRU waste to be accepted for disposal at the WIPP facility were summarized in Chapter 3 based on the WAC for the Waste Isolation Pilot Plant.'' This Safety Analysis Report (SAR) documents the safety analyses that develop and evaluate the adequacy of the Waste Isolation Pilot Plant Contact-Handled Transuranic Wastes (WIPP CH TRU) safety bases necessary to ensure the safety of workers, the public and the environment from the hazards posed by WIPP waste handling and emplacement operations during the disposal phase and hazards associated with the decommissioning and decontamination phase. The analyses of the hazards associated with the long-term (10,000 year) disposal of TRU and TRU mixed waste, and demonstration of compliance with the requirements of 40 CFR 191, Subpart B and 40 CFR 268.6 will be addressed in detail in the WIPP Final Certification Application scheduled for submittal in October 1996 (40 CFR 191) and the No-Migration Variance Petition (40 CFR 268.6) scheduled for submittal in June 1996. Section 5.4, Long-Term Waste Isolation Assessment summarizes the current status of the assessment

Waste Isolation Pilot Plant Safety Analysis Report

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-11-01

The following provides a summary of the specific issues addressed in this FY-95 Annual Update as they relate to the CH TRU safety bases: Executive Summary; Site Characteristics; Principal Design and Safety Criteria; Facility Design and Operation; Hazards and Accident Analysis; Derivation of Technical Safety Requirements; Radiological and Hazardous Material Protection; Institutional Programs; Quality Assurance; and Decontamination and Decommissioning. The System Design Descriptions`` (SDDS) for the WIPP were reviewed and incorporated into Chapter 3, Principal Design and Safety Criteria and Chapter 4, Facility Design and Operation. This provides the most currently available final engineering design information on waste emplacement operations throughout the disposal phase up to the point of permanent closure. Also, the criteria which define the TRU waste to be accepted for disposal at the WIPP facility were summarized in Chapter 3 based on the WAC for the Waste Isolation Pilot Plant.`` This Safety Analysis Report (SAR) documents the safety analyses that develop and evaluate the adequacy of the Waste Isolation Pilot Plant Contact-Handled Transuranic Wastes (WIPP CH TRU) safety bases necessary to ensure the safety of workers, the public and the environment from the hazards posed by WIPP waste handling and emplacement operations during the disposal phase and hazards associated with the decommissioning and decontamination phase. The analyses of the hazards associated with the long-term (10,000 year) disposal of TRU and TRU mixed waste, and demonstration of compliance with the requirements of 40 CFR 191, Subpart B and 40 CFR 268.6 will be addressed in detail in the WIPP Final Certification Application scheduled for submittal in October 1996 (40 CFR 191) and the No-Migration Variance Petition (40 CFR 268.6) scheduled for submittal in June 1996. Section 5.4, Long-Term Waste Isolation Assessment summarizes the current status of the assessment.

Some Examples of Accident Analyses for RB Reactor

International Nuclear Information System (INIS)

Pesic, M.

2002-01-01

The RB reactor is heavy water critical assembly operated in the Vinca Institute of Nuclear Sciences, Belgrade, Yugoslavia, since April 1959. The first Safety Analysis Report of the RB critical assembly was prepared in 1961/62. But, the first accidental analysis was done in late 1958 in aim the examine power transient and total equivalent doses received by the staff during the reactivity accident occurred on October 15, 1958. Since 1960, the RB reactor is modified few times. Beside initial natural uranium metal fuel rods, new fuel (TVR-S types) from 2% enriched metal uranium and 80% enriched UO 2 were available since 1962 and 1976, respectively. Also, modifications in control and safety systems of the reactor were done occasionally. Special reactor cores were created using all three types of fuel elements, among them, the coupled fast-thermal ones. Nuclear Safety Committee of the Vinca Institute, an independent regulatory body approved for usage all these modifications of the RB reactor. For those decisions of the Committee, the Preliminary Safety Analysis Reports were prepared that, beside proposed technical modifications and new regulation rules had included analyses of various possible accidents. Special attention is given and new methodology was proposed for thoroughly analyses of design based accidents related to coupled fast-thermal cores, that include reactor central zones filled by fuel elements without moderator. In these accidents, during assumed flooding of the fast zone by moderator, a very high reactivity could be inserted in the system with very high reactivity rate. It was necessary to provide that the safety system of the reactor had fast response to that accident and had enough high (negative) reactivity to shut down the reactor timely. In this paper, a brief overview of some accidents, methodology and computation tools used for the accident analyses at RB reactor are given. (author)

Industrial Personal Computer based Display for Nuclear Safety System

International Nuclear Information System (INIS)

Kim, Ji Hyeon; Kim, Aram; Jo, Jung Hee; Kim, Ki Beom; Cheon, Sung Hyun; Cho, Joo Hyun; Sohn, Se Do; Baek, Seung Min

2014-01-01

The safety display of nuclear system has been classified as important to safety (SIL:Safety Integrity Level 3). These days the regulatory agencies are imposing more strict safety requirements for digital safety display system. To satisfy these requirements, it is necessary to develop a safety-critical (SIL 4) grade safety display system. This paper proposes industrial personal computer based safety display system with safety grade operating system and safety grade display methods. The description consists of three parts, the background, the safety requirements and the proposed safety display system design. The hardware platform is designed using commercially available off-the-shelf processor board with back plane bus. The operating system is customized for nuclear safety display application. The display unit is designed adopting two improvement features, i.e., one is to provide two separate processors for main computer and display device using serial communication, and the other is to use Digital Visual Interface between main computer and display device. In this case the main computer uses minimized graphic functions for safety display. The display design is at the conceptual phase, and there are several open areas to be concreted for a solid system. The main purpose of this paper is to describe and suggest a methodology to develop a safety-critical display system and the descriptions are focused on the safety requirement point of view

Industrial Personal Computer based Display for Nuclear Safety System

Energy Technology Data Exchange (ETDEWEB)

Kim, Ji Hyeon; Kim, Aram; Jo, Jung Hee; Kim, Ki Beom; Cheon, Sung Hyun; Cho, Joo Hyun; Sohn, Se Do; Baek, Seung Min [KEPCO, Youngin (Korea, Republic of)

2014-08-15

The safety display of nuclear system has been classified as important to safety (SIL:Safety Integrity Level 3). These days the regulatory agencies are imposing more strict safety requirements for digital safety display system. To satisfy these requirements, it is necessary to develop a safety-critical (SIL 4) grade safety display system. This paper proposes industrial personal computer based safety display system with safety grade operating system and safety grade display methods. The description consists of three parts, the background, the safety requirements and the proposed safety display system design. The hardware platform is designed using commercially available off-the-shelf processor board with back plane bus. The operating system is customized for nuclear safety display application. The display unit is designed adopting two improvement features, i.e., one is to provide two separate processors for main computer and display device using serial communication, and the other is to use Digital Visual Interface between main computer and display device. In this case the main computer uses minimized graphic functions for safety display. The display design is at the conceptual phase, and there are several open areas to be concreted for a solid system. The main purpose of this paper is to describe and suggest a methodology to develop a safety-critical display system and the descriptions are focused on the safety requirement point of view.

Bayesian Network Assessment Method for Civil Aviation Safety Based on Flight Delays

OpenAIRE

Huawei Wang; Jun Gao

2013-01-01

Flight delays and safety are the principal contradictions in the sound development of civil aviation. Flight delays often come up and induce civil aviation safety risk simultaneously. Based on flight delays, the random characteristics of civil aviation safety risk are analyzed. Flight delays have been deemed to a potential safety hazard. The change rules and characteristics of civil aviation safety risk based on flight delays have been analyzed. Bayesian networks (BN) have been used to build ...

Thermal hydraulic and safety analyses for Pakistan Research Reactor-1

International Nuclear Information System (INIS)

Bokhari, I.H.; Israr, M.; Pervez, S.

1999-01-01

Thermal hydraulic and safety analysis of Pakistan Research Reactor-1 (PARR-1) utilizing low enriched uranium (LEU) fuel have been performed using computer code PARET. The present core comprises of 29 standard and 5 control fuel elements. Results of the thermal hydraulic analysis show that the core can be operated at a steady-state power level of 10 MW for a flow rate of 950 m 3 /h, with sufficient safety margins against ONB (onset of nucleate boiling) and DNB (departure from nucleate boiling). Safety analysis has been carried out for various modes of reactivity insertions. The events studied include: start-up accident; accidental drop of a fuel element in the core; flooding of a beam tube with water; removal of an in-pile experiment during reactor operation etc. For each of these transients, time histories of reactor power, energy released and clad surface temperature etc. were calculated. The results indicate that the peak clad temperatures remain well below the clad melting temperature during these accidents. It is therefore concluded that the reactor can be safely operated at 10 MW without compromising safety. (author)

Exploring the role of emotional intelligence in behavior-based safety coaching.

Science.gov (United States)

Wiegand, Douglas M

2007-01-01

Safety coaching is an applied behavior analysis technique that involves interpersonal interaction to understand and manipulate environmental conditions that are directing (i.e., antecedent to) and motivating (i.e., consequences of) safety-related behavior. A safety coach must be skilled in interacting with others so as to understand their perspectives, communicate a point clearly, and be persuasive with behavior-based feedback. This article discusses the evidence-based "ability model" of emotional intelligence and its relevance to the interpersonal aspect of the safety coaching process. Emotional intelligence has potential for improving safety-related efforts and other aspects of individuals' work and personal lives. Safety researchers and practitioners are therefore encouraged to gain an understanding of emotional intelligence and conduct and support research applying this construct toward injury prevention.

Systematic safety evaluation of old nuclear power plants

International Nuclear Information System (INIS)

Dredemis, G.; Fourest, B.

1984-01-01

The French safety authorities have undertaken a systematic evaluation of the safety of old nuclear power plants. Apart from a complete revision of safety documents (safety analysis report, general operating rules, incident and accident procedures, internal emergency plan, quality organisation manual), this examination consisted of analysing the operating experience of systems frequently challenged and a systematic examination of the safety-related systems. This paper is based on an exercise at the Ardennes Nuclear Power Plant which has been in operation for 15 years. This paper also summarizes the main surveys and modifications relating to this power plant. (orig.)

Behavioral based safety approaches

International Nuclear Information System (INIS)

Maria Michael Raj, I.

2009-01-01

Approach towards the establishment of positive safety culture at Heavy Water Plant, Tuticorin includes the adoption of several important methodologies focused on human behavior and culminates with achievement of Total Safety Culture where Quality and Productivity are integrated with Safety

Analysis of School Food Safety Programs Based on HACCP Principles

Science.gov (United States)

Roberts, Kevin R.; Sauer, Kevin; Sneed, Jeannie; Kwon, Junehee; Olds, David; Cole, Kerri; Shanklin, Carol

2014-01-01

Purpose/Objectives: The purpose of this study was to determine how school districts have implemented food safety programs based on HACCP principles. Specific objectives included: (1) Evaluate how schools are implementing components of food safety programs; and (2) Determine foodservice employees food-handling practices related to food safety.…

Needs for evidence-based road safety decision making in Europe.

NARCIS (Netherlands)

Dupont, E. Muhlrad, N. Buttler, I. Gitelman, V. Giustiniani, G. Jähi, H. Machata, K. Martensen, H. Papadimitriou, E. Persia, L. Talbot, R. Vallet, G. Wijnen, W. & Yannis, G.

2012-01-01

The objective of this research is the assessment of current needs for evidence-based road safety decision making in Europe, through the consultation of a panel of road safety experts. The members of this Experts Panel have extensive knowledge of road safety management processes and needs in their

Hazard analysis & safety requirements for small drone operations : to what extent do popular drones embed safety?

NARCIS (Netherlands)

Plioutsias, Anastasios; Karanikas, Nektarios; Chatzimichailidou, Maria Mikela

2018-01-01

Currently, published risk analyses for drones refer mainly to commercial systems, use data from civil aviation, and are based on probabilistic approaches without suggesting an inclusive list of hazards and respective requirements. Within this context, this paper presents: (1) a set of safety

A concurrent diagnosis of microbiological food safety output and food safety management system performance: Cases from meat processing industries

NARCIS (Netherlands)

Luning, P.A.; Jacxsens, L.; Rovira, J.; Oses Gomez, S.; Uyttendaele, M.; Marcelis, W.J.

2011-01-01

Stakeholder requirements force companies to analyse their food safety management system (FSMS) performance to improve food safety. Performance is commonly analysed by checking compliance against preset requirements via audits/inspections, or actual food safety (FS) output is analysed by

Reactor safety research and safety technology. Pt. 2

International Nuclear Information System (INIS)

Theenhaus, R.; Wolters, J.

1987-01-01

The state of HTR safety research work reached permits a comprehensive and reliable answer to be given to questions which have been raised by the reactor accident at Chernobyl, regarding HTR safety. Together with the probability safety analyses, the way to a safety concept suitable for an HTR is cleared; instructions are given for design optimisation with regard to safety technique and economy. The consequences of a graphite fire, the neutron physics design and the consequenes of the lack of a safety containment are briefly described. (DG) [de

Safety issues of tooth whitening using peroxide-based materials.

Science.gov (United States)

Li, Y; Greenwall, L

2013-07-01

In-office tooth whitening using hydrogen peroxide (H₂O₂) has been practised in dentistry without significant safety concerns for more than a century. While few disputes exist regarding the efficacy of peroxide-based at-home whitening since its first introduction in 1989, its safety has been the cause of controversy and concern. This article reviews and discusses safety issues of tooth whitening using peroxide-based materials, including biological properties and toxicology of H₂O₂, use of chlorine dioxide, safety studies on tooth whitening, and clinical considerations of its use. Data accumulated during the last two decades demonstrate that, when used properly, peroxide-based tooth whitening is safe and effective. The most commonly seen side effects are tooth sensitivity and gingival irritation, which are usually mild to moderate and transient. So far there is no evidence of significant health risks associated with tooth whitening; however, potential adverse effects can occur with inappropriate application, abuse, or the use of inappropriate whitening products. With the knowledge on peroxide-based whitening materials and the recognition of potential adverse effects associated with the procedure, dental professionals are able to formulate an effective and safe tooth whitening regimen for individual patients to achieve maximal benefits while minimising potential risks.

Pedestrian safety management using the risk-based approach

Directory of Open Access Journals (Sweden)

Romanowska Aleksandra

2017-01-01

Full Text Available The paper presents a concept of a multi-level pedestrian safety management system. Three management levels are distinguished: strategic, tactical and operational. The basis for the proposed approach to pedestrian safety management is a risk-based method. In the approach the elements of behavioural and systemic theories were used, allowing for the development of a formalised and repeatable procedure integrating the phases of risk assessment and response to the hazards of road crashes involving pedestrians. Key to the method are tools supporting pedestrian safety management. According to the risk management approach, the tools can be divided into two groups: tools supporting risk assessment and tools supporting risk response. In the paper attention is paid to selected tools supporting risk assessment, with particular emphasis on the methods for estimating forecasted pedestrian safety measures (at strategic, national and regional level and identification of particularly dangerous locations in terms of pedestrian safety at tactical (regional and local and operational level. The proposed pedestrian safety management methods and tools can support road administration in making rational decisions in terms of road safety, safety of road infrastructure, crash elimination measures or reducing the consequences suffered by road users (particularly pedestrians as a result of road crashes.

Development of FPGA-based safety-related I and C systems

Energy Technology Data Exchange (ETDEWEB)

Goto, Y.; Oda, N.; Miyazaki, T.; Hayashi, T.; Sato, T.; Igawa, S. [08, Shinsugita-cho, Isogo-ku, Yokohama 235-8523 (Japan); 1, Toshiba-cho, Fuchu, Tokyo 183-8511 (Japan)

2006-07-01

Toshiba has developed Non-rewritable (NRW) Field Programmable Gate Array (FPGA)-based safety-related Instrumentation and Control (I and C) system [1]. Considering application to safety-related systems, nonvolatile and non-rewritable FPGA which is impossible to be changed after once manufactured has been adopted in Toshiba FPGA-based system. FPGA is a device which consists only of defined digital circuit: hardware, which performs defined processing. FPGA-based system solves issues existing both in the conventional systems operated by analog circuits (analog-based system) and the systems operated by central processing unit (CPU-based system). The advantages of applying FPGA are to keep the long-life supply of products, improving testability (verification), and to reduce the drift which may occur in analog-based system. The system which Toshiba developed this time is Power Range Monitor (PRM). Toshiba is planning to expand application of FPGA-based technology by adopting this development method to the other safety-related systems from now on. (authors)

Applying interprofessional Team-Based Learning in patient safety: a pilot evaluation study.

Science.gov (United States)

Lochner, Lukas; Girardi, Sandra; Pavcovich, Alessandra; Meier, Horand; Mantovan, Franco; Ausserhofer, Dietmar

2018-03-27

Interprofessional education (IPE) interventions are not always successful in achieving learning outcomes. Team-Based Learning (TBL) would appear to be a suitable pedagogical method for IPE, as it focuses on team performance; however, little is known about interprofessional TBL as an instructional framework for patient safety. In this pilot-study, we aimed to (1) describe participants' reactions to TBL, (2) observe their achievement with respect to interprofessional education learning objectives, and (3) document their attitudinal shifts with regard to patient safety behaviours. We developed and implemented a three-day course for pre-qualifying, non-medical healthcare students to give instruction on non-technical skills related to 'learning from errors'. The course consisted of three sequential modules: 'Recognizing Errors', 'Analysing Errors', and 'Reporting Errors'. The evaluation took place within a quasi-experimental pre-test-post-test study design. Participants completed self-assessments through valid and reliable instruments such as the Mennenga's TBL Student Assessment Instrument and the University of the West of England's Interprofessional Questionnaire. The mean scores of the individual readiness assurance tests were compared with the scores of the group readiness assurance test in order to explore if students learned from each other during group discussions. Data was analysed using descriptive (i.e. mean, standard deviation), parametric (i.e. paired t-test), and non-parametric (i.e. Wilcoxon signed-rank test) methods. Thirty-nine students from five different bachelor's programs attended the course. The participants positively rated TBL as an instructional approach. All teams outperformed the mean score of their individual members during the readiness assurance process. We observed significant improvements in 'communication and teamwork' and 'interprofessional learning' but not in 'interprofessional interaction' and 'interprofessional relationships

School-Based and Community-Based Gun Safety Educational Strategies for Injury Prevention.

Science.gov (United States)

Holly, Cheryl; Porter, Sallie; Kamienski, Mary; Lim, Aubrianne

2018-05-01

Nearly 1,300 children in the United States die because of firearm-related injury each year and another 5,790 survive gunshot wounds, making the prevention of firearm-related unintentional injury to children of vital importance to families, health professionals, and policy makers. To systematically review the evidence on school-based and community-based gun safety programs for children aged 3 to 18 years. Systematic review. Twelve databases were searched from their earliest records to December 2016. Interventional and analytic studies were sought, including randomized controlled trials, quasi-experimental studies, as well as before-and-after studies or cohort studies with or without a control that involved an intervention. The low level of evidence, heterogeneity of studies, and lack of consistent outcome measures precluded a pooled estimate of results. A best evidence synthesis was performed. Results support the premise that programs using either knowledge-based or active learning strategies or a combination of these may be insufficient for teaching gun safety skills to children. Gun safety programs do not improve the likelihood that children will not handle firearms in an unsupervised situation. Stronger research designs with larger samples are needed to determine the most effective way to transfer the use of the gun safety skills outside the training session and enable stronger conclusions to be drawn.

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.

Interim summary report of the safety case 2009

International Nuclear Information System (INIS)

2010-03-01

Following the guidelines set forth by the Ministry of Trade and Industry (now Ministry of Employment and Economy), Posiva is preparing to submit a construction license application for the final disposal spent nuclear fuel at the Olkiluoto site, Finland, by the end of the year 2012. Disposal will take place in a geological repository implemented according to the KBS-3 method. The long-term safety section supporting the license application will be based on a safety case that, according to the internationally adopted definition, will be a compilation of the evidence, analyses and arguments that quantify and substantiate the safety and the level of expert confidence in the safety of the planned repository. The present Interim Summary Report represents a major contribution to the development of this safety case. The report has been compiled in accordance with Posiva's current plan for preparing this safety case. A full safety case for the KBS-3V variant will be developed to support the Preliminary Safety Assessment Report (PSAR) in 2012. The report outlines the current design and safety concept for the planned repository. It summarises the approach used to formulate scenarios for the evolution of the disposal system over time, describes these scenarios and presents the main models and computer codes used to analyse them. It also discusses compliance with Finnish regulatory requirements for long-term safety of a geological repository and gives the main evidence, arguments and analyses that lead to confidence, on the part of Posiva, in the long-term safety of the planned repository. Current understanding of the evolution of the disposal system indicates that, except a few unlikely circumstances affecting a small number of canisters, spent fuel will remain isolated, and the radionuclides contained within the canisters, for hundreds of thousands of years or more, in accordance with the base scenario. Confidence in this base scenario derives, in the first place, from the

Optimization of a Centrifugal Boiler Circulating Pump's Casing Based on CFD and FEM Analyses

Directory of Open Access Journals (Sweden)

Zhigang Zuo

2014-04-01

Full Text Available It is important to evaluate the economic efficiency of boiler circulating pumps in manufacturing process from the manufacturers' point of view. The possibility of optimizing the pump casing with respect to structural pressure integrity and hydraulic performance was discussed. CFD analyses of pump models with different pump casing sizes were firstly carried out for the hydraulic performance evaluation. The effects of the working temperature and the sealing ring on the hydraulic efficiency were discussed. A model with casing diameter of 0.875D40 was selected for further analyses. FEM analyses were then carried out on different combinations of casing sizes, casing wall thickness, and materials, to evaluate its safety related to pressure integrity, with respect to both static and fatigue strength analyses. Two models with forging and cast materials were selected as final results.

Experience with performance based training of nuclear criticality safety engineers

International Nuclear Information System (INIS)

Taylor, R.G.

1993-01-01

For non-reactor nuclear facilities, the U.S. Department of Energy (DOE) does not require that nuclear criticality safety engineers demonstrate qualification for their job. It is likely, however, that more formalism will be required in the future. Current DOE requirements for those positions which do have to demonstrate qualification indicate that qualification should be achieved by using a systematic approach such as performance based training (PBT). Assuming that PBT would be an acceptable mechanism for nuclear criticality safety engineer training in a more formal environment, a site-specific analysis of the nuclear criticality safety engineer job was performed. Based on this analysis, classes are being developed and delivered to a target audience of newer nuclear criticality safety engineers. Because current interest is in developing training for selected aspects of the nuclear criticality safety engineer job, the analysis is incompletely developed in some areas

Safety analyses for transient behavior of plasma and in-vessel components during plasma abnormal events in fusion reactor

International Nuclear Information System (INIS)

Honda, Takuro; Okazaki, Takashi; Bartels, H.W.; Uckan, N.A.; Seki, Yasushi.

1997-01-01

Safety analyses on plasma abnormal events have been performed using a hybrid code of a plasma dynamics model and a heat transfer model of in-vessel components. Several abnormal events, e.g., increase in fueling rate, were selected for the International Thermonuclear Experimental Reactor (ITER) and transient behavior of the plasma and the invessel components during the events was analyzed. The physics model for safety analysis was conservatively prepared. In most cases, the plasma is terminated by a disruption or it returns to the original operation point. When the energy confinement improves by a factor of 2.0 in the steady state, which is a hypothetical assumption under the present plasma data, the maximum fusion power reaches about 3.3 GW at about 3.6 s and the plasma is terminated due to a disruption. However, the results obtained in this study show the confinement boundary of ITER can be kept almost intact during the abnormal plasma transients, as long as the cooling system works normally. Several parametric studies are needed to comprehend the overpower transient including structure behavior, since many uncertainties are connected to the filed of the plasma physics. And, future work will need to discuss the burn control scenario considering confinement mode transition, system specifications, experimental plans and safety regulations, etc. to confirm the safety related to the plasma anomaly. (author)

Risk based maintenance to increase safety and decrease costs

International Nuclear Information System (INIS)

Phillips, J.H.

2000-01-01

Risk-Based techniques have been developed for commercial nuclear power plants for the last eight years by a team working through the ASME Center for Research and Technology Development (CRTD). System boundaries and success criteria is defined using the Probabilistic Risk Analysis or Probabilistic Safety Analysis developed to meet the Individual Plant Evaluation. Final ranking of components is by a plant expert panel similar to the one developed for the Maintenance Rule. Components are identified as being high risk-significant or low risk-significant. Maintenance and resources are focused on those components that have the highest risk-significance. The techniques have been developed and applied at a number of plants. Results from the first risk-based inspection pilot plant indicates safety due to pipe failure can be doubled while the inspection reduced to about 80% when compared with current inspection programs. Pilot studies on risk-based testing indicate that about 60% of pumps and 25 to 30% of valves in plants are high safety-significant The reduction in inspection and testing reduces the person-rem exposure and resulting in further increases in safety. These techniques have been documented in publications by the ASME CRTD which are referenced. (author)

Thermal and stress analyses of meltdown cups for LMFBR safety experiments using SLSF in-reactor loops

Energy Technology Data Exchange (ETDEWEB)

Blomquist, C. A. [Argonne National Lab., IL (United States); Ariman, T. [Univ. of Notre Dame, IN (United States); Pierce, R. D.; Pedersen, D. R. [Argonne National Lab., IL (United States)

1977-07-01

The test trains for the Sodium Loop Safety Facility (SLSF) in-reactor experiments, which simulate hypothetical LMFBR accidents, have a meltdown cup to protect the primary containment from the effects of molten materials. Thermal and stress analyses were performed on the cup which is designed to contain 3.6 kg of molten fuel and 2.4 kg of molten steel. The cup principal components are: 1. A 38 mm diameter tungsten spike which provides initial fuel quenching and prevents fuel boiling, 2. A 73 mm inside diameter tungsten liner to isolate the support vessel from the molten material high initial temperature, 3. An insulator which is an expedient for extending the experiment time, and 4. An Inconel 625 vessel which provides the structural support to withstand the thermal and pressure stresses. The spike, liner, and insulator are supported by a hemispherical tungsten end cap which fits inside the hemispherical bottom of the support vessel. This vessel is attached to the 316 stainless steel test train with an Inconel 750 wire-formed retaining ring. Thermal analyses were performed with the Argonne-modified version of the general heat transfer code THTB, based on the instantaneous addition of 3200/sup 0/K molten fuel with a decay heat of 9 W/gm and 1920/sup 0/K molten steel. These analyses have shown that the cup will adequately cool the molten materials. The maximum temperature occurs at the center of the fuel region but it is always less than the fuel boiling point. The maximum temperature occurs at the center of the fuel region but it is always less than the fuel boiling point. The most severe heating occurs when there is no sodium flow outside the cup. For this case the sodium boils (approximately 1200/sup 0/K) and the Inconel vessel and tungsten liner temperatures are approximately 1250/sup 0/K and 2420/sup 0/K, respectively.

Applications of computer based safety systems in Korea nuclear power plants

International Nuclear Information System (INIS)

Won Young Yun

1998-01-01

With the progress of computer technology, the applications of computer based safety systems in Korea nuclear power plants have increased rapidly in recent decades. The main purpose of this movement is to take advantage of modern computer technology so as to improve the operability and maintainability of the plants. However, in fact there have been a lot of controversies on computer based systems' safety between the regulatory body and nuclear utility in Korea. The Korea Institute of Nuclear Safety (KINS), technical support organization for nuclear plant licensing, is currently confronted with the pressure to set up well defined domestic regulatory requirements from this aspect. This paper presents the current status and the regulatory activities related to the applications of computer based safety systems in Korea. (author)

Safety balance: Analysis of safety systems

International Nuclear Information System (INIS)

Delage, M.; Giroux, C.

1990-12-01

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

Model-based testing for software safety

NARCIS (Netherlands)

Gurbuz, Havva Gulay; Tekinerdogan, Bedir

2017-01-01

Testing safety-critical systems is crucial since a failure or malfunction may result in death or serious injuries to people, equipment, or environment. An important challenge in testing is the derivation of test cases that can identify the potential faults. Model-based testing adopts models of a

Coupling of channel thermalhydraulics and fuel behaviour in ACR-1000 safety analyses

International Nuclear Information System (INIS)

Huang, F.L.; Lei, Q.M.; Zhu, W.; Bilanovic, Z.

2008-01-01

Channel thermalhydraulics and fuel thermal-mechanical behaviour are interlinked. This paper describes a channel thermalhydraulics and fuel behaviour coupling methodology that has been used in ACR-1000 safety analyses. The coupling is done for all 12 fuel bundles in a fuel channel using the channel thermalhydraulics code CATHENA MOD-3.5d/Rev 2 and the transient fuel behaviour code ELOCA 2.2. The coupling approach can be used for every fuel element or every group of fuel elements in the channel. Test cases are presented where a total of 108 fuel element models are set up to allow a full coupling between channel thermalhydraulics and detailed fuel analysis for a channel containing a string of 12 fuel bundles. An additional advantage of this coupling approach is that there is no need for a separate detailed fuel analysis because the coupling analysis, once done, provides detailed calculations for the fuel channel (fuel bundles, pressure tube, and calandria tube) as well as all the fuel elements (or element groups) in the channel. (author)

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?

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)

The Commodity Form of Safety Information

Directory of Open Access Journals (Sweden)

Rodrigo Finkelstein

2015-10-01

Full Text Available The production of safety information is deemed a vital resource to protect human lives at the work site. The injury rate, lost days, incapacity rate, and fatality rate, are key indicators to prop up labour risk awareness and identify job hazards. However, safety information gets highly distorted because it does not only measure risk but serves as a means of exchange. It determines the amount of money to be swapped between Workers’ Compensation Boards and their client corporations. Moreover, as a depository of exchange value, safety information tends to exert pressure over social reality rather than just being a passive reflection of it. This paper discloses the commodity form of safety information. Based on a political economy of information framework, it identifies, describes, and analyses the safety information commodity in its active role of organizing safety and labour health.

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.)

Management by process based systems and safety focus

International Nuclear Information System (INIS)

Rydnert, Bo; Groenlund, Bjoern

2005-12-01

An initiative from The Swedish Nuclear Power Inspectorate led to this study carried out in the late autumn of 2005. The objective was to understand in more detail how an increasing use of process management affects organisations, on the one hand regarding risks and security, on the other hand regarding management by objectives and other management and operative effects. The main method was interviewing representatives of companies and independent experts. More than 20 interviews were carried out. In addition a literature study was made. All participating companies are using Management Systems based on processes. However, the methods chosen, and the results achieved, vary extensively. Thus, there are surprisingly few examples of complete and effective management by processes. Yet there is no doubt that management by processes is effective and efficient. Overall goals are reached, business results are achieved in more reliable ways and customers are more satisfied. The weaknesses found can be translated into a few comprehensive recommendations. A clear, structured and acknowledged model should be used and the processes should be described unambiguously. The changed management roles should be described and obeyed extremely legibly. New types of process objectives need to be formulated. In addition one fact needs to be observed and effectively fended off. Changes are often met by mental opposition on management level, as well as among co-workers. This fact needs attention and leadership. Safety development is closely related to the design and operation of a business management system and its continual improvement. A deep understanding of what constitutes an efficient and effective management system affects the understanding of safety. safety culture and abilities to achieve safety goals. Concerning risk, the opinions were unambiguous. Management by processes as such does not result in any further risks. On the contrary. Processes give a clear view of production and

Evidence-based and data-driven road safety management

Directory of Open Access Journals (Sweden)

Fred Wegman

2015-07-01

Full Text Available Over the past decades, road safety in highly-motorised countries has made significant progress. Although we have a fair understanding of the reasons for this progress, we don't have conclusive evidence for this. A new generation of road safety management approaches has entered road safety, starting when countries decided to guide themselves by setting quantitative targets (e.g. 50% less casualties in ten years' time. Setting realistic targets, designing strategies and action plans to achieve these targets and monitoring progress have resulted in more scientific research to support decision-making on these topics. Three subjects are key in this new approach of evidence-based and data-driven road safety management: ex-post and ex-ante evaluation of both individual interventions and intervention packages in road safety strategies, and transferability (external validity of the research results. In this article, we explore these subjects based on recent experiences in four jurisdictions (Western Australia, the Netherlands, Sweden and Switzerland. All four apply similar approaches and tools; differences are considered marginal. It is concluded that policy-making and political decisions were influenced to a great extent by the results of analysis and research. Nevertheless, to compensate for a relatively weak theoretical basis and to improve the power of this new approach, a number of issues will need further research. This includes ex-post and ex-ante evaluation, a better understanding of extrapolation of historical trends and the transferability of research results. This new approach cannot be realized without high-quality road safety data. Good data and knowledge are indispensable for this new and very promising approach.

Training the Masses ? Web-based Laser Safety Training at LLNL

Energy Technology Data Exchange (ETDEWEB)

Sprague, D D

2004-12-17

The LLNL work smart standard requires us to provide ongoing laser safety training for a large number of persons on a three-year cycle. In order to meet the standard, it was necessary to find a cost and performance effective method to perform this training. This paper discusses the scope of the training problem, specific LLNL training needs, various training methods used at LLNL, the advantages and disadvantages of these methods and the rationale for selecting web-based laser safety training. The tools and costs involved in developing web-based training courses are also discussed, in addition to conclusions drawn from our training operating experience. The ILSC lecture presentation contains a short demonstration of the LLNL web-based laser safety-training course.

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

Safety Cultures in Water-Based Outdoor Activities in Denmark

DEFF Research Database (Denmark)

Andkjær, Søren; Arvidsen, Jan

2015-01-01

In this paper, we report on the study Safe in Nature (Tryg i naturen) in which the aim was to analyze and discuss risk and safety related to outdoor recreation in the coastal regions of Denmark. A cultural perspective is applied to risk management and the safety cultures related to three selected...... water-based outdoor activities: small boat fishing, sea kayaking, and kite surfing. The theoretical framework used was cultural analysis and the methodological approach was mixed methods using case studies with survey and qualitative interviews. The study indicates that safety is a complex matter...... and that safety culture can be understood as the sum and interaction among six categories. The safety culture is closely related to the activity and differs widely among activities. We suggest a broad perspective be taken on risk management wherein risk and safety can be managed at different levels. Small boat...

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

Science.gov (United States)

Schweizer, Anja; Dejager, Sylvie; Foley, James E; Kothny, Wolfgang

2011-01-01

Aim: 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. PMID:21415917

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

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

Energy Technology Data Exchange (ETDEWEB)

NONE

2014-12-15

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

GIS-based Approaches to Catchment Area Analyses of Mass Transit

DEFF Research Database (Denmark)

Andersen, Jonas Lohmann Elkjær; Landex, Alex

2009-01-01

Catchment area analyses of stops or stations are used to investigate potential number of travelers to public transportation. These analyses are considered a strong decision tool in the planning process of mass transit especially railroads. Catchment area analyses are GIS-based buffer and overlay...... analyses with different approaches depending on the desired level of detail. A simple but straightforward approach to implement is the Circular Buffer Approach where catchment areas are circular. A more detailed approach is the Service Area Approach where catchment areas are determined by a street network...... search to simulate the actual walking distances. A refinement of the Service Area Approach is to implement additional time resistance in the network search to simulate obstacles in the walking environment. This paper reviews and compares the different GIS-based catchment area approaches, their level...

Experience with performance based training of nuclear criticality safety engineers

International Nuclear Information System (INIS)

Taylor, R.G.

1993-01-01

Historically, new entrants to the practice of nuclear criticality safety have learned their job primarily by on-the-job training (OJT) often by association with an experienced nuclear criticality safety engineer who probably also learned their job by OJT. Typically, the new entrant learned what he/she needed to know to solve a particular problem and accumulated experience as more problems were solved. It is likely that more formalism will be required in the future. Current US Department of Energy requirements for those positions which have to demonstrate qualification indicate that it should be achieved by using a systematic approach such as performance based training (PBT). Assuming that PBT would be an acceptable mechanism for nuclear criticality safety engineer training in a more formal environment, a site-specific analysis of the nuclear criticality safety engineer job was performed. Based on this analysis, classes are being developed and delivered to a target audience of newer nuclear criticality safety engineers. Because current interest is in developing training for selected aspects of the nuclear criticality safety engineer job, the analysis i's incompletely developed in some areas. Details of this analysis are provided in this report

International validation of safety analyses for nuclear power plants; Mednarodno preverjanje varnostnih analiz za jedrske elektrane

Energy Technology Data Exchange (ETDEWEB)

Gregoric, N; Mavko, B [Institut ' Jozef Stefan' Ljubljana (Yugoslavia)

1988-07-01

Paper describes the participation of 'J.Stefan' Institute in international standard problems for validation of modeling and programs for safety analysis. Listed are main international experimental facilities for collecting data basic for understanding of physical phenomena, code development and validation of modelling and programs. Since the results of international standard problem analyses are published in a joint final report, it is simple to asses the conformance of the results of a particular group with the experiment. Good results from three international exercises done so far, have encouraged the group to currently participate in OECD-ISP-22 which is a model of the Italian three loop PWR. (author)

Analyser-based phase contrast image reconstruction using geometrical optics.

Science.gov (United States)

Kitchen, M J; Pavlov, K M; Siu, K K W; Menk, R H; Tromba, G; Lewis, R A

2007-07-21

Analyser-based phase contrast imaging can provide radiographs of exceptional contrast at high resolution (geometrical optics are satisfied. Analytical phase retrieval can be performed by fitting the analyser rocking curve with a symmetric Pearson type VII function. The Pearson VII function provided at least a 10% better fit to experimentally measured rocking curves than linear or Gaussian functions. A test phantom, a hollow nylon cylinder, was imaged at 20 keV using a Si(1 1 1) analyser at the ELETTRA synchrotron radiation facility. Our phase retrieval method yielded a more accurate object reconstruction than methods based on a linear fit to the rocking curve. Where reconstructions failed to map expected values, calculations of the Takagi number permitted distinction between the violation of the geometrical optics conditions and the failure of curve fitting procedures. The need for synchronized object/detector translation stages was removed by using a large, divergent beam and imaging the object in segments. Our image acquisition and reconstruction procedure enables quantitative phase retrieval for systems with a divergent source and accounts for imperfections in the analyser.

Behavior-based safety on construction sites: a case study.

Science.gov (United States)

Choudhry, Rafiq M

2014-09-01

This work presents the results of a case study and describes an important area within the field of construction safety management, namely behavior-based safety (BBS). This paper adopts and develops a management approach for safety improvements in construction site environments. A rigorous behavioral safety system and its intervention program was implemented and deployed on target construction sites. After taking a few weeks of safety behavior measurements, the project management team implemented the designed intervention and measurements were taken. Goal-setting sessions were arranged on-site with workers' participation to set realistic and attainable targets of performance. Safety performance measurements continued and the levels of performance and the targets were presented on feedback charts. Supervisors were asked to give workers recognition and praise when they acted safely or improved critical behaviors. Observers were requested to have discussions with workers, visit the site, distribute training materials to workers, and provide feedback to crews and display charts. They were required to talk to operatives in the presence of line managers. It was necessary to develop awareness and understanding of what was being measured. In the process, operatives learned how to act safely when conducting site tasks using the designed checklists. Current weekly scores were discussed in the weekly safety meetings and other operational site meetings with emphasis on how to achieve set targets. The reliability of the safety performance measures taken by the company's observers was monitored. A clear increase in safety performance level was achieved across all categories: personal protective equipment; housekeeping; access to heights; plant and equipment, and scaffolding. The research reveals that scores of safety performance at one project improved from 86% (at the end of 3rd week) to 92.9% during the 9th week. The results of intervention demonstrated large decreases in

The influence of environmental conditions on safety management in hospitals: a qualitative study.

Science.gov (United States)

Alingh, Carien W; van Wijngaarden, Jeroen D H; Huijsman, Robbert; Paauwe, Jaap

2018-05-02

Hospitals are confronted with increasing safety demands from a diverse set of stakeholders, including governmental organisations, professional associations, health insurance companies, patient associations and the media. However, little is known about the effects of these institutional and competitive pressures on hospital safety management. Previous research has shown that organisations generally shape their safety management approach along the lines of control- or commitment-based management. Using a heuristic framework, based on the contextually-based human resource theory, we analysed how environmental pressures affect the safety management approach used by hospitals. A qualitative study was conducted into hospital care in the Netherlands. Five hospitals were selected for participation, based on organisational characteristics as well as variation in their reputation for patient safety. We interviewed hospital managers and staff with a central role in safety management. A total of 43 semi-structured interviews were conducted with 48 respondents. The heuristic framework was used as an initial model for analysing the data, though new codes emerged from the data as well. In order to ensure safe care delivery, institutional and competitive stakeholders often impose detailed safety requirements, strong forces for compliance and growing demands for accountability. As a consequence, hospitals experience a decrease in the room to manoeuvre. Hence, organisations increasingly choose a control-based management approach to make sure that safety demands are met. In contrast, in case of more abstract safety demands and an organisational culture which favours patient safety, hospitals generally experience more leeway. This often results in a stronger focus on commitment-based management. Institutional and competitive conditions as well as strategic choices that hospitals make have resulted in various combinations of control- and commitment-based safety management. A balanced

Sorption data base for the cementitious near-field of L/ILW and ILW repositories for provisional safety analyses for SGT-E2

International Nuclear Information System (INIS)

Wieland, E.

2014-11-01

The near-field of the planned Swiss repositories for low- and intermediate-level waste (L/ILW) and long-lived intermediate-level waste (ILW) consists of large quantities of cementitious materials. Hardened cement paste (HCP) is considered to be the most important sorbing material present in the near-field of L/ILW and ILW repositories. Interaction of radionuclides with HCP represents the most important mechanism retarding their migration from the near-field into the host rock. This report describes a cement sorption data base (SDB) for the safety-relevant radionuclides in the waste that will be disposed of in the L/ILW and ILW repositories. The current update on sorption values for radionuclides should be read in conjunction with the earlier SDBs CEM-94, CEM-97 and CEM-02. Sorption values have been selected based on procedures reported in these earlier SDBs. The values are revised if corresponding new information and/or data are available. The basic information results from a survey of sorption studies published between 2002 and 2013. The sorption values recommended in this report have either been selected from in-house experimental studies or from literature data, and they were further assessed with a view to the sorption values recently published in the framework of the safety analysis for the planned near surface disposal facility in Belgium. The report summarizes the sorption properties of HCP and compiles sorption values for safety-relevant radionuclides and low-molecular weight organic molecules on undisturbed and degraded HCP. A list of the safety-relevant radionuclides is provided. The radionuclide inventories are determined by the waste streams to be disposed of in the L/ILW and ILW repositories. Information on the elemental and mineral composition of HCP was obtained from hydration studies. The concentrations of the most important impurity elements in cement were obtained from dissolution studies on HCP. Particular emphasis is placed on summarizing our

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.

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

Energy Technology Data Exchange (ETDEWEB)

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

2004-07-01

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

New method for distance-based close following safety indicator.

Science.gov (United States)

Sharizli, A A; Rahizar, R; Karim, M R; Saifizul, A A

2015-01-01

The increase in the number of fatalities caused by road accidents involving heavy vehicles every year has raised the level of concern and awareness on road safety in developing countries like Malaysia. Changes in the vehicle dynamic characteristics such as gross vehicle weight, travel speed, and vehicle classification will affect a heavy vehicle's braking performance and its ability to stop safely in emergency situations. As such, the aim of this study is to establish a more realistic new distance-based safety indicator called the minimum safe distance gap (MSDG), which incorporates vehicle classification (VC), speed, and gross vehicle weight (GVW). Commercial multibody dynamics simulation software was used to generate braking distance data for various heavy vehicle classes under various loads and speeds. By applying nonlinear regression analysis to the simulation results, a mathematical expression of MSDG has been established. The results show that MSDG is dynamically changed according to GVW, VC, and speed. It is envisaged that this new distance-based safety indicator would provide a more realistic depiction of the real traffic situation for safety analysis.

Fire-safety engineering and performance-based codes

DEFF Research Database (Denmark)

Sørensen, Lars Schiøtt

project administrators, etc. The book deals with the following topics: • Historical presentation on the subject of fire • Legislation and building project administration • European fire standardization • Passive and active fire protection • Performance-based Codes • Fire-safety Engineering • Fundamental......Fire-safety Engineering is written as a textbook for Engineering students at universities and other institutions of higher education that teach in the area of fire. The book can also be used as a work of reference for consulting engineers, Building product manufacturers, contractors, building...... thermodynamics • Heat exchange during the fire process • Skin burns • Burning rate, energy release rate and design fires • Proposal to Risk-based design fires • Proposal to a Fire scale • Material ignition and flame spread • Fire dynamics in buildings • Combustion products and toxic gases • Smoke inhalation...

Nursing leaders' accountability to narrow the safety chasm: insights and implications from the collective evidence base on healthcare safety.

Science.gov (United States)

Jeffs, Lianne; Macmillan, Kathleen; McKey, Colleen; Ferris, Ella

2009-01-01

Challenges continue to exist in bridging the safety gap to ensure that consistent, high-quality nursing care is provided based on the best scientific knowledge available. This paper examines findings from nursing research presented at the symposium Advancing Nursing Leadership for a Safer Healthcare System, held in Toronto, Ontario in 2007. Four central themes emerged: (1) place the patient in safety; (2) generate a broader knowledge base on safety across the continuum of care; (3) create a safe culture and healthy work environment to mitigate current threats to patient safety; and (4) advance translation of evidence to practice at the organizational and clinical levels. The aim of this exchange of knowledge was to equip nursing leaders and their decision partners with evidence that can become a catalyst for mobilizing change in practice to address the safety chasm.

Integrated Deterministic-Probabilistic Safety Assessment Methodologies

Energy Technology Data Exchange (ETDEWEB)

Kudinov, P.; Vorobyev, Y.; Sanchez-Perea, M.; Queral, C.; Jimenez Varas, G.; Rebollo, M. J.; Mena, L.; Gomez-Magin, J.

2014-02-01

IDPSA (Integrated Deterministic-Probabilistic Safety Assessment) is a family of methods which use tightly coupled probabilistic and deterministic approaches to address respective sources of uncertainties, enabling Risk informed decision making in a consistent manner. The starting point of the IDPSA framework is that safety justification must be based on the coupling of deterministic (consequences) and probabilistic (frequency) considerations to address the mutual interactions between stochastic disturbances (e.g. failures of the equipment, human actions, stochastic physical phenomena) and deterministic response of the plant (i.e. transients). This paper gives a general overview of some IDPSA methods as well as some possible applications to PWR safety analyses. (Author)

Generic analyses for evaluation of low Charpy upper-shelf energy effects on safety margins against fracture of reactor pressure vessel materials

International Nuclear Information System (INIS)

Dickson, T.L.

1993-07-01

Appendix G to 10 CFR Part 50 requires that reactor pressure vessel beltline material maintain Charpy upper-shelf energies of no less than 50 ft-lb during the plant operating life, unless it is demonstrated in a manner approved by the Nuclear Regulatory Commission (NRC), that lower values of Charpy upper-shelf energy provide margins of safety against fracture equivalent to those in Appendix G to Section XI of the ASME Code. Analyses based on acceptance criteria and analysis methods adopted in the ASME Code Case N-512 are described herein. Additional information on material properties was provided by the NRC, Office of Nuclear Regulatory Research, Materials Engineering Branch. These cases, specified by the NRC, represent generic applications to boiling water reactor and pressurized water reactor vessels. This report is designated as HSST Report No. 140

Effect of motivational group interviewing-based safety education on Workers' safety behaviors in glass manufacturing.

Science.gov (United States)

Navidian, Ali; Rostami, Zahra; Rozbehani, Nasrin

2015-09-19

Worker safety education using models that identify and reinforce factors affecting behavior is essential. The present study aimed to determine the effect of safety education based on motivational interviewing on awareness of, attitudes toward, and engagement in worker safety in the glass production industry in Hamedan, Iran, in 2014. This was a quasi-experimental interventional study including a total of 70 production line workers at glass production facilities in Hamedan. The workers were randomly assigned to either an intervention or a control group, with 35 workers in each group. Participants in the control group received four one-hour safety education sessions, in the form of traditional lectures. Those in the intervention group received four educational sessions based on motivational group interviewing, which were conducted in four groups of eight to ten participants each. The instruments used included a researcher-developed questionnaire with checklists addressing safety awareness, and attitude and performance, which were completed before and 12 weeks after the intervention. The data were analyzed using descriptive statistics, independent and paired t-tests, and chi-squared tests. Having obtained the differences in scores before and after the intervention, we determined mean changes in the scores of awareness, attitude, and use of personal protective equipment among workers who underwent motivational group interviewing (3.74 ± 2.16, 1.71 ± 3.16, and 3.2 ± 1.92, respectively, p work environment.

Development of IFC based fire safety assesment tools

DEFF Research Database (Denmark)

Taciuc, Anca; Karlshøj, Jan; Dederichs, Anne

2016-01-01

Due to the impact that the fire safety design has on the building's layout and on other complementary systems, as installations, it is important during the conceptual design stage to evaluate continuously the safety level in the building. In case that the task is carried out too late, additional...... 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...

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.)

Review of domestic and international experience on optimization of tests planning for safety related systems at NPP

International Nuclear Information System (INIS)

Skalozubov, V.I.; Komarov, Yu.A.; Kolykanov, V.N.; Kochneva, V.Yu.; Gablaya, T.V.

2009-01-01

There are represented the basic requirements of normative and operating documents on test periodicity of safety related systems at NPPs, sets out the theoretical methods of test optimization of the technical systems, and analyses foreign engineering methods for changing test periodicity of the NPP systems. Based on this review analyses further tasks are formulated for improvement of the methodical base of optimization of tests planning for safety related systems

Comparative analysis of safety related site characteristics

International Nuclear Information System (INIS)

Andersson, Johan

2010-12-01

This document presents a comparative analysis of site characteristics related to long-term safety for the two candidate sites for a final repository for spent nuclear fuel in Forsmark (municipality of Oesthammar) and in Laxemar (municipality of Oskarshamn) from the point of view of site selection. The analyses are based on the updated site descriptions of Forsmark /SKB 2008a/ and Laxemar /SKB 2009a/, together with associated updated repository layouts and designs /SKB 2008b and SKB 2009b/. The basis for the comparison is thus two equally and thoroughly assessed sites. However, the analyses presented here are focussed on differences between the sites rather than evaluating them in absolute terms. The document serves as a basis for the site selection, from the perspective of long-term safety, in SKB's application for a final repository. A full evaluation of safety is made for a repository at the selected site in the safety assessment SR-Site /SKB 2011/, referred to as SR-Site main report in the following

Comparative analysis of safety related site characteristics

Energy Technology Data Exchange (ETDEWEB)

Andersson, Johan (ed.)

2010-12-15

This document presents a comparative analysis of site characteristics related to long-term safety for the two candidate sites for a final repository for spent nuclear fuel in Forsmark (municipality of Oesthammar) and in Laxemar (municipality of Oskarshamn) from the point of view of site selection. The analyses are based on the updated site descriptions of Forsmark /SKB 2008a/ and Laxemar /SKB 2009a/, together with associated updated repository layouts and designs /SKB 2008b and SKB 2009b/. The basis for the comparison is thus two equally and thoroughly assessed sites. However, the analyses presented here are focussed on differences between the sites rather than evaluating them in absolute terms. The document serves as a basis for the site selection, from the perspective of long-term safety, in SKB's application for a final repository. A full evaluation of safety is made for a repository at the selected site in the safety assessment SR-Site /SKB 2011/, referred to as SR-Site main report in the following

Engineering nanomaterials-based biosensors for food safety detection.

Science.gov (United States)

Lv, Man; Liu, Yang; Geng, Jinhui; Kou, Xiaohong; Xin, Zhihong; Yang, Dayong

2018-05-30

Food safety always remains a grand global challenge to human health, especially in developing countries. To solve food safety pertained problems, numerous strategies have been developed to detect biological and chemical contaminants in food. Among these approaches, nanomaterials-based biosensors provide opportunity to realize rapid, sensitive, efficient and portable detection, overcoming the restrictions and limitations of traditional methods such as complicated sample pretreatment, long detection time, and relying on expensive instruments and well-trained personnel. In this review article, we provide a cross-disciplinary perspective to review the progress of nanomaterials-based biosensors for the detection of food contaminants. The review article is organized by the category of food contaminants including pathogens/toxins, heavy metals, pesticides, veterinary drugs and illegal additives. In each category of food contaminant, the biosensing strategies are summarized including optical, colorimetric, fluorescent, electrochemical, and immune- biosensors; the relevant analytes, nanomaterials and biosensors are analyzed comprehensively. Future perspectives and challenges are also discussed briefly. We envision that our review could bridge the gap between the fields of food science and nanotechnology, providing implications for the scientists or engineers in both areas to collaborate and promote the development of nanomaterials-based biosensors for food safety detection. Copyright © 2018 Elsevier B.V. All rights reserved.

Meta-Analyses of Human Cell-Based Cardiac Regeneration Therapies

DEFF Research Database (Denmark)

Gyöngyösi, Mariann; Wojakowski, Wojciech; Navarese, Eliano P

2016-01-01

In contrast to multiple publication-based meta-analyses involving clinical cardiac regeneration therapy in patients with recent myocardial infarction, a recently published meta-analysis based on individual patient data reported no effect of cell therapy on left ventricular function or clinical...

Space nuclear reactor safety

International Nuclear Information System (INIS)

Damon, D.; Temme, M.; Brown, N.

1990-01-01

Definition of safety requirements and design features of the SP-100 space reactor power system has been guided by a mission risk analysis. The analysis quantifies risk from accidental radiological consequences for a reference mission. Results show that the radiological risk from a space reactor can be made very low. The total mission risk from radiological consequences for a shuttle-launched, earth orbit SP-100 mission is estimated to be 0.05 Person-REM (expected values) based on a 1 mREM/yr de Minimus dose. Results are given for each mission phase. The safety benefits of specific design features are evaluated through risk sensitivity analyses

Safety Testing of Ammonium Nitrate Based Mixtures

Science.gov (United States)

Phillips, Jason; Lappo, Karmen; Phelan, James; Peterson, Nathan; Gilbert, Don

2013-06-01

Ammonium nitrate (AN)/ammonium nitrate based explosives have a lengthy documented history of use by adversaries in acts of terror. While historical research has been conducted on AN-based explosive mixtures, it has primarily focused on detonation performance while varying the oxygen balance between the oxidizer and fuel components. Similarly, historical safety data on these materials is often lacking in pertinent details such as specific fuel type, particle size parameters, oxidizer form, etc. A variety of AN-based fuel-oxidizer mixtures were tested for small-scale sensitivity in preparation for large-scale testing. Current efforts focus on maintaining a zero oxygen-balance (a stoichiometric ratio for active chemical participants) while varying factors such as charge geometry, oxidizer form, particle size, and inert diluent ratios. Small-scale safety testing was conducted on various mixtures and fuels. It was found that ESD sensitivity is significantly affected by particle size, while this is less so for impact and friction. Thermal testing is in progress to evaluate hazards that may be experienced during large-scale testing.

Requirements of safety and reliability

International Nuclear Information System (INIS)

Franzen, L.F.

1977-01-01

The safety strategy for nuclear power plants is characterized by the fact that the high level of safety was attained not as a result of experience, but on the basis of preventive accident analyses and the findings derived from such analyses. Although, in these accident analyses, the deterministic approach is predominant it is supplemented by reliability analyses. The accidents analyzed in nuclear licensing procedures cover a wide spectrum from minor incidents to the design basis accidents which determine the design of the safety devices. The initial and boundary conditions, which are essential for accident analyses, and the determination of the loads occuring in various states during regular operation and in accidents flow into the design of the individual systems and components. The inevitable residual risk and its origins are discussed. (orig./HP) [de

Updated safety analysis of ITER

Energy Technology Data Exchange (ETDEWEB)

Taylor, Neill, E-mail: neill.taylor@iter.org [ITER Organization, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Baker, Dennis; Ciattaglia, Sergio; Cortes, Pierre; Elbez-Uzan, Joelle; Iseli, Markus; Reyes, Susana; Rodriguez-Rodrigo, Lina; Rosanvallon, Sandrine; Topilski, Leonid [ITER Organization, CS 90 046, 13067 St Paul Lez Durance Cedex (France)

2011-10-15

An updated version of the ITER Preliminary Safety Report has been produced and submitted to the licensing authorities. It is revised and expanded in response to requests from the authorities after their review of an earlier version in 2008, to reflect enhancements in ITER safety provisions through design changes, to incorporate new and improved safety analyses and to take into account other ITER design evolution. The updated analyses show that changes to the Tokamak cooling water system design have enhanced confinement and reduced potential radiological releases as well as removing decay heat with very high reliability. New and updated accident scenario analyses, together with fire and explosion risk analyses, have shown that design provisions are sufficient to minimize the likelihood of accidents and reduce potential consequences to a very low level. Taken together, the improvements provided a stronger demonstration of the very good safety performance of the ITER design.

Updated safety analysis of ITER

International Nuclear Information System (INIS)

Taylor, Neill; Baker, Dennis; Ciattaglia, Sergio; Cortes, Pierre; Elbez-Uzan, Joelle; Iseli, Markus; Reyes, Susana; Rodriguez-Rodrigo, Lina; Rosanvallon, Sandrine; Topilski, Leonid

2011-01-01

An updated version of the ITER Preliminary Safety Report has been produced and submitted to the licensing authorities. It is revised and expanded in response to requests from the authorities after their review of an earlier version in 2008, to reflect enhancements in ITER safety provisions through design changes, to incorporate new and improved safety analyses and to take into account other ITER design evolution. The updated analyses show that changes to the Tokamak cooling water system design have enhanced confinement and reduced potential radiological releases as well as removing decay heat with very high reliability. New and updated accident scenario analyses, together with fire and explosion risk analyses, have shown that design provisions are sufficient to minimize the likelihood of accidents and reduce potential consequences to a very low level. Taken together, the improvements provided a stronger demonstration of the very good safety performance of the ITER design.

Thermal hydraulic analyses of LVR-15 research reactor with IRT-M fuel

International Nuclear Information System (INIS)

Macek, J.

1997-01-01

The LVR-15 pool-type research reactor has been in operation at the Nuclear Research Institute at Rez since 1955. Following a number of reconstructions and redesigning, the current reactor power is 15 MW. Thermal hydraulic analyses to demonstrate that the core heat will be safely removed during operation as well as in accident situations were performed based on methodology which had been specifically developed for the LVR-15 research reactor. This methodology was applied to stationary thermal hydraulic computations, as well as to transients, particularly with reactivity failure and loss of circulation pumps emergencies. The applied methodology and the core configuration as used in the Safety Report are described. The initial and boundary conditions are then considered and the summary of the calculated failures with regard to the defined safety limits is presented. The results of the core configuration analyses are also discussed with respect to meeting the safety limits and to the applicability of the methodology to this purpose

Analyser-based phase contrast image reconstruction using geometrical optics

International Nuclear Information System (INIS)

Kitchen, M J; Pavlov, K M; Siu, K K W; Menk, R H; Tromba, G; Lewis, R A

2007-01-01

Analyser-based phase contrast imaging can provide radiographs of exceptional contrast at high resolution (<100 μm), whilst quantitative phase and attenuation information can be extracted using just two images when the approximations of geometrical optics are satisfied. Analytical phase retrieval can be performed by fitting the analyser rocking curve with a symmetric Pearson type VII function. The Pearson VII function provided at least a 10% better fit to experimentally measured rocking curves than linear or Gaussian functions. A test phantom, a hollow nylon cylinder, was imaged at 20 keV using a Si(1 1 1) analyser at the ELETTRA synchrotron radiation facility. Our phase retrieval method yielded a more accurate object reconstruction than methods based on a linear fit to the rocking curve. Where reconstructions failed to map expected values, calculations of the Takagi number permitted distinction between the violation of the geometrical optics conditions and the failure of curve fitting procedures. The need for synchronized object/detector translation stages was removed by using a large, divergent beam and imaging the object in segments. Our image acquisition and reconstruction procedure enables quantitative phase retrieval for systems with a divergent source and accounts for imperfections in the analyser

ITER safety

International Nuclear Information System (INIS)

Raeder, J.; Piet, S.; Buende, R.

1991-01-01

As part of the series of publications by the IAEA that summarize the results of the Conceptual Design Activities for the ITER project, this document describes the ITER safety analyses. It contains an assessment of normal operation effluents, accident scenarios, plasma chamber safety, tritium system safety, magnet system safety, external loss of coolant and coolant flow problems, and a waste management assessment, while it describes the implementation of the safety approach for ITER. The document ends with a list of major conclusions, a set of topical remarks on technical safety issues, and recommendations for the Engineering Design Activities, safety considerations for siting ITER, and recommendations with regard to the safety issues for the R and D for ITER. Refs, figs and tabs

Conducting Clinically Based Intimate Partner Violence Research: Safety Protocol Recommendations.

Science.gov (United States)

Anderson, Jocelyn C; Glass, Nancy E; Campbell, Jacquelyn C

Maintaining safety is of utmost importance during research involving participants who have experienced intimate partner violence (IPV). Limited guidance on safety protocols to protect participants is available, particularly information related to technology-based approaches to informed consent, data collection, and contacting participants during the course of a study. The purpose of the article is to provide details on the safety protocol developed and utilized with women receiving care at an urban HIV clinic and who were taking part in an observational study of IPV, mental health symptoms, and substance abuse and their relationship to HIV treatment adherence. The protocol presents the technological strategies to promote safety and allow autonomy in participant decision-making throughout the research process, including Voice over Internet Protocol telephone numbers, and tablet-based eligibility screening and data collection. Protocols for management of participants at risk for suicide and/or intimate partner homicide that included automated high-risk messaging to participants and research staff and facilitated disclosure of risk to clinical staff based on participant preferences are discussed. Use of technology and partnership with clinic staff helped to provide an environment where research regarding IPV could be conducted without undue burden or risk to participants. Utilizing tablet-based survey administration provided multiple practical and safety benefits for participants. Most women who screened into high-risk categories for suicide or intimate partner homicide did not choose to have their results shared with their healthcare providers, indicating the importance of allowing participants control over information sharing whenever possible.

[Patient safety and errors in medicine: development, prevention and analyses of incidents].

Science.gov (United States)

Rall, M; Manser, T; Guggenberger, H; Gaba, D M; Unertl, K

2001-06-01

"Patient safety" and "errors in medicine" are issues gaining more and more prominence in the eyes of the public. According to newer studies, errors in medicine are among the ten major causes of death in association with the whole area of health care. A new era has begun incorporating attention to a "systems" approach to deal with errors and their causes in the health system. In other high-risk domains with a high demand for safety (such as the nuclear power industry and aviation) many strategies to enhance safety have been established. It is time to study these strategies, to adapt them if necessary and apply them to the field of medicine. These strategies include: to teach people how errors evolve in complex working domains and how types of errors are classified; the introduction of critical incident reporting systems that are free of negative consequences for the reporters; the promotion of continuous medical education; and the development of generic problem-solving skills incorporating the extensive use of realistic simulators wherever possible. Interestingly, the field of anesthesiology--within which realistic simulators were developed--is referred to as a model for the new patient safety movement. Despite this proud track record in recent times though, there is still much to be done even in the field of anesthesiology. Overall though, the most important strategy towards a long-term improvement in patient safety will be a change of "culture" throughout the entire health care system. The "culture of blame" focused on individuals should be replaced by a "safety culture", that sees errors and critical incidents as a problem of the whole organization. The acceptance of human fallability and an open-minded non-punitive analysis of errors in the sense of a "preventive and proactive safety culture" should lead to solutions at the systemic level. This change in culture can only be achieved with a strong commitment from the highest levels of an organization. Patient

Protocol for a multicentre, multistage, prospective study in China using system-based approaches for consistent improvement in surgical safety.

Science.gov (United States)

Yu, Xiaochu; Jiang, Jingmei; Liu, Changwei; Shen, Keng; Wang, Zixing; Han, Wei; Liu, Xingrong; Lin, Guole; Zhang, Ye; Zhang, Ying; Ma, Yufen; Bo, Haixin; Zhao, Yupei

2017-06-15

Surgical safety has emerged as a crucial global health issue in the past two decades. Although several safety-enhancing tools are available, the pace of large-scale improvement remains slow, especially in developing countries such as China. The present project (Modern Surgery and Anesthesia Safety Management System Construction and Promotion) aims to develop and validate system-based integrated approaches for reducing perioperative deaths and complications using a multicentre, multistage design. The project involves collection of clinical and outcome information for 1 20 000 surgical inpatients at four regionally representative academic/teaching general hospitals in China during three sequential stages: preparation and development, effectiveness validation and improvement of implementation for promotion. These big data will provide the evidence base for the formulation, validation and improvement processes of a system-based stratified safety intervention package covering the entire surgical pathway. Attention will be directed to managing inherent patient risks and regulating medical safety behaviour. Information technology will facilitate data collection and intervention implementation, provide supervision mechanisms and guarantee transfer of key patient safety messages between departments and personnel. Changes in rates of deaths, surgical complications during hospitalisation, length of stay, system adoption and implementation rates will be analysed to evaluate effectiveness and efficiency. This study was approved by the institutional review boards of Peking Union Medical College Hospital, First Hospital of China Medical University, Qinghai Provincial People's Hospital, Xiangya Hospital Central South University and the Institute of Basic Medical Sciences, Chinese Academy of Medical Sciences. Study findings will be disseminated via peer-reviewed journals, conference presentations and patent papers. © Article author(s) (or their employer(s) unless otherwise

Risk-based evaluation tool for safety-related maintenance involving scaffolding

International Nuclear Information System (INIS)

Stevens, C.; Azizi, M.; Massman, M.

1988-01-01

The US Nuclear Regulatory Commission (NRC) has expressed a general concern that transient materials in and around safety systems at nuclear power plants represent a seismic safety hazard to the plant, in particular, the uncontrolled use of scaffolding during maintenance activities. Currently, most plants perform a seismic safety analysis for all uses of scaffolding near safety-related equipment to determine appropriate tie-down locations, scaffolding reinforcements, etc. This is both time-consuming and, for the most part, unnecessary. A workable engineering solution based on risk analysis techniques has been developed and is being used at the Palo Verde nuclear generating station (PVNGS)

LWR safety studies. Analyses and further assessments relating to the German Risk Assessment Study on Nuclear Power Plants. Vol. 3

International Nuclear Information System (INIS)

1983-01-01

Critical review of the analyses of the German Risk Assessment Study on Nuclear Power Plants (DRS) concerning the reliability of the containment under accident conditions and the conditions of fission product release (transport and distribution in the environment). Main point of interest in this context is an explosion in the steam section and its impact on the containment. Critical comments are given on the models used in the DRS for determining the accident consequences. The analyses made deal with the mathematical models and database for propagation calculations, the methods of dose computation and assessment of health hazards, and the modelling of protective and safety measures. Social impacts of reactor accidents are also considered. (RF) [de

Developing safety performance functions incorporating reliability-based risk measures.

Science.gov (United States)

Ibrahim, Shewkar El-Bassiouni; Sayed, Tarek

2011-11-01

Current geometric design guides provide deterministic standards where the safety margin of the design output is generally unknown and there is little knowledge of the safety implications of deviating from these standards. Several studies have advocated probabilistic geometric design where reliability analysis can be used to account for the uncertainty in the design parameters and to provide a risk measure of the implication of deviation from design standards. However, there is currently no link between measures of design reliability and the quantification of safety using collision frequency. The analysis presented in this paper attempts to bridge this gap by incorporating a reliability-based quantitative risk measure such as the probability of non-compliance (P(nc)) in safety performance functions (SPFs). Establishing this link will allow admitting reliability-based design into traditional benefit-cost analysis and should lead to a wider application of the reliability technique in road design. The present application is concerned with the design of horizontal curves, where the limit state function is defined in terms of the available (supply) and stopping (demand) sight distances. A comprehensive collision and geometric design database of two-lane rural highways is used to investigate the effect of the probability of non-compliance on safety. The reliability analysis was carried out using the First Order Reliability Method (FORM). Two Negative Binomial (NB) SPFs were developed to compare models with and without the reliability-based risk measures. It was found that models incorporating the P(nc) provided a better fit to the data set than the traditional (without risk) NB SPFs for total, injury and fatality (I+F) and property damage only (PDO) collisions. Copyright © 2011 Elsevier Ltd. All rights reserved.

Safety study application guide

International Nuclear Information System (INIS)

1993-07-01

Martin Marietta Energy Systems, Inc., (Energy Systems) is committed to performing and documenting safety analyses for facilities it manages for the Department of Energy (DOE). Included are analyses of existing facilities done under the aegis of the Safety Analysis Report Upgrade Program, and analyses of new and modified facilities. A graded approach is used wherein the level of analysis and documentation for each facility is commensurate with the magnitude of the hazard(s), the complexity of the facility and the stage of the facility life cycle. Safety analysis reports (SARs) for hazard Category 1 and 2 facilities are usually detailed and extensive because these categories are associated with public health and safety risk. SARs for Category 3 are normally much less extensive because the risk to public health and safety is slight. At Energy Systems, safety studies are the name given to SARs for Category 3 (formerly open-quotes lowclose quotes) facilities. Safety studies are the appropriate instrument when on-site risks are limited to irreversible consequences to a few people, and off-site consequences are limited to reversible consequences to a few people. This application guide provides detailed instructions for performing safety studies that meet the requirements of DOE Orders 5480.22, open-quotes Technical Safety Requirements,close quotes and 5480.23, open-quotes Nuclear Safety Analysis Reports.close quotes A seven-chapter format has been adopted for safety studies. This format allows for discussion of all the items required by DOE Order 5480.23 and for the discussions to be readily traceable to the listing in the order. The chapter titles are: (1) Introduction and Summary, (2) Site, (3) Facility Description, (4) Safety Basis, (5) Hazardous Material Management, (6) Management, Organization, and Institutional Safety Provisions, and (7) Accident Analysis

Probabilistic safety analysis and interpretation thereof

International Nuclear Information System (INIS)

Steininger, U.; Sacher, H.

1999-01-01

Increasing use of the instrumentation of PSA is being made in Germany for quantitative technical safety assessment, for example with regard to incidents which must be reported and forwarding of information, especially in the case of modification of nuclear plants. The Commission for Nuclear Reactor Safety recommends regular execution of PSA on a cycle period of ten years. According to the PSA guidance instructions, probabilistic analyses serve for assessing the degree of safety of the entire plant, expressed as the expectation value for the frequency of endangering conditions. The authors describe the method, action sequence and evaluation of the probabilistic safety analyses. The limits of probabilistic safety analyses arise in the practical implementation. Normally the guidance instructions for PSA are confined to the safety systems, so that in practice they are at best suitable for operational optimisation only to a limited extent. The present restriction of the analyses has a similar effect on power output operation of the plant. This seriously degrades the utilitarian value of these analyses for the plant operators. In order to further develop PSA as a supervisory and operational optimisation instrument, both authors consider it to be appropriate to bring together the specific know-how of analysts, manufacturers, plant operators and experts. (orig.) [de

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

Crane Safety Assessment Method Based on Entropy and Cumulative Prospect Theory

Directory of Open Access Journals (Sweden)

Aihua Li

2017-01-01

Full Text Available Assessing the safety status of cranes is an important problem. To overcome the inaccuracies and misjudgments in such assessments, this work describes a safety assessment method for cranes that combines entropy and cumulative prospect theory. Firstly, the proposed method transforms the set of evaluation indices into an evaluation vector. Secondly, a decision matrix is then constructed from the evaluation vectors and evaluation standards, and an entropy-based technique is applied to calculate the index weights. Thirdly, positive and negative prospect value matrices are established from reference points based on the positive and negative ideal solutions. Thus, this enables the crane safety grade to be determined according to the ranked comprehensive prospect values. Finally, the safety status of four general overhead traveling crane samples is evaluated to verify the rationality and feasibility of the proposed method. The results demonstrate that the method described in this paper can precisely and reasonably reflect the safety status of a crane.

Diversity for security: case assessment for FPGA-based safety-critical systems

Directory of Open Access Journals (Sweden)

Kharchenko Vyacheslav

2016-01-01

Full Text Available Industrial safety critical instrumentation and control systems (I&Cs are facing more with information (in general and cyber, in particular security threats and attacks. The application of programmable logic, first of all, field programmable gate arrays (FPGA in critical systems causes specific safety deficits. Security assessment techniques for such systems are based on heuristic knowledges and the expert judgment. Main challenge is how to take into account features of FPGA technology for safety critical I&Cs including systems in which are applied diversity approach to minimize risks of common cause failure. Such systems are called multi-version (MV systems. The goal of the paper is in description of the technique and tool for case-based security assessment of MV FPGA-based I&Cs.

Safety analysis for research reactors

International Nuclear Information System (INIS)

2008-01-01

The aim of safety analysis for research reactors is to establish and confirm the design basis for items important to safety using appropriate analytical tools. The design, manufacture, construction and commissioning should be integrated with the safety analysis to ensure that the design intent has been incorporated into the as-built reactor. Safety analysis assesses the performance of the reactor against a broad range of operating conditions, postulated initiating events and other circumstances, in order to obtain a complete understanding of how the reactor is expected to perform in these situations. Safety analysis demonstrates that the reactor can be kept within the safety operating regimes established by the designer and approved by the regulatory body. This analysis can also be used as appropriate in the development of operating procedures, periodic testing and inspection programmes, proposals for modifications and experiments and emergency planning. The IAEA Safety Requirements publication on the Safety of Research Reactors states that the scope of safety analysis is required to include analysis of event sequences and evaluation of the consequences of the postulated initiating events and comparison of the results of the analysis with radiological acceptance criteria and design limits. This Safety Report elaborates on the requirements established in IAEA Safety Standards Series No. NS-R-4 on the Safety of Research Reactors, and the guidance given in IAEA Safety Series No. 35-G1, Safety Assessment of Research Reactors and Preparation of the Safety Analysis Report, providing detailed discussion and examples of related topics. Guidance is given in this report for carrying out safety analyses of research reactors, based on current international good practices. The report covers all the various steps required for a safety analysis; that is, selection of initiating events and acceptance criteria, rules and conventions, types of safety analysis, selection of

Software safety hazard analysis

International Nuclear Information System (INIS)

Lawrence, J.D.

1996-02-01

Techniques for analyzing the safety and reliability of analog-based electronic protection systems that serve to mitigate hazards in process control systems have been developed over many years, and are reasonably well understood. An example is the protection system in a nuclear power plant. The extension of these techniques to systems which include digital computers is not well developed, and there is little consensus among software engineering experts and safety experts on how to analyze such systems. One possible technique is to extend hazard analysis to include digital computer-based systems. Software is frequently overlooked during system hazard analyses, but this is unacceptable when the software is in control of a potentially hazardous operation. In such cases, hazard analysis should be extended to fully cover the software. A method for performing software hazard analysis is proposed in this paper

Safety climate in OHSAS 18001-certified organisations: antecedents and consequences of safety behaviour.

Science.gov (United States)

Fernández-Muñiz, Beatriz; Montes-Peón, José Manuel; Vázquez-Ordás, Camilo José

2012-03-01

The occupational health and safety standard OHSAS 18001 has gained considerable acceptance worldwide, and firms from diverse sectors and of varying sizes have implemented it. Despite this, very few studies have analysed safety management or the safety climate in OHSAS 18001-certified organisations. The current work aims to analyse the safety climate in these organisations, identify its dimensions, and propose and test a structural equation model that will help determine the antecedents and consequences of employees' safety behaviour. For this purpose, the authors carry out an empirical study using a sample of 131 OHSAS 18001-certified organisations located in Spain. The results show that management's commitment, and particularly communication, have an effect on safety behaviour and on safety performance, employee satisfaction, and firm competitiveness. These findings are particularly important for management since they provide evidence about the factors that should be encouraged to reduce risks and improve performance in this type of organisation. Copyright © 2011 Elsevier Ltd. All rights reserved.

ASN guide project. Safety policy and management in INBs (base nuclear installations)

International Nuclear Information System (INIS)

2010-01-01

This guide presents the recommendations of the French Nuclear Safety Authority (ASN) in the field of safety policy and management (PMS) for base nuclear installations (INBs). It gives an overview and comments of some prescriptions of the so-called INB order and PMS decision. These regulatory texts define a framework for provisions any INB operator must implement to establish his safety policy, to define and implement a system which allows the safety to be maintained, the improvement of his INB safety to be permanently looked for. The following issues are addressed: operator's safety policy, identification of elements important for safety, of activities pertaining to safety, and of associated requirements, safety management organization and system, management of activities pertaining to safety, documentation and archiving

Patient Safety Based Knowledge Management SECI to Improve Nusrsing Students Competency

Directory of Open Access Journals (Sweden)

Joanggi Wiriatarina Harianto

2015-10-01

Full Text Available Introduction: Patient safety is an important component of health services quality,and  basic principles of patient care. Nursing students also have a great potential to make an action that could endanger the patient, because hospital is one of student practice area. The purpose of this study was to improve the nursing students competency in patient safety by using knowledge management SECI approached. Method: The study used exploratory survey, and quasy experiment. The samples were some of nursing students of STIKes Muhammadiyah Samarinda who were on internship programme that selected using simple random sampling technique, in total of 54 students. This research’s variables were the knowledge management SECI based-patient safety and nursing student’s competency. The data were collected by using questionnaires and observation. The data were analyze by using Partial Least Square (PLS. Result: The result showed that there were significant influence the implementation of a model patient safety based knowledge management seci on increased competence nursing students. Discussion: Improved student competency in patient safety using SECI knowledge management was carried out in four phases, that is Socialization, Externalization, Combination, and Internalization. The result was a new knowledge related to patient safety that able to improve the student’s competency.. Keywords: Patient safety, Knowledge management, SECI, competency

A bromine-based dichroic X-ray polarization analyser

CERN Document Server

Collins, S P; Brown, S D; Thompson, P

2001-01-01

We have demonstrated the advantages offered by dichroic X-ray polarization filters for linear polarization analysis, and describe such a device, based on a dibromoalkane/urea inclusion compound. The polarizer has been successfully tested by analysing the polarization of magnetic diffraction from holmium.

Occupational Therapy Home Safety Intervention via Telehealth

Directory of Open Access Journals (Sweden)

Lori E. Breeden

2016-07-01

Full Text Available Photography can be an effective addition for education-based telehealth services delivered by an occupational therapist.  In this study, photography was used as antecedent to telehealth sessions delivered by an occupational therapist focused on narrative learning about home safety.  After taking photographs of past home safety challenges, six participants experienced three web-based occupational therapy sessions each.  Sessions were recorded and transcribed.  Data were examined using content analysis.  A content analysis identified the following themes as well as an understanding of the learning process.  Analyses yielded themes of: the value of photos to support learning, the value of narrative learning related to home safety education, abstract versus concrete learners.  Procedural findings are included to support future endeavors.  Findings indicate that within a wellness context, home safety education for older adults can be delivered effectively via telehealth when using photography as a part of an occupational therapy intervention.

Safety of nuclear power plants: Design. Safety requirements

International Nuclear Information System (INIS)

2000-01-01

The present publication supersedes the Code on the Safety of Nuclear Power Plants: Design (Safety Series No. 50-C-D (Rev. 1), issued in 1988). It takes account of developments relating to the safety of nuclear power plants since the Code on Design was last revised. These developments include the issuing of the Safety Fundamentals publication, The Safety of Nuclear Installations, and the present revision of various safety standards and other publications relating to safety. Requirements for nuclear safety are intended to ensure adequate protection of site personnel, the public and the environment from the effects of ionizing radiation arising from nuclear power plants. It is recognized that technology and scientific knowledge advance, and nuclear safety and what is considered adequate protection are not static entities. Safety requirements change with these developments and this publication reflects the present consensus. This Safety Requirements publication takes account of the developments in safety requirements by, for example, including the consideration of severe accidents in the design process. Other topics that have been given more detailed attention include management of safety, design management, plant ageing and wearing out effects, computer based safety systems, external and internal hazards, human factors, feedback of operational experience, and safety assessment and verification. This publication establishes safety requirements that define the elements necessary to ensure nuclear safety. These requirements are applicable to safety functions and the associated structures, systems and components, as well as to procedures important to safety in nuclear power plants. It is expected that this publication will be used primarily for land based stationary nuclear power plants with water cooled reactors designed for electricity generation or for other heat production applications (such as district heating or desalination). It is recognized that in the case of

Safety of nuclear power plants: Design. Safety requirements

International Nuclear Information System (INIS)

2004-01-01

The present publication supersedes the Code on the Safety of Nuclear Power Plants: Design (Safety Series No. 50-C-D (Rev. 1), issued in 1988). It takes account of developments relating to the safety of nuclear power plants since the Code on Design was last revised. These developments include the issuing of the Safety Fundamentals publication, The Safety of Nuclear Installations, and the present revision of various safety standards and other publications relating to safety. Requirements for nuclear safety are intended to ensure adequate protection of site personnel, the public and the environment from the effects of ionizing radiation arising from nuclear power plants. It is recognized that technology and scientific knowledge advance, and nuclear safety and what is considered adequate protection are not static entities. Safety requirements change with these developments and this publication reflects the present consensus. This Safety Requirements publication takes account of the developments in safety requirements by, for example, including the consideration of severe accidents in the design process. Other topics that have been given more detailed attention include management of safety, design management, plant ageing and wearing out effects, computer based safety systems, external and internal hazards, human factors, feedback of operational experience, and safety assessment and verification. This publication establishes safety requirements that define the elements necessary to ensure nuclear safety. These requirements are applicable to safety functions and the associated structures, systems and components, as well as to procedures important to safety in nuclear power plants. It is expected that this publication will be used primarily for land based stationary nuclear power plants with water cooled reactors designed for electricity generation or for other heat production applications (such as district heating or desalination). It is recognized that in the case of

Preclosure Safety Analysis Guide

International Nuclear Information System (INIS)

D.D. Orvis

2003-01-01

A preclosure safety analysis (PSA) is a required element of the License Application (LA) for the high- level radioactive waste repository at Yucca Mountain. This guide provides analysts and other Yucca Mountain Repository Project (the Project) personnel with standardized methods for developing and documenting the PSA. The definition of the PSA is provided in 10 CFR 63.2, while more specific requirements for the PSA are provided in 10 CFR 63.112, as described in Sections 1.2 and 2. The PSA requirements described in 10 CFR Part 63 were developed as risk-informed performance-based regulations. These requirements must be met for the LA. The PSA addresses the safety of the Geologic Repository Operations Area (GROA) for the preclosure period (the time up to permanent closure) in accordance with the radiological performance objectives of 10 CFR 63.111. Performance objectives for the repository after permanent closure (described in 10 CFR 63.113) are not mentioned in the requirements for the PSA and they are not considered in this guide. The LA will be comprised of two phases: the LA for construction authorization (CA) and the LA amendment to receive and possess (R and P) high-level radioactive waste (HLW). PSA methods must support the safety analyses that will be based on the differing degrees of design detail in the two phases. The methods described herein combine elements of probabilistic risk assessment (PRA) and deterministic analyses that comprise a risk-informed performance-based safety analysis. This revision to the PSA guide was prepared for the following objectives: (1) To correct factual and typographical errors. (2) To provide additional material suggested from reviews by the Project, the U.S. Department of Energy (DOE), and U.S. Nuclear Regulatory Commission (NRC) Staffs. (3) To update material in accordance with approaches and/or strategies adopted by the Project. In addition, a principal objective for the planned revision was to ensure that the methods and

Safety KPIs - Monitoring of safety performance

Directory of Open Access Journals (Sweden)

Andrej Lališ

2014-09-01

Full Text Available This paper aims to provide brief overview of aviation safety development focusing on modern trends represented by implementation of Safety Key Performance Indicators. Even though aviation is perceived as safe means of transport, it is still struggling with its complexity given by long-term growth and robustness which it has reached today. Thus nowadays safety issues are much more complex and harder to handle than ever before. We are more and more concerned about organizational factors and control mechanisms which have potential to further increase level of aviation safety. Within this paper we will not only introduce the concept of Key Performance Indicators in area of aviation safety as an efficient control mechanism, but also analyse available legislation and documentation. Finally we will propose complex set of indicators which could be applied to Czech Air Navigation Service Provider.

Design provisions for safety

International Nuclear Information System (INIS)

Birkhofer, A.

1983-01-01

Design provisions for safety of nuclear power plants are based on a well balanced concept: the public is protected against a release of radioactive material by multiple barriers. These barriers are protected according to a 'defence-in-depth' principle. The reactor safety concept is primarily aimed at the prevention of accidents, especially fuel damage. Additionally, measures for consequence limitation are provided in order to prevent a severe release of radioactivity to the environment. However, it is difficult to judge the overall effectiveness of such devices. In a comprehensive safety analysis it has to be shown that the protection systems and safeguards work with sufficient reliability in the event of an accident. For the reliability assessment deterministic criteria (single failure, redundancy, fail-safe, demand for diversity) play an important role. Increasing efforts have been made to assess reliability quantitatively by means of probabilistic methods. It is now usual to perform reliability analyses of essential systems of nuclear power plants in the course of licensing procedures. As an additional level of emergency measures for a further reduction of hazards a reasonable amount of accident information has to be transferred. Operational experience may be considered as an important feedback to the design of plant safety features. Operator training has to include, besides skill in performing of operating procedures, the training of a flexible response to different accident situations. Experience has shown that the design provisions for safety could prevent dangerous release of the radioactive material to the environment after an accident has occurred. For future developments of reactor safety, extensive analyses of operating experience are of great importance. The main goal should be to enhance the reliability of measures for accident prevention, which prevent the core from meltdown or other damages

Methods and Effects of Safety Enhancement in Korean PSR

International Nuclear Information System (INIS)

Kim, Young Gab; Park, Jong Woon

2009-01-01

Periodic Safety Review (PSR) is a comprehensive study on a nuclear power plant safety, taking into account aspects such as operational history, ageing, safety analyses and advances in code and standards since the time of construction. In Korea, PSRs have been performed for 20 units and have been effectively used to obtain an overall view of actual plant safety to determine reasonable and practical modifications that should be made in order to obtain a higher level of safety approaching that of modern plants. Among many safety enhancements achieved from Korean PSRs, new safety analyses are the important methods to confirm plant safety by increasing safety margin for specific safety issues. Methods and effects of safety enhancements applied in Korean PSRs are reviewed in this paper in light of new safety analyses to obtain additional safety margins

Toward a Safety Risk-Based Classification of Unmanned Aircraft

Science.gov (United States)

Torres-Pomales, Wilfredo

2016-01-01

There is a trend of growing interest and demand for greater access of unmanned aircraft (UA) to the National Airspace System (NAS) as the ongoing development of UA technology has created the potential for significant economic benefits. However, the lack of a comprehensive and efficient UA regulatory framework has constrained the number and kinds of UA operations that can be performed. This report presents initial results of a study aimed at defining a safety-risk-based UA classification as a plausible basis for a regulatory framework for UA operating in the NAS. Much of the study up to this point has been at a conceptual high level. The report includes a survey of contextual topics, analysis of safety risk considerations, and initial recommendations for a risk-based approach to safe UA operations in the NAS. The next phase of the study will develop and leverage deeper clarity and insight into practical engineering and regulatory considerations for ensuring that UA operations have an acceptable level of safety.

Safety; Avertissement

Energy Technology Data Exchange (ETDEWEB)

NONE

2001-07-01

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.)

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.

Qualitative safety analysis in accelerator based systems

International Nuclear Information System (INIS)

Sarkar, P.K.; Chowdhury, Lekha M.

2006-01-01

In recent developments connected to high energy and high current accelerators, the accelerator driven systems (ADS) and the Radioactive Ion Beam (RIB) facilities come in the forefront of application. For medical and industrial applications high current accelerators often need to be located in populated areas. These facilities pose significant radiological hazard during their operation and accidental situations. We have done a qualitative evaluation of radiological safety analysis using the probabilistic safety analysis (PSA) methods for accelerator-based systems. The major contribution to hazard comes from a target rupture scenario in both ADS and RIB facilities. Other significant contributors to hazard in the facilities are also discussed using fault tree and event tree methodologies. (author)

Formal safety assessment based on relative risks model in ship navigation

Energy Technology Data Exchange (ETDEWEB)

Hu Shenping [Merchant Marine College, Shanghai Maritime University, 1550, Pudong Dadao, Shanghai 200135 (China)]. E-mail: sphu@mmc.shmtu.edu.cn; Fang Quangen [Merchant Marine College, Shanghai Maritime University, 1550, Pudong Dadao, Shanghai 200135 (China)]. E-mail: qgfang@mmc.shmtu.edu.cn; Xia Haibo [Merchant Marine College, Shanghai Maritime University, 1550, Pudong Dadao, Shanghai 200135 (China)]. E-mail: hbxia@mmc.shmtu.edu.cn; Xi Yongtao [Merchant Marine College, Shanghai Maritime University, 1550, Pudong Dadao, Shanghai 200135 (China)]. E-mail: xiyt@mmc.shmtu.edu.cn

2007-03-15

Formal safety assessment (FSA) is a structured and systematic methodology aiming at enhancing maritime safety. It has been gradually and broadly used in the shipping industry nowadays around the world. On the basis of analysis and conclusion of FSA approach, this paper discusses quantitative risk assessment and generic risk model in FSA, especially frequency and severity criteria in ship navigation. Then it puts forward a new model based on relative risk assessment (MRRA). The model presents a risk-assessment approach based on fuzzy functions and takes five factors into account, including detailed information about accident characteristics. It has already been used for the assessment of pilotage safety in Shanghai harbor, China. Consequently, it can be proved that MRRA is a useful method to solve the problems in the risk assessment of ship navigation safety in practice.

Formal safety assessment based on relative risks model in ship navigation

International Nuclear Information System (INIS)

Hu Shenping; Fang Quangen; Xia Haibo; Xi Yongtao

2007-01-01

Formal safety assessment (FSA) is a structured and systematic methodology aiming at enhancing maritime safety. It has been gradually and broadly used in the shipping industry nowadays around the world. On the basis of analysis and conclusion of FSA approach, this paper discusses quantitative risk assessment and generic risk model in FSA, especially frequency and severity criteria in ship navigation. Then it puts forward a new model based on relative risk assessment (MRRA). The model presents a risk-assessment approach based on fuzzy functions and takes five factors into account, including detailed information about accident characteristics. It has already been used for the assessment of pilotage safety in Shanghai harbor, China. Consequently, it can be proved that MRRA is a useful method to solve the problems in the risk assessment of ship navigation safety in practice

Implementing the Comprehensive Unit-Based Safety Program (CUSP) to Improve Patient Safety in an Academic Primary Care Practice.

Science.gov (United States)

Pitts, Samantha I; Maruthur, Nisa M; Luu, Ngoc-Phuong; Curreri, Kimberly; Grimes, Renee; Nigrin, Candace; Sateia, Heather F; Sawyer, Melinda D; Pronovost, Peter J; Clark, Jeanne M; Peairs, Kimberly S

2017-11-01

While there is growing awareness of the risk of harm in ambulatory health care, most patient safety efforts have focused on the inpatient setting. The Comprehensive Unit-based Safety Program (CUSP) has been an integral part of highly successful safety efforts in inpatient settings. In 2014 CUSP was implemented in an academic primary care practice. As part of CUSP implementation, staff and clinicians underwent training on the science of safety and completed a two-question safety assessment survey to identify safety concerns in the practice. The concerns identified by team members were used to select two initial safety priorities. The impact of CUSP on safety climate and teamwork was assessed through a pre-post comparison of results on the validated Safety Attitudes Questionnaire. Ninety-six percent of staff completed science of safety training as part of CUSP implementation, and 100% of staff completed the two-question safety assessment. The most frequently identified safety concerns were related to medications (n = 11, 28.2), diagnostic testing (n = 9, 25), and communication (n = 5, 14). The CUSP team initially prioritized communication and infection control, which led to standardization of work flows within the practice. Six months following CUSP implementation, large but nonstatistically significant increases were found for the percentage of survey respondents who reported knowledge of the proper channels for questions about patient safety, felt encouraged to report safety concerns, and believed that the work setting made it easy to learn from the errors of others. CUSP is a promising tool to improve safety climate and to identify and address safety concerns within ambulatory health care. Copyright © 2017 The Joint Commission. Published by Elsevier Inc. All rights reserved.

Safety climate and culture: Integrating psychological and systems perspectives.

Science.gov (United States)

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

2017-07-01

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

Functional Safety Specification of Communication Profile PROFIsafe

Directory of Open Access Journals (Sweden)

Jan Rofar

2006-01-01

Full Text Available Paper maps the trends in area of safety-related communication within PROFIBUS and PROFINET industry networks. There are analyses safety measures and Fail-safe parameters of PROFIsafe profile in version V2 and their localisation in Safety Communication Layer SCL, which guarantees Safety Integrity Level SIL according to standard IEC 61508. The last chapter analyses the reaction in the event of fault during transmission of messages.

Balancing safety and economics

International Nuclear Information System (INIS)

Kroeger, W.; Fischer, P.U.

2000-01-01

The safety requirements of NPPs have always aimed at limiting societal risks. This risk approach initially resulted in deterministic design criteria and concepts. In the 1980s the paradigm 'safety at all costs' arose and often led to questionable backfitting measures. Conflicts between new requirements, classical design concepts and operational demands were often ignored. The design requirements for advanced reactors ensure enhanced protection against severe accidents. Still, it is questionable whether the 'no-damage-outside-the-fence' criteria can be achieved deterministically and at competitive costs. Market deregulation and utility privatisation call for a balance between safety and costs, without jeopardising basic safety concepts. An ideal approach must be risk-based and imply modern PSAs and new methods for cost-benefit and ALARA analyses, embed nuclear risks in a wider risk spectrum, but also make benefits transparent within the context of a broader life experience. Governments should define basic requirements, minimum standards and consistent comparison criteria, and strengthen operator responsibility. Internationally sufficient and binding safety requirements must be established and nuclear technology transfer handled in a responsible way, while existing plants, with their continuous backfitting investments, should receive particular attention. (orig.)

Implementing evidence-based policy in a network setting: road safety policy in the Netherlands.

Science.gov (United States)

Bax, Charlotte; de Jong, Martin; Koppenjan, Joop

2010-01-01

In the early 1990s, in order to improve road safety in The Netherlands, the Institute for Road Safety Research (SWOV) developed an evidence-based "Sustainable Safety" concept. Based on this concept, Dutch road safety policy, was seen as successful and as a best practice in Europe. In The Netherlands, the policy context has now changed from a sectoral policy setting towards a fragmented network in which safety is a facet of other transport-related policies. In this contribution, it is argued that the implementation strategy underlying Sustainable Safety should be aligned with the changed context. In order to explore the adjustments needed, two perspectives of policy implementation are discussed: (1) national evidence-based policies with sectoral implementation; and (2) decentralized negotiation on transport policy in which road safety is but one aspect. We argue that the latter approach matches the characteristics of the newly evolved policy context best, and conclude with recommendations for reformulating the implementation strategy.

Safety applications of computer based systems for the process industry

International Nuclear Information System (INIS)

Bologna, Sandro; Picciolo, Giovanni; Taylor, Robert

1997-11-01

Computer based systems, generally referred to as Programmable Electronic Systems (PESs) are being increasingly used in the process industry, also to perform safety functions. The process industry as they intend in this document includes, but is not limited to, chemicals, oil and gas production, oil refining and power generation. Starting in the early 1970's the wide application possibilities and the related development problems of such systems were recognized. Since then, many guidelines and standards have been developed to direct and regulate the application of computers to perform safety functions (EWICS-TC7, IEC, ISA). Lessons learnt in the last twenty years can be summarised as follows: safety is a cultural issue; safety is a management issue; safety is an engineering issue. In particular, safety systems can only be properly addressed in the overall system context. No single method can be considered sufficient to achieve the safety features required in many safety applications. Good safety engineering approach has to address not only hardware and software problems in isolation but also their interfaces and man-machine interface problems. Finally, the economic and industrial aspects of the safety applications and development of PESs in process plants are evidenced throughout all the Report. Scope of the Report is to contribute to the development of an adequate awareness of these problems and to illustrate technical solutions applied or being developed

An Adaptive Information Quantity-Based Broadcast Protocol for Safety Services in VANET

Directory of Open Access Journals (Sweden)

Wenjie Wang

2016-01-01

Full Text Available Vehicle-to-vehicle communication plays a significantly important role in implementing safe and efficient road traffic. When disseminating safety messages in the network, the information quantity on safety packets changes over time and space. However, most of existing protocols view each packet the same to disseminate, preventing vehicles from collecting more recent and precise safety information. Hence, an information quantity-based broadcast protocol is proposed in this paper to ensure the efficiency of safety messages dissemination. In particular, we propose the concept of emergency-degree to evaluate packets’ information quantity. Then we present EDCast, an emergency-degree-based broadcast protocol. EDCast differentiates each packet’s priority for accessing the channel based on its emergency-degree so as to provide vehicles with more safety information timely and accurately. In addition, an adaptive scheme is presented to ensure fast dissemination of messages in different network condition. We compare the performance of EDCast with those of three other representative protocols in a typical highway scenario. Simulation results indicate that EDCast achieves higher broadcast efficiency and less redundancy with less delivery delay. What we found demonstrates that it is feasible and necessary for incorporating information quantity of messages in designing an efficient safety message broadcast protocol.

Drug safety data mining with a tree-based scan statistic.

Science.gov (United States)

Kulldorff, Martin; Dashevsky, Inna; Avery, Taliser R; Chan, Arnold K; Davis, Robert L; Graham, David; Platt, Richard; Andrade, Susan E; Boudreau, Denise; Gunter, Margaret J; Herrinton, Lisa J; Pawloski, Pamala A; Raebel, Marsha A; Roblin, Douglas; Brown, Jeffrey S

2013-05-01

In post-marketing drug safety surveillance, data mining can potentially detect rare but serious adverse events. Assessing an entire collection of drug-event pairs is traditionally performed on a predefined level of granularity. It is unknown a priori whether a drug causes a very specific or a set of related adverse events, such as mitral valve disorders, all valve disorders, or different types of heart disease. This methodological paper evaluates the tree-based scan statistic data mining method to enhance drug safety surveillance. We use a three-million-member electronic health records database from the HMO Research Network. Using the tree-based scan statistic, we assess the safety of selected antifungal and diabetes drugs, simultaneously evaluating overlapping diagnosis groups at different granularity levels, adjusting for multiple testing. Expected and observed adverse event counts were adjusted for age, sex, and health plan, producing a log likelihood ratio test statistic. Out of 732 evaluated disease groupings, 24 were statistically significant, divided among 10 non-overlapping disease categories. Five of the 10 signals are known adverse effects, four are likely due to confounding by indication, while one may warrant further investigation. The tree-based scan statistic can be successfully applied as a data mining tool in drug safety surveillance using observational data. The total number of statistical signals was modest and does not imply a causal relationship. Rather, data mining results should be used to generate candidate drug-event pairs for rigorous epidemiological studies to evaluate the individual and comparative safety profiles of drugs. Copyright © 2013 John Wiley & Sons, Ltd.

Model-based Recursive Partitioning for Subgroup Analyses

OpenAIRE

Seibold, Heidi; Zeileis, Achim; Hothorn, Torsten

2016-01-01

The identification of patient subgroups with differential treatment effects is the first step towards individualised treatments. A current draft guideline by the EMA discusses potentials and problems in subgroup analyses and formulated challenges to the development of appropriate statistical procedures for the data-driven identification of patient subgroups. We introduce model-based recursive partitioning as a procedure for the automated detection of patient subgroups that are identifiable by...

Prediction of main factors’ values of air transportation system safety based on system dynamics

Science.gov (United States)

Spiridonov, A. Yu; Rezchikov, A. F.; Kushnikov, V. A.; Ivashchenko, V. A.; Bogomolov, A. S.; Filimonyuk, L. Yu; Dolinina, O. N.; Kushnikova, E. V.; Shulga, T. E.; Tverdokhlebov, V. A.; Kushnikov, O. V.; Fominykh, D. S.

2018-05-01

On the basis of the system-dynamic approach [1-8], a set of models has been developed that makes it possible to analyse and predict the values of the main safety indicators for the operation of aviation transport systems.

Computational methods for nuclear criticality safety analysis

International Nuclear Information System (INIS)

Maragni, M.G.

1992-01-01

Nuclear criticality safety analyses require the utilization of methods which have been tested and verified against benchmarks results. In this work, criticality calculations based on the KENO-IV and MCNP codes are studied aiming the qualification of these methods at the IPEN-CNEN/SP and COPESP. The utilization of variance reduction techniques is important to reduce the computer execution time, and several of them are analysed. As practical example of the above methods, a criticality safety analysis for the storage tubes for irradiated fuel elements from the IEA-R1 research has been carried out. This analysis showed that the MCNP code is more adequate for problems with complex geometries, and the KENO-IV code shows conservative results when it is not used the generalized geometry option. (author)

Patterns of use and impact of standardised MedDRA query analyses on the safety evaluation and review of new drug and biologics license applications.

Directory of Open Access Journals (Sweden)

Lin-Chau Chang

Full Text Available Standardised MedDRA Queries (SMQs have been developed since the early 2000's and used by academia, industry, public health, and government sectors for detecting safety signals in adverse event safety databases. The purpose of the present study is to characterize how SMQs are used and the impact in safety analyses for New Drug Application (NDA and Biologics License Application (BLA submissions to the United States Food and Drug Administration (USFDA.We used the PharmaPendium database to capture SMQ use in Summary Basis of Approvals (SBoAs of drugs and biologics approved by the USFDA. Characteristics of the drugs and the SMQ use were employed to evaluate the role of SMQ safety analyses in regulatory decisions and the veracity of signals they revealed.A comprehensive search of the SBoAs yielded 184 regulatory submissions approved from 2006 to 2015. Search strategies more frequently utilized restrictive searches with "narrow terms" to enhance specificity over strategies using "broad terms" to increase sensitivity, while some involved modification of search terms. A majority (59% of 1290 searches used descriptive statistics, however inferential statistics were utilized in 35% of them. Commentary from reviewers and supervisory staff suggested that a small, yet notable percentage (18% of 1290 searches supported regulatory decisions. The searches with regulatory impact were found in 73 submissions (40% of the submissions investigated. Most searches (75% of 227 searches with regulatory implications described how the searches were confirmed, indicating prudence in the decision-making process.SMQs have an increasing role in the presentation and review of safety analysis for NDAs/BLAs and their regulatory reviews. This study suggests that SMQs are best used for screening process, with descriptive statistics, description of SMQ modifications, and systematic verification of cases which is crucial for drawing regulatory conclusions.

Patterns of use and impact of standardised MedDRA query analyses on the safety evaluation and review of new drug and biologics license applications.

Science.gov (United States)

Chang, Lin-Chau; Mahmood, Riaz; Qureshi, Samina; Breder, Christopher D

2017-01-01

Standardised MedDRA Queries (SMQs) have been developed since the early 2000's and used by academia, industry, public health, and government sectors for detecting safety signals in adverse event safety databases. The purpose of the present study is to characterize how SMQs are used and the impact in safety analyses for New Drug Application (NDA) and Biologics License Application (BLA) submissions to the United States Food and Drug Administration (USFDA). We used the PharmaPendium database to capture SMQ use in Summary Basis of Approvals (SBoAs) of drugs and biologics approved by the USFDA. Characteristics of the drugs and the SMQ use were employed to evaluate the role of SMQ safety analyses in regulatory decisions and the veracity of signals they revealed. A comprehensive search of the SBoAs yielded 184 regulatory submissions approved from 2006 to 2015. Search strategies more frequently utilized restrictive searches with "narrow terms" to enhance specificity over strategies using "broad terms" to increase sensitivity, while some involved modification of search terms. A majority (59%) of 1290 searches used descriptive statistics, however inferential statistics were utilized in 35% of them. Commentary from reviewers and supervisory staff suggested that a small, yet notable percentage (18%) of 1290 searches supported regulatory decisions. The searches with regulatory impact were found in 73 submissions (40% of the submissions investigated). Most searches (75% of 227 searches) with regulatory implications described how the searches were confirmed, indicating prudence in the decision-making process. SMQs have an increasing role in the presentation and review of safety analysis for NDAs/BLAs and their regulatory reviews. This study suggests that SMQs are best used for screening process, with descriptive statistics, description of SMQ modifications, and systematic verification of cases which is crucial for drawing regulatory conclusions.

A progressive methodology for seismic safety evaluation of gravity dams

International Nuclear Information System (INIS)

Ghrib, F.; Leger, P.; Tinawi, R.; Lupien, R.; Veilleux, M.

1995-01-01

A progressive methodology for the seismic safety evaluation of existing concrete gravity dams was described. The methodology was based on five structural analysis levels with increasing complexity to represent inertia forces, dam-foundation and dam-interaction mechanisms, as well as concrete cracking. The five levels were (1) preliminary screening, (2) pseudo-static method, (3) pseudo-dynamic method, (4) linear time history analysis, and (5) non-linear history analysis. The first four levels of analysis were applied for the seismic safety evaluation of Paugan gravity dam (Quebec). Results showed that internal forces from pseudo-dynamic, response spectra and transient finite element analyses could be used to interpret the dynamic stability of dams from familiar strength-based criteria. However, as soon as the base was cracked, the seismically induced forces were modified, and level IV analyses proved more suitable to handle rationally these complexities. 8 refs., 7 figs., 1 tab

Operation safety of control systems. Principles and methods

International Nuclear Information System (INIS)

Aubry, J.F.; Chatelet, E.

2008-01-01

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

Systematic review of economic analyses in patient safety: a protocol designed to measure development in the scope and quality of evidence.

Science.gov (United States)

Carter, Alexander W; Mandavia, Rishi; Mayer, Erik; Marti, Joachim; Mossialos, Elias; Darzi, Ara

2017-08-18

Recent avoidable failures in patient care highlight the ongoing need for evidence to support improvements in patient safety. According to the most recent reviews, there is a dearth of economic evidence related to patient safety. These reviews characterise an evidence gap in terms of the scope and quality of evidence available to support resource allocation decisions. This protocol is designed to update and improve on the reviews previously conducted to determine the extent of methodological progress in economic analyses in patient safety. A broad search strategy with two core themes for original research (excluding opinion pieces and systematic reviews) in 'patient safety' and 'economic analyses' has been developed. Medline, Econlit and National Health Service Economic Evaluation Database bibliographic databases will be searched from January 2007 using a combination of medical subject headings terms and research-derived search terms (see table 1). The method is informed by previous reviews on this topic, published in 2012. Screening, risk of bias assessment (using the Cochrane collaboration tool) and economic evaluation quality assessment (using the Drummond checklist) will be conducted by two independent reviewers, with arbitration by a third reviewer as needed. Studies with a low risk of bias will be assessed using the Drummond checklist. High-quality economic evaluations are those that score >20/35. A qualitative synthesis of evidence will be performed using a data collection tool to capture the study design(s) employed, population(s), setting(s), disease area(s), intervention(s) and outcome(s) studied. Methodological quality scores will be compared with previous reviews where possible. Effect size(s) and estimate uncertainty will be captured and used in a quantitative synthesis of high-quality evidence, where possible. Formal ethical approval is not required as primary data will not be collected. The results will be disseminated through a peer

Causal Relationship Analysis of the Patient Safety Culture Based on Safety Attitudes Questionnaire in Taiwan

Science.gov (United States)

Zeng, Pei-Shan; Huang, Chih-Hsuan

2018-01-01

This study uses the decision-making trial and evaluation laboratory method to identify critical dimensions of the safety attitudes questionnaire in Taiwan in order to improve the patient safety culture from experts' viewpoints. Teamwork climate, stress recognition, and perceptions of management are three causal dimensions, while safety climate, job satisfaction, and working conditions are receiving dimensions. In practice, improvements on effect-based dimensions might receive little effects when a great amount of efforts have been invested. In contrast, improving a causal dimension not only improves itself but also results in better performance of other dimension(s) directly affected by this particular dimension. Teamwork climate and perceptions of management are found to be the most critical dimensions because they are both causal dimensions and have significant influences on four dimensions apiece. It is worth to note that job satisfaction is the only dimension affected by the other dimensions. In order to effectively enhance the patient safety culture for healthcare organizations, teamwork climate, and perceptions of management should be closely monitored. PMID:29686825

Causal Relationship Analysis of the Patient Safety Culture Based on Safety Attitudes Questionnaire in Taiwan

Directory of Open Access Journals (Sweden)

Yii-Ching Lee

2018-01-01

Full Text Available This study uses the decision-making trial and evaluation laboratory method to identify critical dimensions of the safety attitudes questionnaire in Taiwan in order to improve the patient safety culture from experts’ viewpoints. Teamwork climate, stress recognition, and perceptions of management are three causal dimensions, while safety climate, job satisfaction, and working conditions are receiving dimensions. In practice, improvements on effect-based dimensions might receive little effects when a great amount of efforts have been invested. In contrast, improving a causal dimension not only improves itself but also results in better performance of other dimension(s directly affected by this particular dimension. Teamwork climate and perceptions of management are found to be the most critical dimensions because they are both causal dimensions and have significant influences on four dimensions apiece. It is worth to note that job satisfaction is the only dimension affected by the other dimensions. In order to effectively enhance the patient safety culture for healthcare organizations, teamwork climate, and perceptions of management should be closely monitored.

Report of a consultants meeting on backfittings and safety enhancement measures in NPPs with WWER 440/213 reactors. Extrabudgetary programme on the safety of WWER NPPS

International Nuclear Information System (INIS)

1994-01-01

The purpose of this Consultants' Meeting held by the IAEA in Vienna from 11-15 April 1994 within the framework of the Extrabudgetary Programme on WWER Safety was to review and analyze safety issues revealed during operation and through analyses of NPPs with WWER 440/213 reactors. The initial list of safety issues based on the available reports from various studies had been prepared by the IAEA secretariat before the meeting, together with indications of safety enhancement measures proposed in various NPP units. During the meeting, the underlying safety concerns and actual technical status of the plants were discussed and the ranking of the safety issues was considered. 58 refs, 1 tab

A root cause analysis project in a medication safety course.

Science.gov (United States)

Schafer, Jason J

2012-08-10

To develop, implement, and evaluate team-based root cause analysis projects as part of a required medication safety course for second-year pharmacy students. Lectures, in-class activities, and out-of-class reading assignments were used to develop students' medication safety skills and introduce them to the culture of medication safety. Students applied these skills within teams by evaluating cases of medication errors using root cause analyses. Teams also developed error prevention strategies and formally presented their findings. Student performance was assessed using a medication errors evaluation rubric. Of the 211 students who completed the course, the majority performed well on root cause analysis assignments and rated them favorably on course evaluations. Medication error evaluation and prevention was successfully introduced in a medication safety course using team-based root cause analysis projects.

Safety management of an underground-based gravitational wave telescope: KAGRA

Science.gov (United States)

Ohishi, Naoko; Miyoki, Shinji; Uchiyama, Takashi; Miyakawa, Osamu; Ohashi, Masatake

2014-08-01

KAGRA is a unique gravitational wave telescope with its location underground and use of cryogenic mirrors. Safety management plays an important role for secure development and operation of such a unique and large facility. Based on relevant law in Japan, Labor Standard Act and Industrial Safety and Health Law, various countermeasures are mandated to avoid foreseeable accidents and diseases. In addition to the usual safety management of hazardous materials, such as cranes, organic solvents, lasers, there are specific safety issues in the tunnel. Prevention of collapse, flood, and fire accidents are the most critical issues for the underground facility. Ventilation is also important for prevention of air pollution by carbon monoxide, carbon dioxide, organic solvents and radon. Oxygen deficiency should also be prevented.

The role of CFD combustion modeling in hydrogen safety management-II: Validation based on homogeneous hydrogen-air experiments

Energy Technology Data Exchange (ETDEWEB)

Sathiah, Pratap, E-mail: sathiah@nrg.eu [Nuclear Research and Consultancy Group (NRG), Westerduinweg 3, 1755 ZG Petten (Netherlands); Haren, Steven van, E-mail: vanharen@nrg.eu [Nuclear Research and Consultancy Group (NRG), Westerduinweg 3, 1755 ZG Petten (Netherlands); Komen, Ed, E-mail: komen@nrg.eu [Nuclear Research and Consultancy Group (NRG), Westerduinweg 3, 1755 ZG Petten (Netherlands); Roekaerts, Dirk, E-mail: d.j.e.m.roekaerts@tudelft.nl [Department of Multi-Scale Physics, Delft University of Technology, P.O. Box 5, 2600 AA Delft (Netherlands)

2012-11-15

Highlights: Black-Right-Pointing-Pointer A CFD based method is proposed for the simulation of hydrogen deflagration. Black-Right-Pointing-Pointer A dynamic grid adaptation method is proposed to resolve turbulent flame brush thickness. Black-Right-Pointing-Pointer The predictions obtained using this method is in good agreement with the static grid method. Black-Right-Pointing-Pointer TFC model results are in good agreement with large-scale homogeneous hydrogen-air experiments. - Abstract: During a severe accident in a PWR, large quantities of hydrogen can be generated and released into the containment. The generated hydrogen, when mixed with air, can lead to hydrogen combustion. The dynamic pressure loads resulting from hydrogen combustion can be detrimental to the structural integrity of the reactor safety systems and the reactor containment. Therefore, accurate prediction of these pressure loads is an important safety issue. In a previous article, we presented a CFD based method to determine these pressure loads. This CFD method is based on the application of a turbulent flame speed closure combustion model. The validation analyses in our previous paper demonstrated that it is of utmost importance to apply successive mesh and time step refinement in order to get reliable results. In this article, we first determined to what extent the required computational effort required for our CFD approach can be reduced by the application of adaptive mesh refinement, while maintaining the accuracy requirements. Experiments performed within a small fan stirred explosion bomb were used for this purpose. It could be concluded that adaptive grid adaptation is a reliable and efficient method for usage in hydrogen deflagration analyses. For the two-dimensional validation analyses, the application of dynamic grid adaptation resulted in a reduction of the required computational effort by about one order of magnitude. In a second step, the considered CFD approach including adaptive

Safety features of TR-2 reactor

International Nuclear Information System (INIS)

Tuerker, T.

2001-01-01

TR-2 is a swimming pool type research reactor with 5 MW thermal power and uses standard MTR plate type fuel elements. Each standard fuel element consist of 23 fuel plates with a meat + cladding thickness of 0.127 cm, coolant channel clearance is 0.21 cm. Originally TR-2 is designed for %93 enriched U-Al. Alloy fuel meat.This work is based on the preparation of the Final Safety Analyses Report (FSAR) of the TR-2 reactor. The main aspect is to investigate the behaviour of TR-2 reactor under the accident and abnormal operating conditions, which cowers the accident spectrum unique for the TR-2 reactor. This presentation covers some selected transient analyses which are important for the safety aspects of the TR-2 reactor like reactivity induced startup accidents, pump coast down (Loss of Flow Accident, LOFA) and other accidents which are charecteristic to the TR-2

Phenomenological analyses and their application to the Defense Waste Processing Facility probabilistic safety analysis accident progression event tree. Revision 1

International Nuclear Information System (INIS)

Kalinich, D.A.; Thomas, J.K.; Gough, S.T.; Bailey, R.T.; Kearnaghan, D.P.

1995-01-01

In the Defense Waste Processing Facility (DWPF) Safety Analysis Reports (SARs) for the Savannah River Site (SRS), risk-based perspectives have been included per US Department of Energy (DOE) Order 5480.23. The NUREG-1150 Level 2/3 Probabilistic Risk Assessment (PRA) methodology was selected as the basis for calculating facility risk. The backbone of this methodology is the generation of an Accident Progression Event Tree (APET), which is solved using the EVNTRE computer code. To support the development of the DWPF APET, deterministic modeling of accident phenomena was necessary. From these analyses, (1) accident progressions were identified for inclusion into the APET; (2) branch point probabilities and any attendant parameters were quantified; and (3) the radionuclide releases to the environment from accidents were determined. The phenomena of interest for accident progressions included explosions, fires, a molten glass spill, and the response of the facility confinement system during such challenges. A variety of methodologies, from hand calculations to large system-model codes, were used in the evaluation of these phenomena

77 FR 76003 - Submission for OMB Review; Comment Request-Safety Standard for Omnidirectional Citizens Band Base...

Science.gov (United States)

2012-12-26

... Request--Safety Standard for Omnidirectional Citizens Band Base Station Antennas AGENCY: Consumer Product... information associated with the Commission's safety standard for omnidirectional citizens band base station... information required in the Safety Standard for Omnidirectional Citizens Band Base Station (16 CFR Part 1204...

Safety assessment of automated vehicle functions by simulation-based fault injection

OpenAIRE

Juez, Garazi; Amparan, Estibaliz; Lattarulo, Ray; Rastelli, Joshue Perez; Ruiz, Alejandra; Espinoza, Huascar

2017-01-01

As automated driving vehicles become more sophisticated and pervasive, it is increasingly important to assure its safety even in the presence of faults. This paper presents a simulation-based fault injection approach (Sabotage) aimed at assessing the safety of automated vehicle functions. In particular, we focus on a case study to forecast fault effects during the model-based design of a lateral control function. The goal is to determine the acceptable fault detection interval for pe...

Bayesian-network-based safety risk analysis in construction projects

International Nuclear Information System (INIS)

Zhang, Limao; Wu, Xianguo; Skibniewski, Miroslaw J.; Zhong, Jingbing; Lu, Yujie

2014-01-01

This paper presents a systemic decision support approach for safety risk analysis under uncertainty in tunnel construction. Fuzzy Bayesian Networks (FBN) is used to investigate causal relationships between tunnel-induced damage and its influential variables based upon the risk/hazard mechanism analysis. Aiming to overcome limitations on the current probability estimation, an expert confidence indicator is proposed to ensure the reliability of the surveyed data for fuzzy probability assessment of basic risk factors. A detailed fuzzy-based inference procedure is developed, which has a capacity of implementing deductive reasoning, sensitivity analysis and abductive reasoning. The “3σ criterion” is adopted to calculate the characteristic values of a triangular fuzzy number in the probability fuzzification process, and the α-weighted valuation method is adopted for defuzzification. The construction safety analysis progress is extended to the entire life cycle of risk-prone events, including the pre-accident, during-construction continuous and post-accident control. A typical hazard concerning the tunnel leakage in the construction of Wuhan Yangtze Metro Tunnel in China is presented as a case study, in order to verify the applicability of the proposed approach. The results demonstrate the feasibility of the proposed approach and its application potential. A comparison of advantages and disadvantages between FBN and fuzzy fault tree analysis (FFTA) as risk analysis tools is also conducted. The proposed approach can be used to provide guidelines for safety analysis and management in construction projects, and thus increase the likelihood of a successful project in a complex environment. - Highlights: • A systemic Bayesian network based approach for safety risk analysis is developed. • An expert confidence indicator for probability fuzzification is proposed. • Safety risk analysis progress is extended to entire life cycle of risk-prone events. • A typical

Fusion safety codes International modeling with MELCOR and ATHENA- INTRA

CERN Document Server

Marshall, T; Topilski, L; Merrill, B

2002-01-01

For a number of years, the world fusion safety community has been involved in benchmarking their safety analyses codes against experiment data to support regulatory approval of a next step fusion device. This paper discusses the benchmarking of two prominent fusion safety thermal-hydraulic computer codes. The MELCOR code was developed in the US for fission severe accident safety analyses and has been modified for fusion safety analyses. The ATHENA code is a multifluid version of the US-developed RELAP5 code that is also widely used for fusion safety analyses. The ENEA Fusion Division uses ATHENA in conjunction with the INTRA code for its safety analyses. The INTRA code was developed in Germany and predicts containment building pressures, temperatures and fluid flow. ENEA employs the French-developed ISAS system to couple ATHENA and INTRA. This paper provides a brief introduction of the MELCOR and ATHENA-INTRA codes and presents their modeling results for the following breaches of a water cooling line into the...

Probabilistic safety assessment for research reactors

International Nuclear Information System (INIS)

1986-12-01

Increasing interest in using Probabilistic Safety Assessment (PSA) methods for research reactor safety is being observed in many countries throughout the world. This is mainly because of the great ability of this approach in achieving safe and reliable operation of research reactors. There is also a need to assist developing countries to apply Probabilistic Safety Assessment to existing nuclear facilities which are simpler and therefore less complicated to analyse than a large Nuclear Power Plant. It may be important, therefore, to develop PSA for research reactors. This might also help to better understand the safety characteristics of the reactor and to base any backfitting on a cost-benefit analysis which would ensure that only necessary changes are made. This document touches on all the key aspects of PSA but placed greater emphasis on so-called systems analysis aspects rather than the in-plant or ex-plant consequences

Early Safety Assessment of Automotive Systems Using Sabotage Simulation-Based Fault Injection Framework

OpenAIRE

Juez, Garazi; Amparan, Estíbaliz; Lattarulo, Ray; Ruíz, Alejandra; Perez, Joshue; Espinoza, Huascar

2017-01-01

As road vehicles increase their autonomy and the driver reduces his role in the control loop, novel challenges on dependability assessment arise. Model-based design combined with a simulation-based fault injection technique and a virtual vehicle poses as a promising solution for an early safety assessment of automotive systems. To start with, the design, where no safety was considered, is stimulated with a set of fault injection simulations (fault forecasting). By doing so, safety strategies ...

Simulation study of coal mine safety investment based on system dynamics

Institute of Scientific and Technical Information of China (English)

Tong Lei; Dou Yuanyuan

2014-01-01

To generate dynamic planning for coal mine safety investment, this study applies system dynamics to decision-making, classifying safety investments by accident type. It validates the relationship between safety investments and accident cost, by structurally analyzing the causality between safety investments and their influence factors. Our simulation model, based on Vensim software, conducts simulation anal-ysis on a series of actual data from a coalmine in Shanxi Province. Our results indicate a lag phase in safety investments, and that increasing pre-phase safety investment reduces accident costs. We found that a 24%increase in initial safety investment could help reach the target accident costs level 14 months earlier. Our simulation test included nine kinds of variation trends of accident costs brought by different investment ratios on accident prevention. We found an optimized ratio of accident prevention invest-ments allowing a mine to reach accident cost goals 4 months earlier, without changing its total investment.

Are area-based initiatives able to improve area safety in deprived areas? A quasi-experimental evaluation of the Dutch District Approach.

Science.gov (United States)

Kramer, Daniëlle; Jongeneel-Grimen, Birthe; Stronks, Karien; Droomers, Mariël; Kunst, Anton E

2015-07-28

Numerous area-based initiatives have been implemented in deprived areas across Western-Europe with the aim to improve the socio-economic and environmental conditions in these areas. Only few of these initiatives have been scientifically evaluated for their impact on key social determinants of health, like perceived area safety. Therefore, this study aimed to assess the impact of a Dutch area-based initiative called the District Approach on trends in perceived area safety and underlying problems in deprived target districts. A quasi-experimental design was used. Repeated cross-sectional data on perceived area safety and underlying problems were obtained from the National Safety Monitor (2005-2008) and its successor the Integrated Safety Monitor (2008-2011). Study population consisted of 133,522 Dutch adults, including 3,595 adults from target districts. Multilevel logistic regression analyses were performed to assess trends in self-reported general safety, physical order, social order, and non-victimization before and after the start of the District Approach mid-2008. Trends in target districts were compared with trends in various control groups. Residents of target districts felt less safe, perceived less physical and social order, and were victimized more often than adults elsewhere in the Netherlands. For non-victimization, target districts showed a somewhat more positive change in trend after the start of the District Approach than the rest of the Netherlands or other deprived districts. Differences were only statistically significant in women, older adults, and lower educated adults. For general safety, physical order, and social order, there were no differences in trend change between target districts and control groups. Results suggest that the District Approach has been unable to improve perceptions of area safety and disorder in deprived areas, but that it did result in declining victimization rates.

Seismic fragility analyses

International Nuclear Information System (INIS)

Kostov, Marin

2000-01-01

In the last two decades there is increasing number of probabilistic seismic risk assessments performed. The basic ideas of the procedure for performing a Probabilistic Safety Analysis (PSA) of critical structures (NUREG/CR-2300, 1983) could be used also for normal industrial and residential buildings, dams or other structures. The general formulation of the risk assessment procedure applied in this investigation is presented in Franzini, et al., 1984. The probability of failure of a structure for an expected lifetime (for example 50 years) can be obtained from the annual frequency of failure, β E determined by the relation: β E ∫[d[β(x)]/dx]P(flx)dx. β(x) is the annual frequency of exceedance of load level x (for example, the variable x may be peak ground acceleration), P(fI x) is the conditional probability of structure failure at a given seismic load level x. The problem leads to the assessment of the seismic hazard β(x) and the fragility P(fl x). The seismic hazard curves are obtained by the probabilistic seismic hazard analysis. The fragility curves are obtained after the response of the structure is defined as probabilistic and its capacity and the associated uncertainties are assessed. Finally the fragility curves are combined with the seismic loading to estimate the frequency of failure for each critical scenario. The frequency of failure due to seismic event is presented by the scenario with the highest frequency. The tools usually applied for probabilistic safety analyses of critical structures could relatively easily be adopted to ordinary structures. The key problems are the seismic hazard definitions and the fragility analyses. The fragility could be derived either based on scaling procedures or on the base of generation. Both approaches have been presented in the paper. After the seismic risk (in terms of failure probability) is assessed there are several approaches for risk reduction. Generally the methods could be classified in two groups. The

Stakes and Solutions for current and up-coming Licensing Challenges in PWR and BWR Reload and Safety Analysis

Energy Technology Data Exchange (ETDEWEB)

Curca-Tiving, F.; Opel, S.

2014-07-01

Regulatory requirements for reloads and safety analyses are evolving: New safety criteria, requests for enlarged qualification databases, statistical applications, uncertainty propagation... In order to address these challenges and access more predictable licensing processes, AREVA implements a consistent code and methodology suite for PWR and BWR core design and safety analysis, based on a first principles modeling with an extremely broad international verification and validation data base. (Author)

A fuzzy-based model to implement the global safety buildings index assessment for agri-food buildings

Directory of Open Access Journals (Sweden)

Francesco Barreca

2014-06-01

Full Text Available The latest EU policies focus on the issue of food safety with a view to ensuring adequate and standard quality levels for the food produced and/or consumed within the EC. To that purpose, the environment where agricultural products are manufactured and processed plays a crucial role in achieving food hygiene. As a consequence, it is of the outmost importance to adopt proper building solutions which meet health and hygiene requirements as well as to use suitable tools to measure the levels achieved. Similarly, it is necessary to verify and evaluate the level of workers’ safety and welfare in their working environment. Workers’ safety has not only an ethical and social value but also an economic implication, since possible accidents or environmental stressors are the major causes of the lower efficiency and productivity of workers. Therefore, it is fundamental to design suitable models of analysis that allow assessing buildings as a whole, taking into account both health and hygiene safety as well as workers’ safety and welfare. Hence, this paper proposes an assessment model that, based on an established study protocol and on the application of a fuzzy logic procedure, allows assessing the global safety level of an agri-food building by means of a global safety buildings index. The model here presented is original since it uses fuzzy logic to evaluate the performances of both the technical and environmental systems of an agri-food building in terms of health and hygiene safety of the manufacturing process as well as of workers’ health and safety. The result of the assessment is expressed through a triangular fuzzy membership function which allows carrying out comparative analyses of different buildings. A specific procedure was developed to apply the model to a case study which tested its operational simplicity and the validity of its results. The proposed model allows obtaining a synthetic and global value of the building performance of

Exploring the Effects of Cultural Variables in the Implementation of Behavior-Based Safety in Two Organizations

Science.gov (United States)

Bumstead, Alaina; Boyce, Thomas E.

2005-01-01

The present case study examines how culture can influence behavior-based safety in different organizational settings and how behavior-based safety can impact different organizational cultures. Behavior-based safety processes implemented in two culturally diverse work settings are described. Specifically, despite identical implementation plans,…

Examination of the bases for proposed innovations in reactor safety technology

International Nuclear Information System (INIS)

Moses, D.L.

1986-01-01

This paper employs the criteria for evaluations from the Nuclear Power Option Viability Study to examine the bases for proposed innovations in light water reactor safety technology. These bases for innovation fall into four broad categories as follows: (1) virtually exclusive reliance on passive safety features to preclude core damage in all situations, (2) design simplification using some passive safety features to reduce the frequency of core damage to less than about 10 -6 per reactor-year, (3) passive containment to preclude releases from any accident, and (4) designing to limit licensing attention to one or at least a few systems. Of these, only the first two, and perhaps only the second, hold significant promise for providing for the viability of advanced light water reactors

Safety assessment as basis for the decision making process

International Nuclear Information System (INIS)

Ilie, P.; Didita, L.; Danchiv, A.

2005-01-01

This paper deals with the safety assessment for a new near surface repository, particularly for the early stage of repository development using ISAM (Improvement of Safety Assessment Methodologies for Near Surface Disposal Facilities) safety assessment methodology. In this stage of the repository life cycle the main purpose of the safety assessment is to demonstrate that the plant is capable to be constructed and operated safely. The paper is based on development of the ASAM (Application of the Safety Assessment Methodologies for Near-Surface Disposal Facilities) Decision Support Subgroup of the Common Aspects Working Group. The implications of decision making for the application of the ISAM methodology on post-closure safety assessment are analysed. Some important elements of the decision-making process with impact on key components of the ISAM process are described. Following the development of Decision Support Subgroup of the ASAM Common Aspects Working Group the proposed change of ISAM methodology is analysed. This approach puts all activities in a decision context where the first iteration of the safety assessment is based on the existing state of knowledge and the initial engineering design. Confidence in the process is accomplished through the direct inclusion of all decision makers and stakeholders in the formulation of decisions, the definition of the state of knowledge, and decision making activities. The decision process is developed in context of undertaking assessments with little site-specific information, this situation is specifically for new planned repository. Limited site-specific information can result in a high degree of uncertainty, therefore it is important first of all to identify the sources of uncertainty arising from the limited nature of the site-specific information and then to apply appropriate approaches to manage the uncertainties and to determine whether the uncertainties are important to the overall safety of the disposal facility

Economics of the specification 6M safety re-evaluation and regulatory requirements

International Nuclear Information System (INIS)

Hopper, C.M.

1985-01-01

The objective of this work was to examine the potential economic impact of the DOT Specification 6M criticality safety re-evaluation and regulatory requirements. The examination was based upon comparative analyses of current authorized fissile material load limits for the 6M, current Federal regulations (and interpretations) limiting the contents of Type B fissile material packages, limiting aggregates of fissile material packages, and recent proposed fissile material mass limits derived from specialized criticality safety analyses of the 6M package. The work examines influences on cost in transportation, handling, and storage of fissile materials. Depending upon facility throughput requirements (and assumed incremental costs of fissile material packaging, storage, and transport), operating, facility storage capacity, and transportation costs can be reduced significantly. As an example of the pricing algorithm application based upon reasonable cost influences, the magnitude of the first year cost reductions could extend beyond four times the cost of the packaging nuclear criticality safety re-evaluation. 1 tab

SRP reactor safety evolution

International Nuclear Information System (INIS)

Rankin, D.B.

1984-01-01

The Savannah River Plant reactors have operated for over 100 reactor years without an incident of significant consequence to on or off-site personnel. The reactor safety posture incorporates a conservative, failure-tolerant design; extensive administrative controls carried out through detailed operating and emergency written procedures; and multiple engineered safety systems backed by comprehensive safety analyses, adapting through the years as operating experience, changes in reactor operational modes, equipment modernization, and experience in the nuclear power industry suggested. Independent technical reviews and audits as well as a strong organizational structure also contribute to the defense-in-depth safety posture. A complete review of safety history would discuss all of the above contributors and the interplay of roles. This report, however, is limited to evolution of the engineered safety features and some of the supporting analyses. The discussion of safety history is divided into finite periods of operating history for preservation of historical perspective and ease of understanding by the reader. Programs in progress are also included. The accident at Three Mile Island was assessed for its safety implications to SRP operation. Resulting recommendations and their current status are discussed separately at the end of the report. 16 refs., 3 figs

LFR safety approach and main ELFR safety analysis results

International Nuclear Information System (INIS)

Bubelis, E.; Schikorr, M.; Frogheri, M.; Mansani, L.; Bandini, G.; Burgazzi, L.; Mikityuk, K.; Zhang, Y.; Lo Frano, R.; Forgione, N.

2013-01-01

LFR safety approach: → A global safety approach for the LFR reference plant has been assessed and the safety analyses methodology has been developed. → LFR follows the general guidelines of the Generation IV safety concept recommendations. Thus, improved safety and higher reliability are recognized as an essential priority. → The fundamental safety objectives and the Defence-in-Depth (DiD) approach, as described by IAEA Safety Guides, have been preserved. → The recommendations of the Risk and Safety Working Group (RSWG) of GEN-IV IF has been taken into account: • safety is to be “built-in” in the fundamental design rather than “added on”; • full implementation of the Defence-in-Depth principles in a manner that is demonstrably exhaustive, progressive, tolerant, forgiving and well-balanced; • “risk-informed” approach - deterministic approach complemented with a probabilistic one; • adoption of an integrated methodology that can be used to evaluate and document the safety of Gen IV nuclear systems - ISAM. In particular the OPT tool is the fundamental methodology used throughout the design process

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.)

Intensive care nurses' perceptions of simulation-based team training for building patient safety in intensive care: a descriptive qualitative study.

Science.gov (United States)

Ballangrud, Randi; Hall-Lord, Marie Louise; Persenius, Mona; Hedelin, Birgitta

2014-08-01

To describe intensive care nurses' perceptions of simulation-based team training for building patient safety in intensive care. Failures in team processes are found to be contributory factors to incidents in an intensive care environment. Simulation-based training is recommended as a method to make health-care personnel aware of the importance of team working and to improve their competencies. The study uses a qualitative descriptive design. Individual qualitative interviews were conducted with 18 intensive care nurses from May to December 2009, all of which had attended a simulation-based team training programme. The interviews were analysed by qualitative content analysis. One main category emerged to illuminate the intensive care nurse perception: "training increases awareness of clinical practice and acknowledges the importance of structured work in teams". Three generic categories were found: "realistic training contributes to safe care", "reflection and openness motivates learning" and "finding a common understanding of team performance". Simulation-based team training makes intensive care nurses more prepared to care for severely ill patients. Team training creates a common understanding of how to work in teams with regard to patient safety. Copyright © 2014 Elsevier Ltd. All rights reserved.

NPP Temelin safety analysis reports and PSA status

International Nuclear Information System (INIS)

Mlady, O.

1999-01-01

To enhance the safety level of Temelin NPP, recommendations of the international reviews were implemented into the design as well as into organization of the plant construction and preparation for operation. The safety assessment of these design changes has been integrated and reflected in the Safety Analysis Reports, which follow the internationally accepted guidelines. All safety analyses within Safety Analysis Reports were repeated carefully considering technical improvements and replacements to complement preliminary safety documentation. These analyses were performed by advanced western computer codes to the depth and in the structure required by western standards. The Temelin NPP followed a systematic approach in the functional design of the Reactor Protection System and related safety analyses. Modifications of reactor protection system increase defense in depth and facilitate demonstrating that LOCA and radiological limits are met for non-LOCA events. The rigorous safety analysis methodology provides assurance that LOCA and radiological limits are met. Established and accepted safety analysis methodology and accepted criteria were applied to Temelin NPP meeting US NRC and Czech Republic requirements. IAEA guidelines and recommendations

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

In service monitoring based on fatigue analyses, possibilities and limitations

International Nuclear Information System (INIS)

Dittmar, S.; Binder, F.

2004-01-01

German LWR reactors are equipped with monitoring systems which are to enable a comparison of real transients with load case catalogues and fatigue catalogues for fatigue analyses. The information accuracy depends on the accuracy of measurements, on the consideration of parameters influencing fatigue (medium, component surface, component size, etc.), and on the accuracy of the load analyses. The contribution attempts a critical evaluation, also inview of the fact that real fatigue damage often are impossible to quantify on the basis of fatigue analyses at a later stage. The effects of the consideration or non-consideration of various influencing factors are discussed, as well as the consequences of the scatter of material characteristics on which the analyses are based. Possible measures to be taken in operational monitoring are derived. (orig.) [de

Preliminary scoping safety analyses of the limiting design basis protected accidents for the Fast Flux Test Facility tritium production core

International Nuclear Information System (INIS)

Heard, F.J.

1997-01-01

The SAS4A/SASSYS-l computer code is used to perform a series of analyses for the limiting protected design basis transient events given a representative tritium and medical isotope production core design proposed for the Fast Flux Test Facility. The FFTF tritium and isotope production mission will require a different core loading which features higher enrichment fuel, tritium targets, and medical isotope production assemblies. Changes in several key core parameters, such as the Doppler coefficient and delayed neutron fraction will affect the transient response of the reactor. Both reactivity insertion and reduction of heat removal events were analyzed. The analysis methods and modeling assumptions are described. Results of the analyses and comparison against fuel pin performance criteria are presented to provide quantification that the plant protection system is adequate to maintain the necessary safety margins and assure cladding integrity

Probabilistic safety criteria at the safety function/system level

International Nuclear Information System (INIS)

1989-09-01

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

Evaluation of aviation-based safety team training in a hospital in The Netherlands.

Science.gov (United States)

De Korne, Dirk F; Van Wijngaarden, Jeroen D H; Van Dyck, Cathy; Hiddema, U Francis; Klazinga, Niek S

2014-01-01

The purpose of this paper is to evaluate the implementation of a broad-scale team resource management (TRM) program on safety culture in a Dutch eye hospital, detailing the program's content and procedures. Aviation-based TRM training is recognized as a useful approach to increase patient safety, but little is known about how it affects safety culture. Pre- and post-assessments of the hospitals' safety culture was based on interviews with ophthalmologists, anesthesiologists, residents, nurses, and support staff. Interim observations were made at training sessions and in daily hospital practice. The program consisted of safety audits of processes and (team) activities, interactive classroom training sessions by aviation experts, a flight simulator session, and video recording of team activities with subsequent feedback. Medical professionals considered aviation experts inspiring role models and respected their non-hierarchical external perspective and focus on medical-technical issues. The post-assessment showed that ophthalmologists and other hospital staff had become increasingly aware of safety issues. The multidisciplinary approach promoted social (team) orientation that replaced the former functionally-oriented culture. The number of reported near-incidents greatly increased; the number of wrong-side surgeries stabilized to a minimum after an initial substantial reduction. The study was observational and the hospital's variety of efforts to improve safety culture prevented us from establishing a causal relation between improvement and any one specific intervention. Aviation-based TRM training can be a useful to stimulate safety culture in hospitals. Safety and quality improvements are not single treatment interventions but complex socio-technical interventions. A multidisciplinary system approach and focus on "team" instead of "profession" seems both necessary and difficult in hospital care.

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.

Perceived school safety is strongly associated with adolescent mental health problems.

Science.gov (United States)

Nijs, Miesje M; Bun, Clothilde J E; Tempelaar, Wanda M; de Wit, Niek J; Burger, Huibert; Plevier, Carolien M; Boks, Marco P M

2014-02-01

School environment is an important determinant of psychosocial function and may also be related to mental health. We therefore investigated whether perceived school safety, a simple measure of this environment, is related to mental health problems. In a population-based sample of 11,130 secondary school students, we analysed the relationship of perceived school safety with mental health problems using multiple logistic regression analyses to adjust for potential confounders. Mental health problems were defined using the clinical cut-off of the self-reported Strengths and Difficulties Questionnaire. School safety showed an exposure-response relationship with mental health problems after adjustment for confounders. Odds ratios increased from 2.48 ("sometimes unsafe") to 8.05 ("very often unsafe"). The association was strongest in girls and young and middle-aged adolescents. Irrespective of the causal background of this association, school safety deserves attention either as a risk factor or as an indicator of mental health problems.

Behavior of underclad cracks in reactor pressure vessels - evaluation of mechanical analyses with tests on cladded mock-ups

International Nuclear Information System (INIS)

Moinereau, D.; Rousselier, G.; Bethmont, M.

1993-01-01

Innocuity of underclad flaws in the reactor pressure vessels must be demonstrated in the French safety analyses, particularly in the case of a severe transient at the end of the pressure vessel lifetime, because of the radiation embrittlement of the vessel material. Safety analyses are usually performed with elastic and elasto-plastic analyses taking into account the effect of the stainless steel cladding. EDF has started a program including experiments on large size cladded specimens and their interpretations. The purpose of this program is to evaluate the different methods of fracture analysis used in safety studies. Several specimens made of ferritic steel A508 C1 3 with stainless steel cladding, containing small artificial defects, are loaded in four-point bending. Experiments are performed at very low temperature to simulate radiation embrittlement and to obtain crack instability by cleavage fracture. Three tests have been performed on mock-ups containing a small underclad crack (with depth about 5 mn) and a fourth test has been performed on one mock-up with a larger crack (depth about 13 mn). In each case, crack instability occurred by cleavage fracture in the base metal, without crack arrest, at a temperature of about - 170 deg C. Each test is interpreted using linear elastic analysis and elastic-plastic analysis by two-dimensional finite element computations. The fracture are conservatively predicted: the stress intensity factors deduced from the computations (K cp or K j ) are always greater than the base metal toughness. The comparison between the elastic analyses (including two plasticity corrections) and the elastic-plastic analyses shows that the elastic analyses are often conservative. The beneficial effect of the cladding in the analyses is also shown : the analyses are too conservative if the cladding effects is not taken into account. (authors). 9 figs., 6 tabs., 10 refs

Safety climate and injuries: an examination of theoretical and empirical relationships.

Science.gov (United States)

Beus, Jeremy M; Payne, Stephanie C; Bergman, Mindy E; Arthur, Winfred

2010-07-01

Our purpose in this study was to meta-analytically address several theoretical and empirical issues regarding the relationships between safety climate and injuries. First, we distinguished between extant safety climate-->injury and injury-->safety climate relationships for both organizational and psychological safety climates. Second, we examined several potential moderators of these relationships. Meta-analyses revealed that injuries were more predictive of organizational safety climate than safety climate was predictive of injuries. Additionally, the injury-->safety climate relationship was stronger for organizational climate than for psychological climate. Moderator analyses revealed that the degree of content contamination in safety climate measures inflated effects, whereas measurement deficiency attenuated effects. Additionally, moderator analyses showed that as the time period over which injuries were assessed lengthened, the safety climate-->injury relationship was attenuated. Supplemental meta-analyses of specific safety climate dimensions also revealed that perceived management commitment to safety is the most robust predictor of occupational injuries. Contrary to expectations, the operationalization of injuries did not meaningfully moderate safety climate-injury relationships. Implications and recommendations for future research and practice are discussed.

Ageing study of the engineered safety features actuation system of the Loviisa NPP

International Nuclear Information System (INIS)

Simola, K.; Maskuniitty, M.

1995-06-01

An ageing study of the engineered safety features actuation system of the Loviisa nuclear power plant has been performed. The operating experience, including failure and maintenance histories of analog measuring devices, logics for safety signal formation and individual control electronics of pumps and valves, has been collected and analysed. The safety importance of system components has been studied with a fault tree analysis of a selected safety function. Based on the results of the analysis of operating experiences and the fault tree analysis, some components were selected for deeper analyses. According to the operating experience, the amount of failures in the Loviisa plant safety system has been low and no increasing trend in the failure history can yet be observed. Only a few failures had prohibited the propagation of the safety signal, mostly the failures have caused a false alarm. The failures reported have concerned mainly limit signal units, transmitters, and priority units. According to the fault tree analysis of one safety function, the most important components of this subsystem are individual control units and pulse/DC converters. Failure modes and effect analyses were performed for priority and individual control unit, limit signal unit and comparator and pulse/DC converter in order to identify the critical failure modes of these devices. (orig.) (15 refs., 26 figs., 9 tabs.)

A dynamic Bayesian network based approach to safety decision support in tunnel construction

International Nuclear Information System (INIS)

Wu, Xianguo; Liu, Huitao; Zhang, Limao; Skibniewski, Miroslaw J.; Deng, Qianli; Teng, Jiaying

2015-01-01

This paper presents a systemic decision approach with step-by-step procedures based on dynamic Bayesian network (DBN), aiming to provide guidelines for dynamic safety analysis of the tunnel-induced road surface damage over time. The proposed DBN-based approach can accurately illustrate the dynamic and updated feature of geological, design and mechanical variables as the construction progress evolves, in order to overcome deficiencies of traditional fault analysis methods. Adopting the predictive, sensitivity and diagnostic analysis techniques in the DBN inference, this approach is able to perform feed-forward, concurrent and back-forward control respectively on a quantitative basis, and provide real-time support before and after an accident. A case study in relating to dynamic safety analysis in the construction of Wuhan Yangtze Metro Tunnel in China is used to verify the feasibility of the proposed approach, as well as its application potential. The relationships between the DBN-based and BN-based approaches are further discussed according to analysis results. The proposed approach can be used as a decision tool to provide support for safety analysis in tunnel construction, and thus increase the likelihood of a successful project in a dynamic project environment. - Highlights: • A dynamic Bayesian network (DBN) based approach for safety decision support is developed. • This approach is able to perform feed-forward, concurrent and back-forward analysis and control. • A case concerning dynamic safety analysis in Wuhan Yangtze Metro Tunnel in China is presented. • DBN-based approach can perform a higher accuracy than traditional static BN-based approach

Exploring the relationship between quality of management and safety climate in a large scale Danish cross‐sectional study

DEFF Research Database (Denmark)

Sønderstrup-Andersen, Hans H. K.; Carlsen, Kathrine; Kines, Pete

2011-01-01

. In addition, we consider the impact of age, gender, education, job type and seniority as well as company size and industrial sector on the rating of safety climate. Predictors of safety climate ratings are analysed by use of multiple regression analysis. Our results show that the leadership style measured...... and transformational leadership and safety climate, and to explore how safety climate is affected by a number of socio-demographic factors and within different industries and company sizes. The analyses are based on data from a recent Danish work environment cross-sectional study including 3681 employees from a wide...... range of industries and who report that safety climate is relevant for their job. We use two safety climate items, (one regarding management safety empowerment; one regarding co-workers’ safety priority), one question about transactional leadership and two scales concerning transformational leadership...

'BeSAFE', effect-evaluation of internet-based, tailored safety information combined with personal counselling on parents' child safety behaviours: study design of a randomized controlled trial

Directory of Open Access Journals (Sweden)

van Beeck Eduard F

2010-08-01

Full Text Available Abstract Background Injuries in or around the home are the most important cause of death among children aged 0-4 years old. It is also a major source of morbidity and loss of quality of life. In order to reduce the number of injuries, the Consumer Safety Institute introduced the use of Safety Information Leaflets in the Netherlands to provide safety education to parents of children aged 0-4 years. Despite current safety education, necessary safety behaviours are still not taken by a large number of parents, causing unnecessary risk of injury among young children. In an earlier study an E-health module with internet-based, tailored safety information was developed and applied. It concerns an advice for parents on safety behaviours in their homes regarding their child. The aim of this study is to evaluate the effect of this safety information combined with personal counselling on parents' child safety behaviours. Methods/Design Parents who are eligible for the regular well-child visit with their child at child age 5-8 months are invited to participate in this study. Participating parents are randomized into one of two groups: 1 internet-based, tailored safety information combined with personal counselling (intervention group, or 2 personal counselling using the Safety Information Leaflets of the Consumer Safety Institute in the Netherlands for children aged 12 to 24 months (control group. All parents receive safety information on safety topics regarding the prevention of falling, poisoning, drowning and burning. Parents of the intervention group will access the internet-based, tailored safety information module when their child is approximately 10 months old. After completion of the assessment questions, the program compiles a tailored safety advice. The parents are asked to devise and inscribe a personal implementation intention. During the next well-child visit, the Child Health Clinic professional will discuss this tailored safety information

Seismic and tsunami safety margin assessment

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-08-15

Nuclear Regulation Authority is going to establish new seismic and tsunami safety guidelines to increase the safety of NPPs. The main purpose of this research is testing structures/components important to safety and tsunami resistant structures/components, and evaluating the capacity of them against earthquake and tsunami. Those capacity data will be utilized for the seismic and tsunami back-fit review based on the new seismic and tsunami safety guidelines. The summary of the program in 2012 is as follows. 1. Component seismic capacity test and quantitative seismic capacity evaluation. PWR emergency diesel generator partial-model seismic capacity tests have been conducted and quantitative seismic capacities have been evaluated. 2. Seismic capacity evaluation of switching-station electric equipment. Existing seismic test data investigation, specification survey and seismic response analyses have been conducted. 3. Tsunami capacity evaluation of anti-inundation measure facilities. Tsunami pressure test have been conducted utilizing a small breakwater model and evaluated basic characteristics of tsunami pressure against seawall structure. (author)

Seismic and tsunami safety margin assessment